Testing in Support of Fission Surface Power System Qualification

NASA Technical Reports Server (NTRS)

Houts, Mike; Bragg-Sitton, Shannon; Godfroy, Tom; Martin, Jim; Pearson, Boise; VanDyke, Melissa

2007-01-01

The strategy for qualifying a FSP system could have a significant programmatic impact. The US has not qualified a space fission power system since launch of the SNAP-10A in 1965. This paper explores cost-effective options for obtaining data that would be needed for flight qualification of a fission system. Qualification data could be obtained from both nuclear and non-nuclear testing. The ability to perform highly realistic nonnuclear testing has advanced significantly throughout the past four decades. Instrumented thermal simulators were developed during the 1970s and 1980s to assist in the development, operation, and assessment of terrestrial fission systems. Instrumented thermal simulators optimized for assisting in the development, operation, and assessment of modern FSP systems have been under development (and utilized) since 1998. These thermal simulators enable heat from fission to be closely mimicked (axial power profile, radial power profile, temperature, heat flux, etc.) and extensive data to be taken from the core region. For transient testing, pin power during a transient is calculated based on the reactivity feedback that would occur given measured values of test article temperature and/or dimensional changes. The reactivity feedback coefficients needed for the test are either calculated or measured using cold/warm zero-power criticals. In this way non-nuclear testing can be used to provide very realistic information related to nuclear operation. Non-nuclear testing can be used at all levels, including component, subsystem, and integrated system testing. FSP fuels and materials are typically chosen to ensure very high confidence in operation at design burnups, fluences, and temperatures. However, facilities exist (e.g. ATR, HFIR) for affordably performing in-pile fuel and materials irradiations, if such testing is desired. Ex-core materials and components (such as alternator materials, control drum drives, etc.) could be irradiated in university or DOE reactors to ensure adequate radiation resistance. Facilities also exist for performing warm and cold zero-power criticals.

Ground test facility for SEI nuclear rocket engines

NASA Astrophysics Data System (ADS)

Harmon, Charles D.; Ottinger, Cathy A.; Sanchez, Lawrence C.; Shipers, Larry R.

1992-07-01

Nuclear (fission) thermal propulsion has been identified as a critical technology for a manned mission to Mars by the year 2019. Facilities are required that will support ground tests to qualify the nuclear rocket engine design, which must support a realistic thermal and neutronic environment in which the fuel elements will operate at a fraction of the power for a flight weight reactor/engine. This paper describes the design of a fuel element ground test facility, with a strong emphasis on safety and economy. The details of major structures and support systems of the facility are discussed, and a design diagram of the test facility structures is presented.

Fission-gas-release rates from irradiated uranium nitride specimens

NASA Technical Reports Server (NTRS)

Weinstein, M. B.; Kirchgessner, T. A.; Tambling, T. N.

1973-01-01

Fission-gas-release rates from two 93 percent dense UN specimens were measured using a sweep gas facility. Specimen burnup rates averaged .0045 and .0032 percent/hr, and the specimen temperatures ranged from 425 to 1323 K and from 552 to 1502 K, respectively. Burnups up to 7.8 percent were achieved. Fission-gas-release rates first decreased then increased with burnup. Extensive interconnected intergranular porosity formed in the specimen operated at over 1500 K. Release rate variation with both burnup and temperature agreed with previous irradiation test results.

Fission Activities of the Nuclear Reactions Group in Uppsala

NASA Astrophysics Data System (ADS)

Al-Adili, A.; Alhassan, E.; Gustavsson, C.; Helgesson, P.; Jansson, K.; Koning, A.; Lantz, M.; Mattera, A.; Prokofiev, A. V.; Rakopoulos, V.; Sjöstrand, H.; Solders, A.; Tarrío, D.; Österlund, M.; Pomp, S.

This paper highlights some of the main activities related to fission of the nuclear reactions group at Uppsala University. The group is involved for instance in fission yield experiments at the IGISOL facility, cross-section measurements at the NFS facility, as well as fission dynamics studies at the IRMM JRC-EC. Moreover, work is ongoing on the Total Monte Carlo (TMC) methodology and on including the GEF fission code into the TALYS nuclear reaction code. Selected results from these projects are discussed.

Simulations for the future converter of the e-linac for the TRIUMF ARIEL facility

NASA Astrophysics Data System (ADS)

Lebois, M.; Bricault, P.

2011-09-01

In the next years, TRIUMF activity will be focused on building a new facility to produce very intense neutron rich radioactive ion beams. Unlike others ISOL facilities, the e-linac primary beam, that will induce the fission, is an intense electron beam (50 MeV energy and 10 mA intensity). This challenging choice, which make this installation unique, despite the ALTO facility, makes an average fission rate of 1013-14fissions/s in the target.This beam is sent on an uranium carbide target (UCx), but due to its power, it is essential to insert a "converter" on the beam path to avoid a target overheating. The purpose of this converter is to convert electrons into Bremsstralhung radiation. The γ rays produce excite the dipole resonance of 23892U (15 MeV) inducing fission. Energy deposition, fission rate and thermal behavior were simulated using Monte Carlo techniques are presented in this paper

Overview of the ISOL facility for the RISP

NASA Astrophysics Data System (ADS)

Woo, H. J.; Kang, B. H.; Tshoo, K.; Seo, C. S.; Hwang, W.; Park, Y.-H.; Yoon, J. W.; Yoo, S. H.; Kim, Y. K.; Jang, D. Y.

2015-02-01

The key feature of the Isotope Separation On-Line (ISOL) facility is its ability to provide high-intensity and high-quality beams of neutron-rich isotopes with masses in the range of 80-160 by means of a 70-MeV proton beam directly impinging on uranium-carbide thin-disc targets to perform forefront research in nuclear structure, nuclear astrophysics, reaction dynamics and interdisciplinary fields like medical, biological and material sciences. The technical design of the 10-kW and the 35-kW direct fission targets with in-target fission rates of up to 1014 fissions/s has been finished, and for the development of the ISOL fission-target chemistry an initial effort has been made to produce porous lanthanum-carbide (LaCx) discs as a benchmark for the final production of porous UCx discs. For the production of various beams, three classes of ion sources are under development at RISP (Rare Isotope Science Project), the surface ion source, the plasma ion source (FEBIAD), the laser ion source, and the engineering design of the FEBIAD is in progress for prototype fabrication. The engineering design of the ISOL target/ion source front-end system is also in progress, and a prototype will be used for an off-line test facility in front of the pre-separator. The technical designs of other basic elements at the ISOL facility, such as the RF-cooler, the high-resolution mass separator, and the A/q separator, have been finished, and the results, along with the future plans, are introduced.

Capabilities and Testing of the Fission Surface Power Primary Test Circuit (FSP-PTC)

NASA Technical Reports Server (NTRS)

Garber, Anne E.

2007-01-01

An actively pumped alkali metal flow circuit, designed and fabricated at the NASA Marshall Space Flight Center, is currently undergoing testing in the Early Flight Fission Test Facility (EFF-TF). Sodium potassium (NaK), which was used in the SNAP-10A fission reactor, was selected as the primary coolant. Basic circuit components include: simulated reactor core, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, liquid metal flowmeter, load/drain reservoir, expansion reservoir, test section, and instrumentation. Operation of the circuit is based around a 37-pin partial-array core (pin and flow path dimensions are the same as those in a full core), designed to operate at 33 kWt. NaK flow rates of greater than 1 kg/sec may be achieved, depending upon the power applied to the EM pump. The heat exchanger provides for the removal of thermal energy from the circuit, simulating the presence of an energy conversion system. The presence of the test section increases the versatility of the circuit. A second liquid metal pump, an energy conversion system, and highly instrumented thermal simulators are all being considered for inclusion within the test section. This paper summarizes the capabilities and ongoing testing of the Fission Surface Power Primary Test Circuit (FSP-PTC).

Hardware Progress Made in the Early Flight Fission Test Facilities (EFF-TF) To Support Near-Term Space Fission Systems

NASA Technical Reports Server (NTRS)

VanDyke, Melissa; Martin, James

2005-01-01

The EFF-TF provides a facility to experimentally evaluate thermal hydraulic issues through the use of highly effective non-nuclear testing. These techniques provide a rapid, more cost effective method of evaluating designs and support development risk mitigation when concerns are associated with non-nuclear aspects of space nuclear systems. For many systems, electrical resistance thermal simulators can be used to closely mimic the heat deposition of the fission process, providing axial and radial profiles. A number of experimental and design programs were underway in 2004. Initial evaluation of the SAFE-100a (19 module stainless steel/sodium heat pipe reactor with integral gas neat exchanger) was performed with tests up to 17.5 kW of input power at core temperatures of 1000 K. A stainless steel sodium SAFE-100 heat pipe module was placed through repeated freeze/thaw cyclic testing accumulating over 200 restarts to a temperature of 1000 K. Additionally, the design of a 37-fuel pin stainless steel pumped sodium/potassium (NaK) loop was finalized and components procured. Ongoing testing at the EFF-TF is geared towards facilitating both research and development necessary to field a near term space nuclear system. Efforts are coordinated with DOE laboratories, industry, universities, and other NASA centers. This paper describes some of the 2004 efforts.

Determination of relative krypton fission product yields from 14 MeV neutron induced fission of 238U at the National Ignition Facility

DOE PAGES

Edwards, E. R.; Cassata, W. S.; Velsko, C. A.; ...

2016-09-22

Precisely-known fission yield distributions are needed to determine a fissioning isotope and the incident neutron energy in nuclear security applications. 14 MeV neutrons from DT fusion at the National Ignition Facility (NIF) induce fission in depleted uranium (DU) contained in the target assembly hohlraum. The fission yields of Kr isotopes (85m, 87, 88, and 89) are measured relative to the cumulative yield of 88Kr and compared to previously tabulated values. Here, the results from this experiment and England and Rider are in agreement, except for the 85mKr/ 88Kr ratio, which may be the result of incorrect nuclear data.

Determination of relative krypton fission product yields from 14 MeV neutron induced fission of 238U at the National Ignition Facility.

PubMed

Edwards, E R; Cassata, W S; Velsko, C A; Yeamans, C B; Shaughnessy, D A

2016-11-01

Precisely-known fission yield distributions are needed to determine a fissioning isotope and the incident neutron energy in nuclear security applications. 14 MeV neutrons from DT fusion at the National Ignition Facility induce fission in depleted uranium contained in the target assembly hohlraum. The fission yields of Kr isotopes (85m, 87, 88, and 89) are measured relative to the cumulative yield of 88 Kr and compared to previously tabulated values. The results from this experiment and England and Rider are in agreement, except for the 85m Kr/ 88 Kr ratio, which may be the result of incorrect nuclear data.

Fission yield measurements at IGISOL

NASA Astrophysics Data System (ADS)

Lantz, M.; Al-Adili, A.; Gorelov, D.; Jokinen, A.; Kolhinen, V. S.; Mattera, A.; Moore, I.; Penttilä, H.; Pomp, S.; Prokofiev, A. V.; Rakopoulos, V.; Rinta-Antila, S.; Simutkin, V.; Solders, A.

2016-06-01

The fission product yields are an important characteristic of the fission process. In fundamental physics, knowledge of the yield distributions is needed to better understand the fission process. For nuclear energy applications good knowledge of neutroninduced fission-product yields is important for the safe and efficient operation of nuclear power plants. With the Ion Guide Isotope Separator On-Line (IGISOL) technique, products of nuclear reactions are stopped in a buffer gas and then extracted and separated by mass. Thanks to the high resolving power of the JYFLTRAP Penning trap, at University of Jyväskylä, fission products can be isobarically separated, making it possible to measure relative independent fission yields. In some cases it is even possible to resolve isomeric states from the ground state, permitting measurements of isomeric yield ratios. So far the reactions U(p,f) and Th(p,f) have been studied using the IGISOL-JYFLTRAP facility. Recently, a neutron converter target has been developed utilizing the Be(p,xn) reaction. We here present the IGISOL-technique for fission yield measurements and some of the results from the measurements on proton induced fission. We also present the development of the neutron converter target, the characterization of the neutron field and the first tests with neutron-induced fission.

Prompt fission neutron spectra from fission induced by 1 to 8 MeV neutrons on U235 and Pu239 using the double time-of-flight technique

NASA Astrophysics Data System (ADS)

Noda, S.; Haight, R. C.; Nelson, R. O.; Devlin, M.; O'Donnell, J. M.; Chatillon, A.; Granier, T.; Bélier, G.; Taieb, J.; Kawano, T.; Talou, P.

2011-03-01

Prompt fission neutron spectra from U235 and Pu239 were measured for incident neutron energies from 1 to 200 MeV at the Weapons Neutron Research facility (WNR) of the Los Alamos Neutron Science Center, and the experimental data were analyzed with the Los Alamos model for the incident neutron energies of 1-8 MeV. A CEA multiple-foil fission chamber containing deposits of 100 mg U235 and 90 mg Pu239 detected fission events. Outgoing neutrons were detected by the Fast Neutron-Induced γ-Ray Observer array of 20 liquid organic scintillators. A double time-of-flight technique was used to deduce the neutron incident energies from the spallation target and the outgoing energies from the fission chamber. These data were used for testing the Los Alamos model, and the total kinetic energy parameters were optimized to obtain a best fit to the data. The prompt fission neutron spectra were also compared with the Evaluated Nuclear Data File (ENDF/B-VII.0). We calculate average energies from both experimental and calculated fission neutron spectra.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Iseki, Tadahiro; Inaba, Makoto; Takahashi, Naoki

During the second and third steps of Active Test at Rokkasho Reprocessing Plant (RRP), the performances of the Separation Facility have been checked; (A) diluent washing efficiency, (B) plutonium stripping efficiency, (C) decontamination factor of fission products and (D) plutonium and uranium leakage into raffinate and spent solvent. Test results were equivalent to or better than expected. (authors)

Kilopower: Small and Affordable Fission Power Systems for Space

NASA Technical Reports Server (NTRS)

Mason, Lee; Palac, Don; Gibson, Marc

2017-01-01

The Nuclear Systems Kilopower Project was initiated by NASA's Space Technology Mission Directorate Game Changing Development Program in fiscal year 2015 to demonstrate subsystem-level technology readiness of small space fission power in a relevant environment (Technology Readiness Level 5) for space science and human exploration power needs. The Nuclear Systems Kilopower Project centerpiece is the Kilopower Reactor Using Stirling Technology (KRUSTY) test, which consists of the development and testing of a fission ground technology demonstrator of a 1 kWe-class fission power system. The technologies to be developed and validated by KRUSTY are extensible to space fission power systems from 1 to 10 kWe, which can enable higher power future potential deep space science missions, as well as modular surface fission power systems for exploration. The Kilopower Project is cofounded by NASA and the Department of Energy National Nuclear Security Administration (NNSA).KRUSTY include the reactor core, heat pipes to transfer the heat from the core to the power conversion system, and the power conversion system. Los Alamos National Laboratory leads the design of the reactor, and the Y-12 National Security Complex is fabricating it. NASA Glenn Research Center (GRC) has designed, built, and demonstrated the balance of plant heat transfer and power conversion portions of the KRUSTY experiment. NASA MSFC developed an electrical reactor simulator for non-nuclear testing, and the design of the reflector and shielding for nuclear testing. In 2016, an electrically heated non-fissionable Depleted Uranium (DU) core was tested at GRC in a configuration identical to the planned nuclear test. Once the reactor core has been fabricated and shipped to the Device Assembly Facility at the NNSAs Nevada National Security Site, the KRUSTY nuclear experiment will be assembled and tested. Completion of the KRUSTY experiment will validate the readiness of 1 to 10 kWe space fission technology for NASAs future requirements for sunlight-independent space power. An early opportunity for demonstration of In-Situ Resource Utilization (ISRU) capability on the surface of Mars is currently being considered for 2026 launch. Since a space fission system is the leading option for power generation for the first Mars human outpost, a smaller version of a planetary surface fission power system could be built to power the ISRU demonstration and ensure its end-to-end validity. Planning is underway to start the hardware development of this subscale flight demonstrator in 2018.

Application of thin-film breakdown counters for characterization of neutron field of the VESUVIO instrument at the ISIS spallation source

NASA Astrophysics Data System (ADS)

Smirnov, A. N.; Pietropaolo, A.; Prokofiev, A. V.; Rodionova, E. E.; Frost, C. D.; Ansell, S.; Schooneveld, E. M.; Gorini, G.

2012-09-01

The high-energy neutron field of the VESUVIO instrument at the ISIS facility has been characterized using the technique of thin-film breakdown counters (TFBC). The technique utilizes neutron-induced fission reactions of natU and 209Bi with detection of fission fragments by TFBCs. Experimentally determined count rates of the fragments are ≈50% higher than those calculated using spectral neutron flux simulated with the MCNPX code. This work is a part of the project to develop ChipIr, a new dedicated facility for the accelerated testing of electronic components and systems for neutron-induced single event effects in the new Target Station 2 at ISIS. The TFBC technique has shown to be applicable for on-line monitoring of the neutron flux in the neutron energy range 1-800 MeV at the position of the device under test (DUT).

Heater Development, Fabrication, and Testing: Analysis of Fabricated Heaters

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Dickens, R. E.; Farmer, J. T.; Davis, J. D.; Adams, M. R.; Martin, J. J.; Webster, K. L.

2008-01-01

Thermal simulators (highly designed heater elements) developed at the Early Flight Fission Test Facility (EFF-TF) are used to simulate the heat from nuclear fission in a variety of reactor concepts. When inserted into the reactor geometry, the purpose of the thermal simulators is to deliver thermal power to the test article in the same fashion as if nuclear fuel were present. Considerable effort has been expended to mimic heat from fission as closely as possible. To accurately represent the fuel, the simulators should be capable of matching the overall properties of the nuclear fuel rather than simply matching the fuel temperatures. This includes matching thermal stresses in the pin, pin conductivities, total core power, and core power profile (axial and radial). This Technical Memorandum discusses the historical development of the thermal simulators used in nonnuclear testing at the EFF-TF and provides a basis for the development of the current series of thermal simulators. The status of current heater fabrication and testing is assessed, providing data and analyses for both successes and failures experienced in the heater development and testing program.

Electrical-thermal-structural finite element simulation and experimental study of a plasma ion source for the production of radioactive ion beams

NASA Astrophysics Data System (ADS)

Manzolaro, M.; Meneghetti, G.; Andrighetto, A.; Vivian, G.

2016-03-01

The production target and the ion source constitute the core of the selective production of exotic species (SPES) facility. In this complex experimental apparatus for the production of radioactive ion beams, a 40 MeV, 200 μA proton beam directly impinges a uranium carbide target, generating approximately 1013 fissions per second. The transfer line enables the unstable isotopes generated by the 238U fissions in the target to reach the ion source, where they can be ionized and finally accelerated to the subsequent areas of the facility. In this work, the plasma ion source currently adopted for the SPES facility is analyzed in detail by means of electrical, thermal, and structural numerical models. Next, theoretical results are compared with the electric potential difference, temperature, and displacement measurements. Experimental tests with stable ion beams are also presented and discussed.

Comparative evaluation of solar, fission, fusion, and fossil energy resources. Part 2: Power from nuclear fission

NASA Technical Reports Server (NTRS)

Clement, J. D.

1973-01-01

Different types of nuclear fission reactors and fissionable materials are compared. Special emphasis is placed upon the environmental impact of such reactors. Graphs and charts comparing reactor facilities in the U. S. are presented.

Measurement of fission yields and isomeric yield ratios at IGISOL

NASA Astrophysics Data System (ADS)

Pomp, Stephan; Mattera, Andrea; Rakopoulos, Vasileios; Al-Adili, Ali; Lantz, Mattias; Solders, Andreas; Jansson, Kaj; Prokofiev, Alexander V.; Eronen, Tommi; Gorelov, Dimitri; Jokinen, Ari; Kankainen, Anu; Moore, Iain D.; Penttilä, Heikki; Rinta-Antila, Sami

2018-03-01

Data on fission yields and isomeric yield ratios (IYR) are tools to study the fission process, in particular the generation of angular momentum. We use the IGISOL facility with the Penning trap JYFLTRAP in Jyväskylä, Finland, for such measurements on 232Th and natU targets. Previously published fission yield data from IGISOL concern the 232Th(p,f) and 238U(p,f) reactions at 25 and 50 MeV. Recently, a neutron source, using the Be(p,n) reaction, has been developed, installed and tested. We summarize the results for (p,f) focusing on the first measurement of IYR by direct ion counting. We also present first results for IYR and relative yields for Sn and Sb isotopes in the 128-133 mass range from natU(n,f) based on γ-spectrometry. We find a staggering behaviour in the cumulative yields for Sn and a shift in the independent fission yields for Sb as compared to current evaluations. Plans for the future experimental program on fission yields and IYR measurements are discussed.

77 FR 67679 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-13

... diesel generator surveillance requirements. Margin of safety is related to the ability of the fission... surveillance tests, allowing testing in any MODE of operation. The Division 3 AC sources, including the diesel generator (DG) and its associated emergency loads are accident mitigating features, not accident initiators...

Prompt γ rays and neutrons from fission

NASA Astrophysics Data System (ADS)

Kwan, E.; Wu, C. Y.; Chyzh, A.; Gostic, J.; Henderson, R.; Haight, R. C.; Lee, H. Y.; O'Donnell, J. M.; Perdue, B. A.; Taddeucci, T. N.

2011-10-01

Nuclear data are needed to test the accuracy of calculations from nuclear reaction codes. Information on the prompt γ-ray distributions from fission is sparse and only a handful of published experiments data that measured the prompt γ-ray distribution above incident neutron energies of 1 MeV can be found. In addition, improvement on the accuracy and shape of neutron spectrum from the fission of actinides been requested by the nuclear data community. An investigation on the shapes of the neutron and γ-ray distributions from the spontaneous fission of 252Cf and the neutron-induced fission of 235U was undertaken using the Chi-Nu detector array at the Weapons Neutron Research Facility of the Los Alamos Neutron Science Center. Preliminary results will be presented. This work is performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344 and the Los Alamos National Laboratory under Contract DE-AC52-06NA25396.

Nuclear Systems Kilopower Overview

NASA Technical Reports Server (NTRS)

Palac, Don; Gibson, Marc; Mason, Lee; Houts, Michael; McClure, Patrick; Robinson, Ross

2016-01-01

The Nuclear Systems Kilopower Project was initiated by NASAs Space Technology Mission Directorate Game Changing Development Program in fiscal year 2015 to demonstrate subsystem-level technology readiness of small space fission power in a relevant environment (Technology Readiness Level 5) for space science and human exploration power needs. The Nuclear Systems Kilopower Project consists of two elements. The primary element is the Kilopower Prototype Test, also called the Kilopower Reactor Using Stirling Technology(KRUSTY) Test. This element consists of the development and testing of a fission ground technology demonstrator of a 1 kWe fission power system. A 1 kWe system matches requirements for some robotic precursor exploration systems and future potential deep space science missions, and also allows a nuclear ground technology demonstration in existing nuclear test facilities at low cost. The second element, the Mars Kilopower Scalability Study, consists of the analysis and design of a scaled-up version of the 1 kWe reference concept to 10 kWe for Mars surface power projected requirements, and validation of the applicability of the KRUSTY experiment to key technology challenges for a 10 kWe system. If successful, these two elements will lead to initiation of planning for a technology demonstration of a 10 kWe fission power capability for Mars surface outpost power.

Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR.

PubMed

Chang, G S; Ambrosek, R G

2005-01-01

The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra.

PubMed

Ďuran, I; Bolshakova, I; Viererbl, L; Sentkerestiová, J; Holyaka, R; Lahodová, Z; Bém, P

2010-10-01

We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10(16) cm(-2) was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

Lanl Neutron-Induced Fission Cross Section Measurement Program

NASA Astrophysics Data System (ADS)

Laptev, A. B.; Tovesson, F.; Hill, T. S.

2014-09-01

A well established program of neutron-induced fission cross section measurement at Los Alamos Neutron Science Center (LANSCE) is supporting the Fuel Cycle Research program (FC R&D). Combining measurements at two LANSCE facilities, the Lujan Center and the Weapons Neutron Research facility (WNR), cover neutron energies over 10 orders of magnitude: from sub-thermal up to 200 MeV. A parallel-plate fission ionization chamber was used as a fission fragment detector. The 235U(n,f) standard was used as the reference. Fission cross sections have been measured for multiple actinides. The new data presented here completes the suite of long-lived Uranium isotopes that were investigated with this experimental approach. The cross section data are presented in comparison with existing evaluations and previous measurements.

Accurate isotopic fission yields of electromagnetically induced fission of 238U measured in inverse kinematics at relativistic energies

NASA Astrophysics Data System (ADS)

Pellereau, E.; Taïeb, J.; Chatillon, A.; Alvarez-Pol, H.; Audouin, L.; Ayyad, Y.; Bélier, G.; Benlliure, J.; Boutoux, G.; Caamaño, M.; Casarejos, E.; Cortina-Gil, D.; Ebran, A.; Farget, F.; Fernández-Domínguez, B.; Gorbinet, T.; Grente, L.; Heinz, A.; Johansson, H.; Jurado, B.; Kelić-Heil, A.; Kurz, N.; Laurent, B.; Martin, J.-F.; Nociforo, C.; Paradela, C.; Pietri, S.; Rodríguez-Sánchez, J. L.; Schmidt, K.-H.; Simon, H.; Tassan-Got, L.; Vargas, J.; Voss, B.; Weick, H.

2017-05-01

SOFIA (Studies On Fission with Aladin) is a novel experimental program, dedicated to accurate measurements of fission-fragment isotopic yields. The setup allows us to fully identify, in nuclear charge and mass, both fission fragments in coincidence for the whole fission-fragment range. It was installed at the GSI facility (Darmstadt), to benefit from the relativistic heavy-ion beams available there, and thus to use inverse kinematics. This paper reports on fission yields obtained in electromagnetically induced fission of 238U.

Development and Testing of Space Fission Technology at NASA-MSFC

NASA Technical Reports Server (NTRS)

Polzin, Kurt; Pearson, J. Boise; Houts, Michael

2008-01-01

The Early Flight Fission Test Facility (EFF-TF) at NASA-Marshall Space Flight Center (MSFC) provides a capability to perform hardware-directed activities to support multiple inspace nuclear reactor concepts by using a non-nuclear test methodology. This includes fabrication and testing at both the module/component level and near prototypic reactor configurations allowing for realistic thermal-hydraulic evaluations of systems. The EFF-TF is currently performing non-nuclear testing of hardware to support a technology development effort related to an affordable fission surface power (AFSP) system that could be deployed on the Lunar surface. The AFSP system is presently based on a pumped liquid metal-cooled reactor design, which builds on US and Russian space reactor technology as well as extensive US and international terrestrial liquid metal reactor experience. An important aspect of the current hardware development effort is the information and insight that can be gained from experiments performed in a relevant environment using realistic materials. This testing can often deliver valuable data and insights with a confidence that is not otherwise available or attainable. While the project is currently focused on potential fission surface power for the lunar surface, many of the present advances, testing capabilities, and lessons learned can be applied to the future development of a low-cost in-space fission power system. The potential development of such systems would be useful in fulfilling the power requirements for certain electric propulsion systems (magnetoplasmadynamic thruster, high-power Hall and ion thrusters). In addition, inspace fission power could be applied towards meeting spacecraft and propulsion needs on missions further from the Sun, where the usefulness of solar power is diminished. The affordable nature of the fission surface power system that NASA may decide to develop in the future might make derived systems generally attractive for powering spacecraft and propulsion systems in space. This presentation will discuss work on space nuclear systems that has been performed at MSFC's EFF-TF over the past 10 years. Emphasis will be place on both ongoing work related to FSP and historical work related to in-space systems potentially useful for powering electric propulsion systems.

THAI Multi-Compartment Containment Test Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kanzleiter, T.; Poss, G.; Funke, F.

2006-07-01

The THAI experimental programme includes combined-effect investigations on thermal hydraulics, hydrogen, and fission product (iodine and aerosols) behaviour in LWR containments under severe accident conditions. An overview on the experiments performed up to now and on the future test program is presented, in combination with a selection of typical results to illustrate the versatility of the test facility and the broad variety of topics investigated. (authors)

Photon-induced Fission Product Yield Measurements on 235U, 238U, and 239Pu

NASA Astrophysics Data System (ADS)

Krishichayan, Fnu; Bhike, M.; Tonchev, A. P.; Tornow, W.

2015-10-01

During the past three years, a TUNL-LANL-LLNL collaboration has provided data on the fission product yields (FPYs) from quasi-monoenergetic neutron-induced fission of 235U, 238U, and 239Pu at TUNL in the 0.5 to 15 MeV energy range. Recently, we have extended these experiments to photo-fission. We measured the yields of fission fragments ranging from 85Kr to 147Nd from the photo-fission of 235U, 238U, and 239Pu using 13-MeV mono-energetic photon beams at the HIGS facility at TUNL. First of its kind, this measurement will provide a unique platform to explore the effect of the incoming probe on the FPYs, i.e., photons vs. neutrons. A dual-fission ionization chamber was used to determine the number of fissions in the targets and these samples (along with Au monitor foils) were gamma-ray counted in the low-background counting facility at TUNL. Details of the experimental set-up and results will be presented and compared to the FPYs obtained from neutron-induced fission at the same excitation energy of the compound nucleus. Work supported in part by the NNSA-SSAA Grant No. DE-NA0001838.

Fission Surface Power Technology Development Update

NASA Technical Reports Server (NTRS)

Palac, Donald T.; Mason, Lee S.; Houts, Michael G.; Harlow, Scott

2011-01-01

Power is a critical consideration in planning exploration of the surfaces of the Moon, Mars, and places beyond. Nuclear power is an important option, especially for locations in the solar system where sunlight is limited or environmental conditions are challenging (e.g., extreme cold, dust storms). NASA and the Department of Energy are maintaining the option for fission surface power for the Moon and Mars by developing and demonstrating technology for a fission surface power system. The Fission Surface Power Systems project has focused on subscale component and subsystem demonstrations to address the feasibility of a low-risk, low-cost approach to space nuclear power for surface missions. Laboratory demonstrations of the liquid metal pump, reactor control drum drive, power conversion, heat rejection, and power management and distribution technologies have validated that the fundamental characteristics and performance of these components and subsystems are consistent with a Fission Surface Power preliminary reference concept. In addition, subscale versions of a non-nuclear reactor simulator, using electric resistance heating in place of the reactor fuel, have been built and operated with liquid metal sodium-potassium and helium/xenon gas heat transfer loops, demonstrating the viability of establishing system-level performance and characteristics of fission surface power technologies without requiring a nuclear reactor. While some component and subsystem testing will continue through 2011 and beyond, the results to date provide sufficient confidence to proceed with system level technology readiness demonstration. To demonstrate the system level readiness of fission surface power in an operationally relevant environment (the primary goal of the Fission Surface Power Systems project), a full scale, 1/4 power Technology Demonstration Unit (TDU) is under development. The TDU will consist of a non-nuclear reactor simulator, a sodium-potassium heat transfer loop, a power conversion unit with electrical controls, and a heat rejection system with a multi-panel radiator assembly. Testing is planned at the Glenn Research Center Vacuum Facility 6 starting in 2012, with vacuum and liquid-nitrogen cold walls to provide simulation of operationally relevant environments. A nominal two-year test campaign is planned including a Phase 1 reactor simulator and power conversion test followed by a Phase 2 integrated system test with radiator panel heat rejection. The testing is expected to demonstrate the readiness and availability of fission surface power as a viable power system option for NASA's exploration needs. In addition to surface power, technology development work within this project is also directly applicable to in-space fission power and propulsion systems.

Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

NASA Technical Reports Server (NTRS)

Garber, Anne E.; Dickens, Ricky E.

2011-01-01

The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

Neutron-induced fission measurements at the time-of-flight facility nELBE

DOE PAGES

Kögler, T.; Beyer, R.; Junghans, A. R.; ...

2015-05-18

Neutron-induced fission of ²⁴²Pu is studied at the photoneutron source nELBE. The relative fast neutron fission cross section was determined using actinide fission chambers in a time-of-flight experiment. A good agreement of present nuclear data with evalua- tions has been achieved in the range of 100 keV to 10 MeV.

Fission product yield measurements using monoenergetic photon beams

NASA Astrophysics Data System (ADS)

Krishichayan; Bhike, M.; Tonchev, A. P.; Tornow, W.

2017-09-01

Measurements of fission products yields (FPYs) are an important source of information on the fission process. During the past couple of years, a TUNL-LANL-LLNL collaboration has provided data on the FPYs from quasi monoenergetic neutron-induced fission on 235U, 238U, and 239Pu and has revealed an unexpected energy dependence of both asymmetric fission fragments at energies below 4 MeV. This peculiar FPY energy dependence was more pronounced in neutron-induced fission of 239Pu. In an effort to understand and compare the effect of the incoming probe on the FPY distribution, we have carried out monoenergetic photon-induced fission experiments on the same 235U, 238U, and 239Pu targets. Monoenergetic photon beams of Eγ = 13.0 MeV were provided by the HIγS facility, the world's most intense γ-ray source. In order to determine the total number of fission events, a dual-fission chamber was used during the irradiation. These irradiated samples were counted at the TUNL's low-background γ-ray counting facility using high efficient HPGe detectors over a period of 10 weeks. Here we report on our first ever photofission product yield measurements obtained with monoenegetic photon beams. These results are compared with neutron-induced FPY data.

NASA's Nuclear Thermal Propulsion Project

NASA Technical Reports Server (NTRS)

Houts, Michael; Mitchell, Sonny; Kim, Tony; Borowski, Stanley; Power, Kevin; Scott, John; Belvin, Anthony; Clement, Steven

2015-01-01

Space fission power systems can provide a power rich environment anywhere in the solar system, independent of available sunlight. Space fission propulsion offers the potential for enabling rapid, affordable access to any point in the solar system. One type of space fission propulsion is Nuclear Thermal Propulsion (NTP). NTP systems operate by using a fission reactor to heat hydrogen to very high temperature (>2500 K) and expanding the hot hydrogen through a supersonic nozzle. First generation NTP systems are designed to have an Isp of approximately 900 s. The high Isp of NTP enables rapid crew transfer to destinations such as Mars, and can also help reduce mission cost, improve logistics (fewer launches), and provide other benefits. However, for NTP systems to be utilized they must be affordable and viable to develop. NASA's Advanced Exploration Systems (AES) NTP project is a technology development project that will help assess the affordability and viability of NTP. Early work has included fabrication of representative graphite composite fuel element segments, coating of representative graphite composite fuel element segments, fabrication of representative cermet fuel element segments, and testing of fuel element segments in the Compact Fuel Element Environmental Tester (CFEET). Near-term activities will include testing approximately 16" fuel element segments in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES), and ongoing research into improving fuel microstructure and coatings. In addition to recapturing fuels technology, affordable development, qualification, and utilization strategies must be devised. Options such as using low-enriched uranium (LEU) instead of highly-enriched uranium (HEU) are being assessed, although that option requires development of a key technology before it can be applied to NTP in the thrust range of interest. Ground test facilities will be required, especially if NTP is to be used in conjunction with high value or crewed missions. There are potential options for either modifying existing facilities or constructing new ground test facilities. At least three potential options exist for reducing (or eliminating) the release of radioactivity into the environment during ground testing. These include fully containing the NTP exhaust during the ground test, scrubbing the exhaust, or utilizing an existing borehole at the Nevada National Security Site (NNSS) to filter the exhaust. Finally, the project is considering the potential for an early flight demonstration of an engine very similar to one that could be used to support human Mars or other ambitious missions. The flight demonstration could be an important step towards the eventual utilization of NTP.

Neutron-Induced Fission Measurements at the Dance and Lsds Facilities at Lanl

NASA Astrophysics Data System (ADS)

Jandel, M.; Bredeweg, T. A.; Bond, E. M.; Chadwick, M. B.; Couture, A.; O'Donnell, J. M.; Fowler, M. M.; Haight, R. C.; Hayes-Sterbenz, A. C.; Rundberg, R. S.; Rusev, G. Y.; Ullmann, J. L.; Vieira, D. J.; Wilhelmy, J. B.; Wu, C. Y.; Becker, J. A.; Alexander, C. W.; Belier, G.

2014-09-01

New results from neutron-induced fission measurements performed at the Detector for Advanced Neutron Capture Experiments (DANCE) and Lead Slowing Down Spectrometer (LSDS) are presented. New correlated data on promptfission γ-ray (PFG) distributions were measured using the DANCE array for resonant neutron-induced fission of 233U, 235U and 239Pu. The deduced properties of PFG emission are presented using a simple parametrization. An accurate knowledge of fission γ-ray spectra enables us to analyze the isomeric states of 236U created after neutron capture on 235U. We briefly discuss these new results. Finally, we review details and preliminary results of the challenging 237U(n,f) cross section measurement at the LSDS facility.

Determination of gaseous fission product yields from 14 MeV neutron induced fission of 238U at the National Ignition Facility

DOE PAGES

Cassata, W. S.; Velsko, C. A.; Stoeffl, W.; ...

2016-01-14

We determined fission yields of xenon ( 133mXe, 135Xe, 135mXe, 137Xe, 138Xe, and 139Xe) resulting from 14 MeV neutron induced fission of depleted uranium at the National Ignition Facility. Measurements begin approximately 20 s after shot time, and yields have been determined for nuclides with half-lives as short as tens of seconds. We determined the relative independent yields of 133mXe, 135Xe, and 135mXe to significantly higher precision than previously reported. The relative fission yields of all nuclides are statistically indistinguishable from values reported by England and Rider (ENDF-349. LA-UR-94-3106, 1994), with exception of the cumulative yield of 139Xe. Furthermore, considerablemore » differences exist between our measured yields and the JEFF-3.1 database values.« less

Benchmark experiments at ASTRA facility on definition of space distribution of {sup 235}U fission reaction rate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bobrov, A. A.; Boyarinov, V. F.; Glushkov, A. E.

2012-07-01

Results of critical experiments performed at five ASTRA facility configurations modeling the high-temperature helium-cooled graphite-moderated reactors are presented. Results of experiments on definition of space distribution of {sup 235}U fission reaction rate performed at four from these five configurations are presented more detail. Analysis of available information showed that all experiments on criticality at these five configurations are acceptable for use them as critical benchmark experiments. All experiments on definition of space distribution of {sup 235}U fission reaction rate are acceptable for use them as physical benchmark experiments. (authors)

Fission of actinide nuclei using multi-nucleon transfer reactions

NASA Astrophysics Data System (ADS)

Léguillon, Romain; Nishio, Katsuhisa; Hirose, Kentaro; Orlandi, Riccardo; Makii, Hiroyuki; Nishinaka, Ichiro; Ishii, Tetsuro; Tsukada, Kazuaki; Asai, Masato; Chiba, Satoshi; Ohtsuki, Tsutomu; Araki, Shohei; Watanabe, Yukinobu; Tatsuzawa, Ryotaro; Takaki, Naoyuki

2014-09-01

We are promoting a campaign to measure fission-fragment mass distributions for neutron-rich actinide nuclei populated by transfer reactions from their ground state up to an excitation energy of several tens MeV. We thus obtain the excitation energy dependence of the mass distribution. The experiment was carried out at the 20 MV JAEA tandem facility at Tokai. We report on the data obtained in the direct reaction 18 O + 232 Th . Transfer-channels and excitation energies of the fissioning nuclei were identified using silicon dE-E detectors located at forward angle. Two fission fragments were detected in coincidence using multi-wire proportional counters. Fission fragment masses were determined by kinematic consideration. We obtained the fission fragment mass distributions for 13 nuclei from actinium to uranium and some fission barrier heights. We are promoting a campaign to measure fission-fragment mass distributions for neutron-rich actinide nuclei populated by transfer reactions from their ground state up to an excitation energy of several tens MeV. We thus obtain the excitation energy dependence of the mass distribution. The experiment was carried out at the 20 MV JAEA tandem facility at Tokai. We report on the data obtained in the direct reaction 18 O + 232 Th . Transfer-channels and excitation energies of the fissioning nuclei were identified using silicon dE-E detectors located at forward angle. Two fission fragments were detected in coincidence using multi-wire proportional counters. Fission fragment masses were determined by kinematic consideration. We obtained the fission fragment mass distributions for 13 nuclei from actinium to uranium and some fission barrier heights. Present study is supported by the Ministry of Education, Culture, Sports, Science and Technology of Japan.

The South African isotope facility project

NASA Astrophysics Data System (ADS)

Bark, R. A.; Barnard, A. H.; Conradie, J. L.; de Villiers, J. G.; van Schalkwyk, P. A.

2018-05-01

The South African Isotope Facility (SAIF) is a project in which iThemba LABS plans to build a radioactive-ion beam (RIB) facility. The project is divided into the Accelerator Centre of Exotic Isotopes (ACE Isotopes) and the Accelerator Centre for Exotic Beams (ACE Beams). For ACE Isotopes, a high-current, 70 MeV cyclotron will be acquired to take radionuclide production off the existing Separated Sector Cyclotron (SSC). A freed up SSC will then be available for an increased tempo of nuclear physics research and to serve as a driver accelerator for the ACE Beams project, in which protons will be used for the direct fission of Uranium, producing beams of fission fragments. The ACE Beams project has begun with "LeRIB" - a Low Energy RIB facility, now under construction. In a collaboration with INFN Legnaro, the target/ion-source "front-end" will be a copy of the front-end developed for the SPES project. A variety of targets may be inserted into the SPES front-end; a uranium-carbide target has been designed to produce up to 2 × 1013 fission/s using a 70 MeV proton beam of 150 µA intensity.

Multi-channel probes to understand fission dynamics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mosby, Shea Morgan

2016-04-15

Explaining the origin of the elements is a major outstanding question in nuclear astrophysics. Observed elemental abundance distribution shows strong nuclear physics effects. In conclusion, neutron-induced reactions are important for nuclear astrophysics and applied fields in nuclear energy and security. LANSCE has a program to address many of these questions directly with neutron beams on (near-)stable nuclei. Increasing demand for correlated data to test details of fission models poses additional challenges. Possibilities exist to extend existing experimental efforts to radioactive beam facilities. Kinematic focusing from using inverse kinematics has potential to circumvent some challenges associated with measuring correlations between fissionmore » output channels.« less

Maximum reasonable radioxenon releases from medical isotope production facilities and their effect on monitoring nuclear explosions.

PubMed

Bowyer, Theodore W; Kephart, Rosara; Eslinger, Paul W; Friese, Judah I; Miley, Harry S; Saey, Paul R J

2013-01-01

Fission gases such as (133)Xe are used extensively for monitoring the world for signs of nuclear testing in systems such as the International Monitoring System (IMS). These gases are also produced by nuclear reactors and by fission production of (99)Mo for medical use. Recently, medical isotope production facilities have been identified as the major contributor to the background of radioactive xenon isotopes (radioxenon) in the atmosphere (Stocki et al., 2005; Saey, 2009). These releases pose a potential future problem for monitoring nuclear explosions if not addressed. As a starting point, a maximum acceptable daily xenon emission rate was calculated, that is both scientifically defendable as not adversely affecting the IMS, but also consistent with what is possible to achieve in an operational environment. This study concludes that an emission of 5 × 10(9) Bq/day from a medical isotope production facility would be both an acceptable upper limit from the perspective of minimal impact to monitoring stations, but also appears to be an achievable limit for large isotope producers. Copyright © 2012 Elsevier Ltd. All rights reserved.

Fission-gas release from uranium nitride at high fission rate density

NASA Technical Reports Server (NTRS)

Weinstein, M. B.; Kirchgessner, T. A.; Tambling, T. N.

1973-01-01

A sweep gas facility has been used to measure the release rates of radioactive fission gases from small UN specimens irradiated to 8-percent burnup at high fission-rate densities. The measured release rates have been correlated with an equation whose terms correspond to direct recoil release, fission-enhanced diffusion, and atomic diffusion (a function of temperature). Release rates were found to increase linearly with burnups between 1.5 and 8 percent. Pore migration was observed after operation at 1550 K to over 6 percent burnup.

Neutron-induced fission cross section measurements for uranium isotopes 236U and 234U at LANSCE

NASA Astrophysics Data System (ADS)

Laptev, A. B.; Tovesson, F.; Hill, T. S.

2013-04-01

A well established program of neutron-induced fission cross section measurement at Los Alamos Neutron Science Center (LANSCE) is supporting the Fuel Cycle Research program (FC R&D). The incident neutron energy range spans from sub-thermal up to 200 MeV by combining two LANSCE facilities, the Lujan Center and the Weapons Neutron Research facility (WNR). The time-of-flight method is implemented to measure the incident neutron energy. A parallel-plate fission ionization chamber was used as a fission fragment detector. The event rate ratio between the investigated foil and a standard 235U foil is converted into a fission cross section ratio. In addition to previously measured data new measurements include 236U data which is being analyzed, and 234U data acquired in the 2011-2012 LANSCE run cycle. The new data complete the full suite of Uranium isotopes which were investigated with this experimental approach. Obtained data are presented in comparison with existing evaluations and previous data.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William J. Jr.; Moran, Robert P.; Pearson, J. Boise

2012-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities

76 FR 16004 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-22

... to the integrity of a fission product barrier, nor are they relied upon as a primary success path to... of safety is related to confidence in the ability of the fission product barriers (fuel cladding... switch is in the Refuel position have no impact on the performance of the fission product barriers since...

Fission Surface Power Technology Development Testing at NASA's Early Flight Fission Test Facility

NASA Technical Reports Server (NTRS)

Houts. Michael G.

2009-01-01

Fission surface power (FSP) systems could be used to provide power anytime, anywhere on the surface of the Moon or Mars. FSP systems could be used at polar locations, at locations away from the poles, or in permanently shaded regions, with excellent performance at all sites. A potential reference 40 kWe option has been devised that is cost-competitive with alternatives while providing more power for less mass anywhere on the lunar surface. The reference FSP system (FSPS) is also readily extensible for use on Mars. At Mars the system would be capable of operating through global dust storms and providing year-round power at any Martian latitude. Under the NASA Exploration Technology Development Program (ETDP), NASA and the Department of Energy (DOE) have begun technology development on Fission Surface Power (FSP). The primary customer for this technology is the NASA Constellation Program which is responsible for the development of surface systems to support human exploration on the moon and Mars. The objectives of the FSP technology project are: 1) Develop FSP concepts that meet expected surface power requirements at reasonable cost with added benefits over other options. 2) Establish a hardware-based technical foundation for FSP design concepts and reduce overall development risk. 3) Reduce the cost uncertainties for FSP and establish greater credibility for flight system cost estimates. 4) Generate the key products to allow Agency decision-makers to consider FSP as a viable option for flight development. To be mass efficient, FSP systems must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial systems. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference FSP system uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance on the surface of the moon or Mars. Recent testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference FSP system, and has helped evaluate methods for system integration. In June, 2009, a representative pumped NaK loop (provided by Marshall Space Flight Center) was coupled to a Stirling power converter (provided by Glenn Research Center) and tested at various conditions representative of those that would be seen during actual FSP system operation. In all areas, performance of the integrated system exceeded project goals. High-temperature NaK pump testing has also been performed at the EFF-TF, as has testing of methods for providing long-duration NaK purity.

Irradiation data for the MFA-1 and MFA-2 tests in the FFTF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nelson, J.V.

This report provides key information on the irradiation environment of the MONJU fuel tests MFA-1 and MFA-2 in the Fast Flux Test Facility (FFTF). This information includes the fission powers, neutron fluxes, sodium temperatures and sodium flow rates in MFA-I, MFA-2 and adjacent assemblies. It also includes MFA-1 and MFA-2 compositions as a function of exposure. The work was performed at the request of Power Reactor and Nuclear Fuels Corporation (PNC) of Japan.

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Code of Federal Regulations, 2010 CFR

2010-01-01

... transferred to a Federal repository no later than 10 years following separation of fission products from the.... Disposal of high-level radioactive fission product waste material will not be permitted on any land other... of the policy stated above with respect to high-level radioactive fission product wastes generated...

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Code of Federal Regulations, 2011 CFR

2011-01-01

... transferred to a Federal repository no later than 10 years following separation of fission products from the.... Disposal of high-level radioactive fission product waste material will not be permitted on any land other... of the policy stated above with respect to high-level radioactive fission product wastes generated...

Reducing Uncertainties in Neutron-Induced Fission Cross Sections Using a Time Projection Chamber

NASA Astrophysics Data System (ADS)

Manning, Brett; Niffte Collaboration

2015-10-01

Neutron-induced fission cross sections for actinides have long been of great interest for nuclear energy and stockpile stewardship. Traditionally, measurements were performed using fission chambers which provided limited information about the detected fission events. For the case of 239Pu(n,f), sensitivity studies have shown a need for more precise measurements. Recently the Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) has developed the fission Time Projection Chamber (fissionTPC) to measure fission cross sections to better than 1% uncertainty by providing 3D tracking of fission fragments. The fissionTPC collected data to calculate the 239Pu(n,f) cross section at the Weapons Neutron Research facility at the Los Alamos Neutron Science Center during the 2014 run cycle. Preliminary analysis has been focused on studying particle identification and target and beam non-uniformities to reduce the uncertainty on the cross section. Additionally, the collaboration is investigating other systematic errors that could not be well studied with a traditional fission chamber. LA-UR-15-24906.

The Neutrons for Science Facility at SPIRAL-2

NASA Astrophysics Data System (ADS)

Ledoux, X.; Aïche, M.; Avrigeanu, M.; Avrigeanu, V.; Audouin, L.; Balanzat, E.; Ban-détat, B.; Ban, G.; Barreau, G.; Bauge, E.; Bélier, G.; Bem, P.; Blideanu, V.; Borcea, C.; Bouffard, S.; Caillaud, T.; Chatillon, A.; Czajkowski, S.; Dessagne, P.; Doré, D.; Fallot, M.; Farget, F.; Fischer, U.; Giot, L.; Granier, T.; Guillous, S.; Gunsing, F.; Gustavsson, C.; Jacquot, B.; Jansson, K.; Jurado, B.; Kerveno, M.; Klix, A.; Landoas, O.; Lecolley, F. R.; Lecouey, J. L.; Majerle, M.; Marie, N.; Materna, T.; Mrazek, J.; Negoita, F.; Novak, J.; Oberstedt, S.; Oberstedt, A.; Panebianco, S.; Perrot, L.; Plompen, A. J. M.; Pomp, S.; Ramillon, J. M.; Ridikas, D.; Rossé, B.; Rudolf, G.; Serot, O.; Simakov, S. P.; Simeckova, E.; Smith, A. G.; Sublet, J. C.; Taieb, J.; Tassan-Got, L.; Tarrio, D.; Takibayev, A.; Thfoin, I.; Tsekhanovich, I.; Varignon, C.

2014-05-01

The Neutrons For Science (NFS) facility is a component of SPIRAL-2 laboratory under construction at Caen (France). SPIRAL-2 is dedicated to the production of high intensity Radioactive Ions Beams (RIB). It is based on a high-power linear accelerator (LINAG) to accelerate deuterons beams in order to produce neutrons by breakup reactions on a C converter. These neutrons will induce fission in 238U for production of radioactive isotopes. Additionally to the RIB production, the proton and deuteron beams delivered by the accelerator will be used in the NFS facility. NFS is composed of a pulsed neutron beam and irradiation stations for cross-section measurements and material studies. The beams delivered by the LINAG will allow producing intense neutron beams in the 100 keV-40 MeV energy range with either a continuous or quasi-mono-energetic spectrum. At NFS available average fluxes will be up to 2 orders of magnitude higher than those of other existing time-of-flight facilities in the 1 MeV - 40 MeV range. NFS will be a very powerful tool for fundamental physics and application related research in support of the transmutation of nuclear waste, design of future fission and fusion reactors, nuclear medicine or test and development of new detectors. The facility and its characteristics are described, and several examples of the first potential experiments are presented.

Enhanced trigger for the NIFFTE fissionTPC in presence of high-rate alpha backgrounds

NASA Astrophysics Data System (ADS)

Bundgaard, Jeremy; Niffte Collaboration

2015-10-01

Nuclear physics and nuclear energy communities call for new, high precision measurements to improve existing fission models and design next generation reactors. The Neutron Induced Fission Fragment Tracking experiment (NIFFTE) has developed the fission Time Projection Chamber (fissionTPC) to measure neutron induced fission with unrivaled precision. The fissionTPC is annually deployed to the Weapons Neutron Research facility at Los Alamos Neutron Science Center where it operates with a neutron beam passing axially through the drift volume, irradiating heavy actinide targets to induce fission. The fissionTPC was developed at the Lawrence Livermore National Laboratory's TPC lab, where it measures spontaneous fission from radioactive sources to characterize detector response, improve performance, and evolve the design. To measure 244Cm, we've developed a fission trigger to reduce the data rate from alpha tracks while maintaining a high fission detection efficiency. In beam, alphas from 239Pu are a large background when detecting fission fragments; implementing the fission trigger will greatly reduce this background. The implementation of the cathode fission trigger in the fissionTPC will be presented along with a detailed study of its efficiency.

An off-line method to characterize the fission product release from uranium carbide-target prototypes developed for SPIRAL2 project

NASA Astrophysics Data System (ADS)

Hy, B.; Barré-Boscher, N.; Özgümüs, A.; Roussière, B.; Tusseau-Nenez, S.; Lau, C.; Cheikh Mhamed, M.; Raynaud, M.; Said, A.; Kolos, K.; Cottereau, E.; Essabaa, S.; Tougait, O.; Pasturel, M.

2012-10-01

In the context of radioactive ion beams, fission targets, often based on uranium compounds, have been used for more than 50 years at isotope separator on line facilities. The development of several projects of second generation facilities aiming at intensities two or three orders of magnitude higher than today puts an emphasis on the properties of the uranium fission targets. A study, driven by Institut de Physique Nucléaire d'Orsay (IPNO), has been started within the SPIRAL2 project to try and fully understand the behavior of these targets. In this paper, we have focused on five uranium carbide based targets. We present an off-line method to characterize their fission product release and the results are examined in conjunction with physical characteristics of each material such as the microstructure, the porosity and the chemical composition.

Fission Fragment characterization with FALSTAFF at NFS

NASA Astrophysics Data System (ADS)

Doré, D.; Farget, F.; Lecolley, F.-R.; Ledoux, X.; Lehaut, G.; Materna, T.; Pancin, J.; Panebianco, S.

2013-03-01

The Neutrons for Science (NFS) facility will be one of the first installations of the SPIRAL2 facility. NFS will be composed of a time-of-flight baseline and irradiation stations and will allow studying neutron-induced reactions for energies going from some hundreds of keV up to 40 MeV. Continuous and quasi-monoenergetic energy neutron beams will be available. Taking advantage of this new installation, the development of an experimental setup for a full characterization of actinide fission fragments in this energy domain has been undertaken. To achieve this goal a new detection system called FALSTAFF (Four Arm cLover for the STudy of Actinide Fission Fragments) in under development. In this paper, the characteristics of the NFS facility will be exposed and the motivations for the FALSTAFF experiment will be presented. The experimental setup will be described and the expected resolutions based on realistic GEANT4 simulations will be discussed.

Development Status of the Fission Power System Technology Demonstration Unit

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Gibson, Marc A.; Geng, Steven M.; Pearson, Jon Boise; Godfoy, Thomas

2012-01-01

This paper summarizes the progress that has been made in the development of the Fission Power System Technology Demonstration Unit (TDU). The reactor simulator core and Annular Linear Induction Pump have been fabricated and assembled into a test loop at the NASA Marshall Space Flight Center. A 12 kWe Power Conversion Unit (PCU) is being developed consisting of two 6 kWe free-piston Stirling engines. The two 6 kWe engines have been fabricated by Sunpower Inc. and are currently being tested separately prior to integration into the PCU. The Facility Cooling System (FCS) used to reject convertor waste heat has been assembled and tested at the NASA Glenn Research Center (GRC). The structural elements, including a Buildup Assembly Platform (BAP) and Upper Truss Structure (UTS) have been fabricated, and will be used to test cold-end components in thermal vacuum prior to TDU testing. Once all components have been fully tested at the subsystem level, they will be assembled into an end-to-end system and tested in thermal vacuum at GRC.

Development Status of the Fission Power System Technology Demonstration Unit

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Gibson, Marc A.; Geng, Steven M; Pearson, Jon Boise; Godfroy, Thomas

2012-01-01

This paper summarizes the progress that has been made in the development of the Fission Power System Technology Demonstration Unit (TDU). The reactor simulator core and Annular Linear Induction Pump have been fabricated and assembled into a test loop at the NASA Marshall Space Flight Center. A 12 kWe Power Conversion Unit (PCU) is being developed consisting of two 6 kWe free-piston Stirling engines. The two 6 kWe engines have been fabricated by Sunpower Inc. and are currently being tested separately prior to integration into the PCU. The Facility Cooling System (FCS) used to reject convertor waste heat has been assembled and tested at the NASA Glenn Research Center (GRC). The structural elements, including a Buildup Assembly Platform (BAP) and Upper Truss Structure (UTS) have been fabricated, and will be used to test cold-end components in thermal vacuum prior to TDU testing. Once all components have been fully tested at the subsystem level, they will be assembled into an end-to-end system and tested in thermal vacuum at NASA GRC.

Fission neutron source in Rome

NASA Astrophysics Data System (ADS)

Coppola, Mario; Di Majo, V.; Ingrao, G.; Rebessi, S.; Testa, A.

1997-02-01

A fission neutron source is operating in Rome at the ENEA Casaccia Research Center since 1971, consisting of a low power fast reactor named RSV-Tapiro. it is employed for a variety of experiments, including dosimetry, material testing, radiation protection and biology. In particular, application to experimental radiobiology includes studies of the biological action of neutrons in the whole-body irradiated animal, or in specialized systems in vivo or in vitro. For his purpose a vertical irradiation facility was originally constructed. Recently, a new horizontal irradiation facility has been designed to allow the exposure of larger samples or larger sample batches at one time. Dosimetry at the sample irradiation positions is routinely carried out by the conventional method of using two ion chambers. This physical dosimetry has recently been compared with the results of biological dosimetry based on the detection of chromosomal aberrations in peripheral blood human lymphocytes irradiated in vitro. A characterization of the radiation quality in the two configurations has been carried out by tissue equivalent proportional counter microdosimetry measurements. Information about the main characteristics of the reactor and the two irradiation facilities is provided and relevant results of the various measurements are summarized. Radiobiological results obtained using this neutron source are also briefly outlined.

Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jason M. Harp; Paul A. Demkowicz

2014-10-01

In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10 -4 to 10 -5) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materialsmore » is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.« less

Nuclear data measurements at the new NFS facility at GANIL

NASA Astrophysics Data System (ADS)

Gustavsson, C.; Pomp, S.; Scian, G.; Lecolley, F.-R.; Tippawan, U.; Watanabe, Y.

2012-10-01

The NFS (Neutrons For Science) facility is part of the SPRIAL 2 project at GANIL, Caen, France. The facility is currently under construction and the first beam is expected in early 2013. NFS will have a white neutron source covering the 1-40 MeV energy range with a neutron flux higher than comparable facilities. A quasi-mono-energetic neutron beam will also be available. In these energy ranges, especially above 14 MeV, there is a large demand for neutron-induced data for a wide range of applications involving dosimetry, medical therapy, single-event upsets in electronics and nuclear energy. Today, there are a few or no cross section data on reactions such as (n, fission), (n, xn), (n, p), (n, d) and (n, α). We propose to install experimental equipment for measuring neutron-induced light-charged particle production and fission relative to the H(n, p) cross section. Both the H(n, p) cross section and the fission cross section for 238U are important reference cross sections used as standards for many other experiments. Nuclear data for certain key elements, such as closed shell nuclei, are also of relevance for the development of nuclear reaction models. Our primary intention is to measure charged particle production (protons, deuterons and alphas) from 12C, 16O, 28Si and 56Fe and neutron-induced fission cross sections from 238U and 232Th.

Thermal Characterization of a Simulated Fission Engine via Distributed Fiber Bragg Gratings

NASA Astrophysics Data System (ADS)

Duncan, Roger G.; Fielder, Robert S.; Seeley, Ryan J.; Kozikowski, Carrie L.; Raum, Matthew T.

2005-02-01

We report the use of distributed fiber Bragg gratings to monitor thermal conditions within a simulated nuclear reactor core located at the Early Flight Fission Test Facility of the NASA Marshall Space Flight Center. Distributed fiber-optic temperature measurements promise to add significant capability and advance the state-of-the-art in high-temperature sensing. For the work reported herein, seven probes were constructed with ten sensors each for a total of 70 sensor locations throughout the core. These discrete temperature sensors were monitored over a nine hour period while the test article was heated to over 700 °C and cooled to ambient through two operational cycles. The sensor density available permits a significantly elevated understanding of thermal effects within the simulated reactor. Fiber-optic sensor performance is shown to compare very favorably with co-located thermocouples where such co-location was feasible.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Phase II Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William J.; Moran, Robert P.; Pearson, J. Bose

2013-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities. Keywords: Nuclear Thermal Propulsion, Simulator

Status of DEMO-FNS development

NASA Astrophysics Data System (ADS)

Kuteev, B. V.; Shpanskiy, Yu. S.; DEMO-FNS Team

2017-07-01

Fusion-fission hybrid facility based on superconducting tokamak DEMO-FNS is developed in Russia for integrated commissioning of steady-state and nuclear fusion technologies at the power level up to 40 MW for fusion and 400 MW for fission reactions. The project status corresponds to the transition from a conceptual design to an engineering one. This facility is considered, in RF, as the main source of technological and nuclear science information, which should complement the ITER research results in the fields of burning plasma physics and control.

A Feasibility Study on Reactor Based Fission Neutron Radiography of 200-l Waste Packages

NASA Astrophysics Data System (ADS)

Bücherl, T.; Kalthoff, O.; von Gostomski, Ch. Lierse

This feasibility study investigates the applicability of fission neutrons for the non-destructive characterization of radioactive waste packages by means of neutron radiography. Based on a number of mock-up drums of different non-radioactive matrices, but being typical for radioactive waste generated in Europe, radiography measurements at the NECTAR and the ITS facility using fission neutrons and 60Co-gamma-rays, respectively, are performed. The resulting radiographs are compared and qualitatively assessed. In addition, a first approach for the stitching of the fission neutron radiographs to visualize the complete area of 200-l waste drums is performed. While the feasibility of fission neutrons is demonstrated successfully, fields for further improvements are identified.

A fission gas release correlation for uranium nitride fuel pins

NASA Technical Reports Server (NTRS)

Weinstein, M. B.; Davison, H. W.

1973-01-01

A model was developed to predict fission gas releases from UN fuel pins clad with various materials. The model was correlated with total release data obtained by different experimentors, over a range of fuel temperatures primarily between 1250 and 1660 K, and fuel burnups up to 4.6 percent. In the model, fission gas is transported by diffusion mechanisms to the grain boundaries where the volume grows and eventually interconnects with the outside surface of the fuel. The within grain diffusion coefficients are found from fission gas release rate data obtained using a sweep gas facility.

Studies of fission fragment properties at the Los Alamos Neutron Science Center (LANSCE)

NASA Astrophysics Data System (ADS)

Tovesson, Fredrik; Mayorov, Dmitriy; Duke, Dana; Manning, Brett; Geppert-Kleinrath, Verena

2017-09-01

Nuclear data related to the fission process are needed for a wide variety of research areas, including fundamental science, nuclear energy and non-proliferation. While some of the relevant data have been measured to the required accuracies there are still many aspects of fission that need further investigation. One such aspect is how Total Kinetic Energy (TKE), fragment yields, angular distributions and other fission observables depend on excitation energy of the fissioning system. Another question is the correlation between mass, charge and energy of fission fragments. At the Los Alamos Neutron Science Center (LANSCE) we are studying neutron-induced fission at incident energies from thermal up to hundreds of MeV using the Lujan Center and Weapons Neutron Research (WNR) facilities. Advanced instruments such as SPIDER (time-of-flight and kinetic energy spectrometer), the NIFFTE Time Projection Chamber (TPC), and Frisch grid Ionization Chambers (FGIC) are used to investigate the properties of fission fragments, and some important results for the major actinides have been obtained.

Studies of fission fragment properties at the Los Alamos Neutron Science Center (LANSCE)

DOE PAGES

Tovesson, Fredrik; Mayorov, Dmitriy; Duke, Dana; ...

2017-09-13

Nuclear data related to the fission process are needed for a wide variety of research areas, including fundamental science, nuclear energy and non-proliferation. While some of the relevant data have been measured to the required accuracies there are still many aspects of fission that need further investigation. One such aspect is how Total Kinetic Energy (TKE), fragment yields, angular distributions and other fission observables depend on excitation energy of the fissioning system. Another question is the correlation between mass, charge and energy of fission fragments. At the Los Alamos Neutron Science Center (LANSCE) we are studying neutron-induced fission at incidentmore » energies from thermal up to hundreds of MeV using the Lujan Center and Weapons Neutron Research (WNR) facilities. Advanced instruments such as SPIDER (time-of-flight and kinetic energy spectrometer), the NIFFTE Time Projection Chamber (TPC), and Frisch grid Ionization Chambers (FGIC) are used to investigate the properties of fission fragments, and some important results for the major actinides have been obtained.« less

Studies of fission fragment properties at the Los Alamos Neutron Science Center (LANSCE)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tovesson, Fredrik; Mayorov, Dmitriy; Duke, Dana

Nuclear data related to the fission process are needed for a wide variety of research areas, including fundamental science, nuclear energy and non-proliferation. While some of the relevant data have been measured to the required accuracies there are still many aspects of fission that need further investigation. One such aspect is how Total Kinetic Energy (TKE), fragment yields, angular distributions and other fission observables depend on excitation energy of the fissioning system. Another question is the correlation between mass, charge and energy of fission fragments. At the Los Alamos Neutron Science Center (LANSCE) we are studying neutron-induced fission at incidentmore » energies from thermal up to hundreds of MeV using the Lujan Center and Weapons Neutron Research (WNR) facilities. Advanced instruments such as SPIDER (time-of-flight and kinetic energy spectrometer), the NIFFTE Time Projection Chamber (TPC), and Frisch grid Ionization Chambers (FGIC) are used to investigate the properties of fission fragments, and some important results for the major actinides have been obtained.« less

Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Pearson, J. Boise; Godfroy, Thomas J.

2011-01-01

Safe, reliable, compact, autonomous, long-life fission systems have numerous potential applications, both terrestrially and in space. Technologies and facilities developed in support of these systems could be useful to a variety of concepts. At moderate power levels, fission systems can be designed to operate for decades without the need for refueling. In addition, fast neutron damage to cladding and structural materials can be maintained at an acceptable level. Nuclear design codes have advanced to the stage where high confidence in the behavior and performance of a system can be achieved prior to initial testing. To help ensure reactor affordability, an optimal strategy must be devised for development and qualification. That strategy typically involves a combination of non-nuclear and nuclear testing. Non-nuclear testing is particularly useful for concepts in which nuclear operating characteristics are well understood and nuclear effects such as burnup and radiation damage are not likely to be significant. To be mass efficient, a SFPS must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial reactors. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference SFPS uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance. In addition, technologies from the SFPS system could be applicable to compact terrestrial systems. Recent non-nuclear testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference SFPS and evaluate methods for system integration. In July, 2011 an Annular Linear Induction Pump (ALIP) provided by Idaho National Laboratory was tested at the EFF-TF to assess performance and verify suitability for use in a10 kWe technology demonstration unit (TDU). In November, 2011 testing of a 37-pin core simulator (designed in conjunction with Los Alamos National Laboratory) for use with the TDU will occur. Previous testing at the EFFTF has included the thermal and mechanical coupling of a pumped NaK loop to Stirling engines (provided by GRC). Testing related to heat pipe cooled systems, gas cooled systems, heat exchangers, and other technologies has also been performed. Integrated TDU testing will begin at GRC in 2013. Thermal simulators developed at the EFF-TF are capable of operating over the temperature and power range typically of interest to compact reactors. Small and large diameter simulators have been developed, and simulators (coupled with the facility) are able to closely match the axial and radial power profile of all potential systems of interest. A photograph of the TDU core simulator during assembly is provided in Figure 2.

Prompt fission neutron emission in the reaction 235U(n,f)

NASA Astrophysics Data System (ADS)

Göök, Alf; Hambsch, Franz-Josef; Oberstedt, Stephan

2018-03-01

Experimental activities at JRC-Geel on prompt fission neutron (PFN) emission in response to OECD/NEA nuclear data requests are presented in this contribution. Specifically, on-going investigations of PFN emission from the reaction 235U(n,f) in the region of the resolved resonances, taking place at the GELINA facility, are presented. The focus of this contribution lies on studies of PFN correlations with fission fragment properties. The experiment employs a scintillation detector array for neutron detection, while fission fragment properties are determined via the double kinetic energy technique using a position sensitive twin ionization chamber. This setup allows us to study several correlations between properties of neutron and fission fragments simultaneously. Results on PFN correlations with fission fragment properties from the present study differ significantly from earlier studies on this reaction, induced by thermal neutrons.

Two detector arrays for fast neutrons at LANSCE

NASA Astrophysics Data System (ADS)

Haight, R. C.; Lee, H. Y.; Taddeucci, T. N.; O'Donnell, J. M.; Perdue, B. A.; Fotiades, N.; Devlin, M.; Ullmann, J. L.; Laptev, A.; Bredeweg, T.; Jandel, M.; Nelson, R. O.; Wender, S. A.; White, M. C.; Wu, C. Y.; Kwan, E.; Chyzh, A.; Henderson, R.; Gostic, J.

2012-03-01

The neutron spectrum from neutron-induced fission needs to be known in designing new fast reactors, predicting criticality for safety analyses, and developing techniques for global security application. The experimental data base of fission neutron spectra is very incomplete and most present evaluated libraries are based on the approach of the Los Alamos Model. To validate these models and to provide improved data for applications, a program is underway to measure the fission neutron spectrum for a wide range of incident neutron energies using the spallation source of fast neutrons at the Weapons Neutron Research (WNR) facility at the Los Alamos Neutron Science Center (LANSCE). In a double time-of-flight experiment, fission neutrons are detected by arrays of neutron detectors to increase the solid angle and also to investigate possible angular dependence of the fission neutrons. The challenge is to measure the spectrum from low energies, down to 100 keV or so, to energies over 10 MeV, where the evaporation-like spectrum decreases by 3 orders of magnitude from its peak around 1 MeV. For these measurements, we are developing two arrays of neutron detectors, one based on liquid organic scintillators and the other on 6Li-glass detectors. The range of fission neutrons detected by organic liquid scintillators extends from about 600 keV to well over 10 MeV, with the lower limit being defined by the limit of pulse-shape discrimination. The 6Li-glass detectors have a range from very low energies to about 1 MeV, where their efficiency then becomes small. Various considerations and tests are in progress to understand important contributing factors in designing these two arrays and they include selection and characterization of photomultiplier tubes (PM), the performance of relatively thin (1.8 cm) 6Li-glass scintillators on 12.5 cm diameter PM tubes, use of 17.5 cm diameter liquid scintillators with 12.5 cm PM tubes, measurements of detector efficiencies with tagged neutrons from the WNR/LANSCE neutron beam, and efficiency calibration with 252Cf spontaneous fission neutrons. Design considerations and test results are presented.

Cold-end Subsystem Testing for the Fission Power System Technology Demonstration Unit

NASA Technical Reports Server (NTRS)

Briggs, Maxwell; Gibson, Marc; Ellis, David; Sanzi, James

2013-01-01

The Fission Power System (FPS) Technology Demonstration Unit (TDU) consists of a pumped sodium-potassium (NaK) loop that provides heat to a Stirling Power Conversion Unit (PCU), which converts some of that heat into electricity and rejects the waste heat to a pumped water loop. Each of the TDU subsystems is being tested independently prior to full system testing at the NASA Glenn Research Center. The pumped NaK loop is being tested at NASA Marshall Space Flight Center; the Stirling PCU and electrical controller are being tested by Sunpower Inc.; and the pumped water loop is being tested at Glenn. This paper describes cold-end subsystem setup and testing at Glenn. The TDU cold end has been assembled in Vacuum Facility 6 (VF 6) at Glenn, the same chamber that will be used for TDU testing. Cold-end testing in VF 6 will demonstrate functionality; validated cold-end fill, drain, and emergency backup systems; and generated pump performance and system pressure drop data used to validate models. In addition, a low-cost proof-of concept radiator has been built and tested at Glenn, validating the design and demonstrating the feasibility of using low-cost metal radiators as an alternative to high-cost composite radiators in an end-to-end TDU test.

Cold-End Subsystem Testing for the Fission Power System Technology Demonstration Unit

NASA Technical Reports Server (NTRS)

Briggs, Mazwell; Gibson, Marc; Ellis, David; Sanzi, James

2013-01-01

The Fission Power System (FPS) Technology Demonstration Unit (TDU) consists of a pumped sodiumpotassium (NaK) loop that provides heat to a Stirling Power Conversion Unit (PCU), which converts some of that heat into electricity and rejects the waste heat to a pumped water loop. Each of the TDU subsystems is being tested independently prior to full system testing at the NASA Glenn Research Center. The pumped NaK loop is being tested at NASA Marshall Space Flight Center; the Stirling PCU and electrical controller are being tested by Sunpower Inc.; and the pumped water loop is being tested at Glenn. This paper describes cold-end subsystem setup and testing at Glenn. The TDU cold end has been assembled in Vacuum Facility 6 (VF 6) at Glenn, the same chamber that will be used for TDU testing. Cold-end testing in VF 6 will demonstrate functionality; validated coldend fill, drain, and emergency backup systems; and generated pump performance and system pressure drop data used to validate models. In addition, a low-cost proof-of concept radiator has been built and tested at Glenn, validating the design and demonstrating the feasibility of using low-cost metal radiators as an alternative to highcost composite radiators in an end-to-end TDU test.

Measurements of fission product yield in the neutron-induced fission of 238U with average energies of 9.35 MeV and 12.52 MeV

NASA Astrophysics Data System (ADS)

Mukerji, Sadhana; Krishnani, Pritam Das; Shivashankar, Byrapura Siddaramaiah; Mulik, Vikas Kaluram; Suryanarayana, Saraswatula Venkat; Naik, Haladhara; Goswami, Ashok

2014-07-01

The yields of various fission products in the neutron-induced fission of 238U with the flux-weightedaveraged neutron energies of 9.35 MeV and 12.52 MeV were determined by using an off-line gammaray spectroscopic technique. The neutrons were generated using the 7Li(p, n) reaction at Bhabha Atomic Research Centre-Tata Institute of Fundamental Research Pelletron facility, Mumbai. The gamma- ray activities of the fission products were counted in a highly-shielded HPGe detector over a period of several weeks to identify the decaying fission products. At both the neutron energies, the fission-yield values are reported for twelve fission product. The results obtained from the present work have been compared with the similar data for mono-energetic neutrons of comparable energy from the literature and are found to be in good agreement. The peak-to-valley (P/V) ratios were calculated from the fission-yield data and were found to decreases for neutron energy from 9.35 to 12.52 MeV, which indicates the role of excitation energy. The effect of the nuclear structure on the fission product-yield is discussed.

Determination of the fast-neutron-induced fission cross-section of 242Pu at nELBE

NASA Astrophysics Data System (ADS)

Kögler, Toni; Beyer, Roland; Junghans, Arnd R.; Schwengner, Ronald; Wagner, Andreas

2018-03-01

The fast-neutron-induced fission cross section of 242Pu was determined in the energy range of 0.5 MeV to 10MeV at the neutron time-of-flight facility nELBE. Using a parallel-plate fission ionization chamber this quantity was measured relative to 235U(n,f). The number of target nuclei was thereby calculated by means of measuring the spontaneous fission rate of 242Pu. An MCNP 6 neutron transport simulation was used to correct the relative cross section for neutron scattering. The determined results are in good agreement with current experimental and evaluated data sets.

Modeling fission product vapor transport in the Falcon facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shepherd, I.M.; Drossinos, Y.; Benson, C.G.

1995-05-01

An extensive database of aerosol Experiments exists and has been used for checking aerosol transport codes. Data for fission product vapor transport are harder to find. Some qualitative data are available, but the Falcon thermal gradient tube tests carried out at AEA Technology`s laboratories in Winfrith, England, mark the first serious attempt to provide a set of experiments suitable for the validation of codes that predict the transport and condensation of realistic mixtures of fission product vapors. Four of these have been analyzed to check how well the computer code VICTORIA can predict the most important phenomena. Of the fourmore » experiments studied, two are reference cases (FAL-17 and FAL-19), one is a case without boric acid (FAL-18), and the other is run in a reducing atmosphere (FAL-20). The results show that once the vapors condense onto aerosols, VICTORIA can predict their deposition rather well. The dominant mechanism is thermophoresis, and each element deposits with more or less the same deposition velocity. The behavior of the vapors is harder to interpret. Essentially, it is important to know the temperature at which each element condenses. It is clear from the measurements that this temperature changed from test to test-caused mostly by the different speciation as the composition of the carrier gas and the relative concentration of other fission products changed. Only in the test with a steam atmosphere and without boric acid was the assumption valid that most of the iodine is cesium iodide and most of the cesium is cesium hydroxide. In general, VICTORIA predicts that, with the exception of cesium, there will be less variation in the speciation-and, hence, variation in the deposition-between tests than is in fact observed. VICTORIA underpredicts the volatility of most elements, and this is partly a consequence of the ideal solution assumption and partly an overestimation of vapor/aerosol interactions.« less

Summary of the Manufacture, Testing and Model Validation of a Full-Scale Radiator for Fission Surface Power Applications

NASA Technical Reports Server (NTRS)

Ellis, David L.; Calder, James; Siamidis, John

2011-01-01

A full-scale radiator for a lunar fission surface power application was manufactured by Material innovations, Inc., for the NASA Glenn Research Center. The radiator was designed to reject 6 kWt with an inlet water temperature of 400 K and a water mass flow rate of 0.5 kg/s. While not flight hardware, the radiator incorporated many potential design features and manufacturing techniques for future flight hardware. The radiator was tested at NASA Glenn Research Center for heat rejection performance. The results showed that the radiator design was capable of rejecting over 6 kWt when operating at the design conditions. The actual performance of the radiator as a function of operational manifolds, inlet water temperature and facility sink temperature was compared to the predictive model developed by NASA Glenn Research Center. The results showed excellent agreement with the model with the actual average face sheet temperature being within 1% of the predicted value. The results will be used in the design and production of NASA s next generation fission power heat rejection systems. The NASA Glenn Research Center s Technology Demonstration Unit will be the first project to take advantage of the newly developed manufacturing techniques and analytical models.

Spent fuel treatment and mineral waste form development at Argonne National Laboratory-West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goff, K.M.; Benedict, R.W.; Bateman, K.

1996-07-01

At Argonne National Laboratory-West (ANL-West) there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Conditioning Facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. Both mineral and metal high-level waste forms will be produced. The mineral waste form will contain the active metal fission products and the transuranics. Cold small-scale waste form testing has been on-going at Argonne in Illinois. Large-scale testing is commencing at ANL-West.

Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test

NASA Astrophysics Data System (ADS)

Godfroy, Thomas J.; Kapernick, Richard J.; Bragg-Sitton, Shannon M.

2004-02-01

One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

First Generation Least Expensive Approach to Fission (FiGLEAF) Testing Results

NASA Technical Reports Server (NTRS)

VanDyke, Melissa; Houts, Mike; Pedersen, Kevin; Godfroy, Tom; Dickens, Ricky; Poston, David; Reid, Bob; Salvail. Pat; Ring, Peter; Schmidt, George R. (Technical Monitor)

2000-01-01

Successful development of space fission systems will require an extensive program of affordable and realistic testing. In addition to tests related to design/development of the fission system, realistic testing of the actual flight unit must also be performed. Testing can be divided into two categories, non-nuclear tests and nuclear tests. Full power nuclear tests of space fission systems are expensive, time consuming, and of limited use, even in the best of programmatic environments. If the system is designed to operate within established radiation damage and fuel burn up limits while simultaneously being designed to allow close simulation of heat from fission using resistance heaters, high confidence in fission system performance and lifetime can be attained through a series of non-nuclear tests. Non-nuclear tests are affordable and timely, and the cause of component and system failures can be quickly and accurately identified. MSFC is leading a Safe Affordable Fission Engine (SAFE) test series whose ultimate goal is the demonstration of a 300 kW flight configuration system using non-nuclear testing. This test series is carried out in collaboration with other NASA centers, other government agencies, industry, and universities. The paper describes the SAFE test series, which includes test article descriptions, test results and conclusions, and future test plans.

Feasibility study of a magnetic fusion production reactor

NASA Astrophysics Data System (ADS)

Moir, R. W.

1986-12-01

A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about 1.4 billion (1982 dollars) in either case. (The direct costs are estimated at 1.1 billion.) The production cost is calculated to be 22,000/g for tritium and 260/g for plutonium of quite high purity (1%240Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells electricity, and (2) there is a risk of not meeting the design goals.

Nuclear Structure Studies with Stable and Radioactive Beams: The SPES radioactive ion beam project

NASA Astrophysics Data System (ADS)

de Angelis, G.; SPES Collaboration; Prete, G.; Andrighetto, A.; Manzolaro, M.; Corradetti, S.; Scarpa, D.; Rossignoli, M.; Monetti, A.; Lollo, M.; Calderolla, M.; Vasquez, J.; Zafiropoulos, D.; Sarchiapone, L.; Benini, D.; Favaron, P.; Rigato, M.; Pegoraro, R.; Maniero, D.; Calabretta, L.; Comunian, M.; Maggiore, M.; Lombardi, A.; Piazza, L.; Porcellato, A. M.; Roncolato, C.; Bisoffi, G.; Pisent, A.; Galatà, A.; Giacchini, M.; Bassato, G.; Canella, S.; Gramegna, F.; Valiente, J.; Bermudez, J.; Mastinu, P. F.; Esposito, J.; Wyss, J.; Russo, A.; Zanella, S.

2015-04-01

A new Radioactive Ion Beam (RIB) facility (SPES) is presently under construction at the Legnaro National Laboratories of INFN. The SPES facility is based on the ISOL method using an UCx Direct Target able to sustain a power of 10 kW. The primary proton beam is provided by a high current Cyclotron accelerator with energy of 35-70 MeV and a beam current of 0.2-0.5 mA. Neutron-rich radioactive ions are produced by proton induced fission on an Uranium target at an expected fission rate of the order of 1013 fissions per second. After ionization and selection the exotic isotopes are re-accelerated by the ALPI superconducting LINAC at energies of 10A MeV for masses in the region A=130 amu. The expected secondary beam rates are of the order of 107 - 109 pps. Aim of the SPES facility is to deliver high intensity radioactive ion beams of neutron rich nuclei for nuclear physics research as well as to be an interdisciplinary research centre for radio-isotopes production for medicine and for neutron beams.

Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

NASA Astrophysics Data System (ADS)

Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

2017-09-01

Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

Experimental Fuels Facility Re-categorization Based on Facility Segmentation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reiss, Troy P.; Andrus, Jason

The Experimental Fuels Facility (EFF) (MFC-794) at the Materials and Fuels Complex (MFC) located on the Idaho National Laboratory (INL) Site was originally constructed to provide controlled-access, indoor storage for radiological contaminated equipment. Use of the facility was expanded to provide a controlled environment for repairing contaminated equipment and characterizing, repackaging, and treating waste. The EFF facility is also used for research and development services, including fuel fabrication. EFF was originally categorized as a LTHC-3 radiological facility based on facility operations and facility radiological inventories. Newly planned program activities identified the need to receive quantities of fissionable materials in excessmore » of the single parameter subcritical limit in ANSI/ANS-8.1, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors” (identified as “criticality list” quantities in DOE-STD-1027-92, “Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports,” Attachment 1, Table A.1). Since the proposed inventory of fissionable materials inside EFF may be greater than the single parameter sub-critical limit of 700 g of U-235 equivalent, the initial re-categorization is Hazard Category (HC) 2 based upon a potential criticality hazard. This paper details the facility hazard categorization performed for the EFF. The categorization was necessary to determine (a) the need for further safety analysis in accordance with LWP-10802, “INL Facility Categorization,” and (b) compliance with 10 Code of Federal Regulations (CFR) 830, Subpart B, “Safety Basis Requirements.” Based on the segmentation argument presented in this paper, the final hazard categorization for the facility is LTHC-3. Department of Energy Idaho (DOE-ID) approval of the final hazard categorization determined by this hazard assessment document (HAD) was required per the DOE-ID Supplemental Guidance for DOE-STD-1027-92 based on the proposed downgrade of the initial facility categorization of Hazard Category 2.« less

Characterization of the Medley setup for measurements of neutron-induced fission cross sections at the GANIL-NFS facility

NASA Astrophysics Data System (ADS)

Tarrío, Diego; Prokofiev, Alexander V.; Gustavsson, Cecilia; Jansson, Kaj; Andersson-Sundén, Erik; Al-Adili, Ali; Pomp, Stephan

2017-09-01

Neutron-induced fission cross sections of 235U and 238U are widely used as standards for monitoring of neutron beams and fields. An absolute measurement of these cross sections at an absolute scale, i.e., versus the H(n,p) scattering cross section, is planned with the white neutron beam under construction at the Neutrons For Science (NFS) facility in GANIL. The experimental setup, based on PPACs and ΔE-ΔE-E telescopes containing Silicon and CsI(Tl) detectors, is described. The expected uncertainties are discussed.

Neutron-induced fission cross-section measurement of 234U with quasi-monoenergetic beams in the keV and MeV range using micromegas detectors

NASA Astrophysics Data System (ADS)

Tsinganis, A.; Kokkoris, M.; Vlastou, R.; Kalamara, A.; Stamatopoulos, A.; Kanellakopoulos, A.; Lagoyannis, A.; Axiotis, M.

2017-09-01

Accurate data on neutron-induced fission cross-sections of actinides are essential for the design of advanced nuclear reactors based either on fast neutron spectra or alternative fuel cycles, as well as for the reduction of safety margins of existing and future conventional facilities. The fission cross-section of 234U was measured at incident neutron energies of 560 and 660 keV and 7.5 MeV with a setup based on `microbulk' Micromegas detectors and the same samples previously used for the measurement performed at the CERN n_TOF facility (Karadimos et al., 2014). The 235U fission cross-section was used as reference. The (quasi-)monoenergetic neutron beams were produced via the 7Li(p,n) and the 2H(d,n) reactions at the neutron beam facility of the Institute of Nuclear and Particle Physics at the `Demokritos' National Centre for Scientific Research. A detailed study of the neutron spectra produced in the targets and intercepted by the samples was performed coupling the NeuSDesc and MCNPX codes, taking into account the energy spread, energy loss and angular straggling of the beam ions in the target assemblies, as well as contributions from competing reactions and neutron scattering in the experimental setup. Auxiliary Monte-Carlo simulations were performed with the FLUKA code to study the behaviour of the detectors, focusing particularly on the reproduction of the pulse height spectra of α-particles and fission fragments (using distributions produced with the GEF code) for the evaluation of the detector efficiency. An overview of the developed methodology and preliminary results are presented.

Monte Carlo criticality source convergence in a loosely coupled fuel storage system.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blomquist, R. N.; Gelbard, E. M.

2003-06-10

The fission source convergence of a very loosely coupled array of 36 fuel subassemblies with slightly non-symmetric reflection is studied. The fission source converges very slowly from a uniform guess to the fundamental mode in which about 40% of the fissions occur in one corner subassembly. Eigenvalue and fission source estimates are analyzed using a set of statistical tests similar to those used in MCNP, including the ''drift-in-mean'' test and a new drift-in-mean test using a linear fit to the cumulative estimate drift, the Shapiro-Wilk test for normality, the relative error test, and the ''1/N'' test. The normality test doesmore » not detect a drifting eigenvalue or fission source. Applied to eigenvalue estimates, the other tests generally fail to detect an unconverged solution, but they are sometimes effective when evaluating fission source distributions. None of the test provides completely reliable indication of convergence, although they can detect nonconvergence.« less

FALSTAFF: A New Tool for Fission Fragment Characterization

NASA Astrophysics Data System (ADS)

Doré, D.; Farget, F.; Lecolley, F.-R.; Lehaut, G.; Materna, T.; Pancin, J.; Panebianco, S.; Papaevangelou, Th.

2014-05-01

The future Neutron For Science (NFS) facility to be installed at SPIRAL2 (Caen, France) will produce high intensity neutron beams from hundreds of keV up to 40 MeV. Taking advantage of this facility, data of particular interest to the nuclear community, in view of the development of fast reactor technology, will be measured. The development of an experimental setup called FALSTAFF for a full characterization of actinide fission fragments has been undertaken. Fission fragment isotopic yields and associated neutron multiplicities will be measured as a function of the neutron energy. Based on time-of-flight and residual energy technique, the setup will allow for the simultaneous measurement of the velocity and energy of the complementary fragments. The performance of the time-of-flight detectors of FALSTAFF will be presented and expected resolutions for fragment masses and neutron multiplicities, based on realistic simulations, will be shown.

AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stempien, John Dennis; Rice, Francine Joyce; Harp, Jason Michael

2016-03-01

The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of themore » rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite rings and fuel compacts). This equipment performed well for separating each capsule in the test train and extracting the capsule components. Only a few problems were encountered. In one case, the outermost ring (the sink ring) was cracked during removal of the capsule through tubes. Although the sink ring will be analyzed in order to obtain a mass balance of fission products in the experiment, these cracks do not pose a major concern because the sink ring will not be analyzed in detail to obtain the spatial distribution of fission products. In Capsules 4 and 5, the compacts could not be removed from the inner rings. Strategies for removing the compacts are being evaluated. Sampling the inner rings with the compacts in-place is also an option. Dimensional measurements were made on the compacts, inner rings, outer rings, and sink rings. The diameters of all compacts decreased by 0.5 to 2.0 %. Generally, the extent of diametric shrinkage increased linearly with increasing neutron fluence. Most compact lengths also decreased. Compact lengths decreased with increasing fluence, reaching maximum shrinkage of about 0.9 % at a fast fluence of 4.0x10 25 n/m 2 E > 0.18 MeV. Above this fluence, the extent of length shrinkage appeared to decrease with fluence, and two compacts from Capsule 7 were found to have slightly increased in length (< 0.1 %) after a fluence of 5.2x10 25 n/m 2.« less

AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stempien, John Dennis; Rice, Francine Joyce; Harp, Jason Michael

The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of themore » rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite rings and fuel compacts). This equipment performed well for separating each capsule in the test train and extracting the capsule components. Only a few problems were encountered. In one case, the outermost ring (the sink ring) was cracked during removal of the capsule through tubes. Although the sink ring will be analyzed in order to obtain a mass balance of fission products in the experiment, these cracks do not pose a major concern because the sink ring will not be analyzed in detail to obtain the spatial distribution of fission products. In Capsules 4 and 5, the compacts could not be removed from the inner rings. Strategies for removing the compacts are being evaluated. Sampling the inner rings with the compacts in-place is also an option. Dimensional measurements were made on the compacts, inner rings, outer rings, and sink rings. The diameters of all compacts decreased by 0.5 to 2.0 %. Generally, the extent of diametric shrinkage increased linearly with increasing neutron fluence. Most compact lengths also decreased. Compact lengths decreased with increasing fluence, reaching maximum shrinkage of about 0.9 % at a fast fluence of 4.0x1025 n/m2 E > 0.18 MeV. Above this fluence, the extent of length shrinkage appeared to decrease with fluence, and two compacts from Capsule 7 were found to have slightly increased in length (< 0.1 %) after a fluence of 5.2x1025 n/m2.« less

Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sheryl Morton; Carl Baily; Tom Hill

Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. Itmore » provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a lowtemperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.« less

Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.

Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. Itmore » provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.« less

Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

NASA Astrophysics Data System (ADS)

Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.; Werner, James E.

2006-01-01

Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

Fission Surface Power Technology Demonstration Unit Test Results

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Gibson, Marc A.; Geng, Steven M.; Sanzi, James L.

2016-01-01

The Fission Surface Power (FSP) Technology Demonstration Unit (TDU) is a system-level demonstration of fission power technology intended for use on manned missions to Mars. The Baseline FSP systems consists of a 190 kWt UO2 fast-spectrum reactor cooled by a primary pumped liquid metal loop. This liquid metal loop transfers heat to two intermediate liquid metal loops designed to isolate fission products in the primary loop from the balance of plant. The intermediate liquid metal loops transfer heat to four Stirling Power Conversion Units (PCU), each of which produce 12 kWe (48 kW total) and reject waste heat to two pumped water loops, which transfer the waste heat to titanium-water heat pipe radiators. The FSP TDU simulates a single leg of the baseline FSP system using an electrically heater core simulator, a single liquid metal loop, a single PCU, and a pumped water loop which rejects the waste heat to a Facility Cooling System (FCS). When operated at the nominal operating conditions (modified for low liquid metal flow) during TDU testing the PCU produced 8.9 kW of power at an efficiency of 21.7 percent resulting in a net system power of 8.1 kW and a system level efficiency of 17.2 percent. The reduction in PCU power from levels seen during electrically heated testing is the result of insufficient heat transfer from the NaK heater head to the Stirling acceptor, which could not be tested at Sunpower prior to delivery to the NASA Glenn Research Center (GRC). The maximum PCU power of 10.4 kW was achieved at the maximum liquid metal temperature of 875 K, minimum water temperature of 350 K, 1.1 kg/s liquid metal flow, 0.39 kg/s water flow, and 15.0 mm amplitude at an efficiency of 23.3 percent. This resulted in a system net power of 9.7 kW and a system efficiency of 18.7 percent.

Fission Surface Power Technology Demonstration Unit Test Results

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Gibson, Marc A.; Geng, Steven; Sanzi, James

2016-01-01

The Fission Surface Power (FSP) Technology Demonstration Unit (TDU) is a system-level demonstration of fission power technology intended for use on manned missions to Mars. The Baseline FSP systems consists of a 190 kWt UO2 fast-spectrum reactor cooled by a primary pumped liquid metal loop. This liquid metal loop transfers heat to two intermediate liquid metal loops designed to isolate fission products in the primary loop from the balance of plant. The intermediate liquid metal loops transfer heat to four Stirling Power Conversion Units (PCU), each of which produce 12 kWe (48 kW total) and reject waste heat to two pumped water loops, which transfer the waste heat to titanium-water heat pipe radiators. The FSP TDU simulates a single leg of the baseline FSP system using an electrically heater core simulator, a single liquid metal loop, a single PCU, and a pumped water loop which rejects the waste heat to a Facility Cooling System (FCS). When operated at the nominal operating conditions (modified for low liquid metal flow) during TDU testing the PCU produced 8.9 kW of power at an efficiency of 21.7% resulting in a net system power of 8.1 kW and a system level efficiency of 17.2%. The reduction in PCU power from levels seen during electrically heated testing is the result of insufficient heat transfer from the NaK heater head to the Stirling acceptor, which could not be tested at Sunpower prior to delivery to GRC. The maximum PCU power of 10.4 kW was achieved at the maximum liquid metal temperature of 875 K, minimum water temperature of 350 K, 1.1 kg/s liquid metal flow, 0.39 kg/s water flow, and 15.0 mm amplitude at an efficiency of 23.3%. This resulted in a system net power of 9.7 kW and a system efficiency of 18.7 %.

Abatement of Xenon and Iodine Emissions from Medical Isotope Production Facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doll, Charles G.; Sorensen, Christina M.; Bowyer, Ted W.

2014-04-01

The capability of the International Monitoring System (IMS) to detect xenon from underground nuclear explosions is dependent on the radioactive xenon background. Adding to the background, medical isotope production (MIP) by fission releases several important xenon isotopes including xenon-133 and iodine-133 that decays to xenon-133. The amount of xenon released from these facilities may be equivalent to or exceed that released from an underground nuclear explosion. Thus the release of gaseous fission products within days of irradiation makes it difficult to distinguish MIP emissions from a nuclear explosion. In addition, recent shortages in molybdenum-99 have created interest and investment opportunitiesmore » to design and build new MIP facilities in the United States and throughout the world. Due to the potential increase in the number of MIP facilities, a discussion of abatement technologies provides insight into how the problem of emission control from MIP facilities can be tackled. A review of practices is provided to delineate methods useful for abatement of medical isotopes.« less

The SPES surface ionization source

NASA Astrophysics Data System (ADS)

Manzolaro, M.; D'Agostini, F.; Monetti, A.; Andrighetto, A.

2017-09-01

Ion sources and target systems play a crucial role in isotope separation on line facilities, determining the main characteristics of the radioactive ion beams available for experiments. In the context of the selective production of exotic species (SPES) facility, a 40 MeV, 200 μA proton beam directly impinges a uranium carbide target, generating approximately 1013 fissions per second. The radioactive isotopes produced by the 238U fissions are delivered to the 1+ ion source by means of a tubular transfer line. Here they can be ionized and subsequently accelerated toward the experimental areas. In this work, the characterization of the surface ionization source currently adopted for the SPES facility is presented, taking as a reference ionization efficiency and transversal emittance measurements. The effects of long term operation at high temperature are also illustrated and discussed.

Development of target ion source systems for radioactive beams at GANIL

NASA Astrophysics Data System (ADS)

Bajeat, O.; Delahaye, P.; Couratin, C.; Dubois, M.; Franberg-Delahaye, H.; Henares, J. L.; Huguet, Y.; Jardin, P.; Lecesne, N.; Lecomte, P.; Leroy, R.; Maunoury, L.; Osmond, B.; Sjodin, M.

2013-12-01

The GANIL facility (Caen, France) is dedicated to the acceleration of heavy ion beams including radioactive beams produced by the Isotope Separation On-Line (ISOL) method at the SPIRAL1 facility. To extend the range of radioactive ion beams available at GANIL, using the ISOL method two projects are underway: SPIRAL1 upgrade and the construction of SPIRAL2. For SPIRAL1, a new target ion source system (TISS) using the VADIS FEBIAD ion source coupled to the SPIRAL1 carbon target will be tested on-line by the end of 2013 and installed in the cave of SPIRAL1 for operation in 2015. The SPIRAL2 project is under construction and is being design for using different production methods as fission, fusion or spallation reactions to cover a large area of the chart of nuclei. It will produce among others neutron rich beams obtained by the fission of uranium induced by fast neutrons. The production target made from uranium carbide and heated at 2000 °C will be associated with several types of ion sources. Developments currently in progress at GANIL for each of these projects are presented.

ATRC Neutron Detector Testing Quick Look Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Troy C. Unruh; Benjamin M. Chase; Joy L. Rempe

2013-08-01

As part of the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program, a joint Idaho State University (ISU) / French Alternative Energies and Atomic Energy Commission (CEA) / Idaho National Laboratory (INL) project was initiated in FY-10 to investigate the feasibility of using neutron sensors to provide online measurements of the neutron flux and fission reaction rate in the ATR Critical Facility (ATRC). A second objective was to provide initial neutron spectrum and flux distribution information for physics modeling and code validation using neutron activation based techniques in ATRC as well as ATR during depressurized operations. Detailed activationmore » spectrometry measurements were made in the flux traps and in selected fuel elements, along with standard fission rate distribution measurements at selected core locations. These measurements provide additional calibration data for the real-time sensors of interest as well as provide benchmark neutronics data that will be useful for the ATR Life Extension Program (LEP) Computational Methods and V&V Upgrade project. As part of this effort, techniques developed by Prof. George Imel will be applied by Idaho State University (ISU) for assessing the performance of various flux detectors to develop detailed procedures for initial and follow-on calibrations of these sensors. In addition to comparing data obtained from each type of detector, calculations will be performed to assess the performance of and reduce uncertainties in flux detection sensors and compare data obtained from these sensors with existing integral methods employed at the ATRC. The neutron detectors required for this project were provided to team participants at no cost. Activation detectors (foils and wires) from an existing, well-characterized INL inventory were employed. Furthermore, as part of an on-going ATR NSUF international cooperation, the CEA sent INL three miniature fission chambers (one for detecting fast flux and two for detecting thermal flux) with associated electronics for assessment. In addition, Prof. Imel, ISU, has access to an inventory of Self-Powered Neutron Detectors (SPNDs) with a range of response times as well as Back-to-Back (BTB) fission chambers from prior research he conducted at the Transient REActor Test Facility (TREAT) facility and Neutron RADiography (NRAD) reactors. Finally, SPNDs from the National Atomic Energy Commission of Argentina (CNEA) were provided in connection with the INL effort to upgrade ATR computational methods and V&V protocols that are underway as part of the ATR LEP. Work during fiscal year 2010 (FY10) focussed on design and construction of Experiment Guide Tubes (EGTs) for positioning the flux detectors in the ATRC N-16 locations as well as obtaining ATRC staff concurrence for the detector evaluations. Initial evaluations with CEA researchers were also started in FY10 but were cut short due to reactor reliability issues. Reactor availability issues caused experimental work to be delayed during FY11/12. In FY13, work resumed; and evaluations were completed. The objective of this "Quick Look" report is to summarize experimental activities performed from April 4, 2013 through May 16, 2013.« less

Reducing Uncertainties in Neutron Induced Fission Cross Sections via a Time Projection Chamber

NASA Astrophysics Data System (ADS)

Magee, Joshua; Niffte Collaboration

2016-09-01

Neutron induced fission cross sections of actinides are of great interest in nuclear energy and stockpile stewardship. Traditionally, measurements of these cross sections have been made with fission chambers, which provide limited information on the actual fragments, and ultimately result in uncertainties on the order of several percent. The Neutron Induced Fission Fragment Tracking Experiment collaboration (NIFFTE) designed and built a fission Time Project Chamber (fission TPC), which provides additional information on these processes, through 3-dimensional tracking, improved particle identification, and in-situ profiles of target and beam non-uniformities. Ultimately, this should provide sub-percent measurements of (n,f) cross-sections. During the 2015 run cycle, measurements of several actinides were performed at the Los Alamos Neutron Science Center (LANSCE) Weapons Neutron Research (WNR) facility. An overview of the fission TPC will be given, as well as the current progress towards a sub-percent measurement of the 239Pu/235U (n,f) cross-section ratio. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

A New Measurement of Neutron Induced Fission Cross Sections

NASA Astrophysics Data System (ADS)

Magee, Joshua; Niffte Collaboration

2017-09-01

Neutron induced fission cross sections of actinides are of great interest in nuclear energy and stockpile stewardship. Traditionally, measurements of these cross sections have been made with fission chambers, which provide limited information on the actual fragments, and ultimately result in uncertainties on the order of several percent. The Neutron Induced Fission ragment Tracking Experiment (NIFFTE) collaboration designed and built a fission Time Projection Chamber (fissionTPC), which provides additional information on these processes, through 3-dimensional tracking, improved particle identification, and in-situ profiles of target and beam non-uniformities. Ultimately, this should provide sub-percent measurements of (n,f) cross-sections. During the 2016 run cycle, measurements of the 238U(n,f)/235U(n,f) cross section shape was performed at the Los Alamos Neutron Science Center (LANSCE) Weapons Neutron Research (WNR) facility. An overview of the fission TPC will be given, as well as these recently reported results. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

Application of pulsed multi-ion irradiations in radiation damage research: A stochastic cluster dynamics simulation study

NASA Astrophysics Data System (ADS)

Hoang, Tuan L.; Nazarov, Roman; Kang, Changwoo; Fan, Jiangyuan

2018-07-01

Under the multi-ion irradiation conditions present in accelerated material-testing facilities or fission/fusion nuclear reactors, the combined effects of atomic displacements with radiation products may induce complex synergies in the structural materials. However, limited access to multi-ion irradiation facilities and the lack of computational models capable of simulating the evolution of complex defects and their synergies make it difficult to understand the actual physical processes taking place in the materials under these extreme conditions. In this paper, we propose the application of pulsed single/dual-beam irradiation as replacements for the expensive steady triple-beam irradiation to study radiation damages in materials under multi-ion irradiation.

Design, fabrication, and operation of capsules for the irradiation testing of candidate advanced space reactor fuel pins

NASA Technical Reports Server (NTRS)

Thoms, K. R.

1975-01-01

Fuel irradiation experiments were designed, built, and operated to test uranium mononitride (UN) fuel clad in tungsten-lined T-111 and uranium dioxide fuel clad in both tungsten-lined T-111 and tungsten-lined Nb-1% Zr. A total of nine fuel pins was irradiated at average cladding temperatures ranging from 931 to 1015 C. The UN experiments, capsules UN-4 and -5, operated for 10,480 and 10,037 hr, respectively, at an average linear heat generation rate of 10 kW/ft. The UO2 experiment, capsule UN-6, operated for 8333 hr at an average linear heat generation rate of approximately 5 kW/ft. Following irradiation, the nine fuel pins were removed from their capsules, externally examined, and sent to the NASA Plum Brook Facility for more detailed postirradiation examination. During visual examination, it was discovered that the cladding of the fuel pin containing dense UN in each of capsules UN-4 and -5 had failed, exposing the UN fuel to the NaK in which the pins were submerged and permitting the release of fission gas from the failed pins. A rough analysis of the fission gas seen in samples of the gas in the fuel pin region indicated fission gas release-to-birth rates from these fuel pins in the range of .00001.

Initial Testing of the Stainless Steel NaK-Cooled Circuit (SNaKC)

NASA Technical Reports Server (NTRS)

Garber, Anne; Godfroy, Thomas

2007-01-01

An actively pumped alkali metal flow circuit, designed and fabricated at the NASA Marshall Space Flight Center, is currently undergoing testing in the Early Flight Fission Test Facility (EFF-TF). Sodium potassium (NaK) was selected as the primary coolant. Basic circuit components include: simulated reactor core, NaK to gas heat exchanger, electromagnetic liquid metal pump, liquid metal flowmeter, load/drain reservoir, expansion reservoir, test section, and instrumentation. Operation of the circuit is based around the 37-pin partial-array core (pin and flow path dimensions are the same as those in a full core), designed to operate at 33 kWt. This presentation addresses the construction, fill and initial testing of the Stainless Steel NaK-Cooled Circuit (SNaKC).

Fission products and nuclear fuel behaviour under severe accident conditions part 3: Speciation of fission products in the VERDON-1 sample

NASA Astrophysics Data System (ADS)

Le Gall, C.; Geiger, E.; Gallais-During, A.; Pontillon, Y.; Lamontagne, J.; Hanus, E.; Ducros, G.

2017-11-01

Qualitative and quantitative analyses on the VERDON-1 sample made it possible to obtain valuable information on fission product behaviour in the fuel during the test. A promising methodology based on the quantitative results of post-test characterisations has been implemented to assess the release fraction of non γ-emitter fission products. The order of magnitude of the estimated release fractions for each fission product was consistent with their class of volatility.

High-pressure swing system for measurements of radioactive fission gases in air samples

NASA Astrophysics Data System (ADS)

Schell, W. R.; Vives-Battle, J.; Yoon, S. R.; Tobin, M. J.

1999-01-01

Radionuclides emitted from nuclear reactors, fuel reprocessing facilities and nuclear weapons tests are distributed widely in the atmosphere but have very low concentrations. As part of the Comprehensive Test Ban Treaty (CTBT), identification and verification of the emission of radionuclides from such sources are fundamental in maintaining nuclear security. To detect underground and underwater nuclear weapons tests, only the gaseous components need to be analyzed. Equipment has now been developed that can be used to collect large volumes of air, separate and concentrate the radioactive gas constituents, such as xenon and krypton, and measure them quantitatively. By measuring xenon isotopes with different half-lives, the time since the fission event can be determined. Developments in high-pressure (3500 kPa) swing chromatography using molecular sieve adsorbents have provided the means to collect and purify trace quantities of the gases from large volumes of air automatically. New scintillation detectors, together with timing and pulse shaping electronics, have provided the low-background levels essential in identifying the gamma ray, X-ray, and electron energy spectra of specific radionuclides. System miniaturization and portability with remote control could be designed for a field-deployable production model.

Data summary report for fission product release test VI-5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osborne, M.F.; Lorenz, R.A.; Travis, J.R.

Test VI-5, the fifth in a series of high-temperature fission product release tests in a vertical test apparatus, was conducted in a flowing mixture of hydrogen and helium. The test specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium which had been irradiated to a burnup of {approximately}42 MWd/kg. Using a hot cell-mounted test apparatus, the fuel rod was heated in an induction furnace under simulated LWR accident conditions to two test temperatures, 2000 K for 20 min and then 2700 K for an additional 20 min. The released fission products were collected inmore » three sequentially operated collection trains on components designed to measure fission product transport characteristics and facilitate sampling and analysis. The results from this test were compared with those obtained in previous tests in this series and with the CORSOR-M and ORNL diffusion release models for fission product release. 21 refs., 19 figs., 12 tabs.« less

Results of Uranium Dioxide-Tungsten Irradiation Test and Post-Test Examination

NASA Technical Reports Server (NTRS)

Collins, J. F.; Debogdan, C. E.; Diianni, D. C.

1973-01-01

A uranium dioxide (UO2) fueled capsule was fabricated and irradiated in the NASA Plum Brook Reactor Facility. The capsule consisted of two bulk UO2 specimens clad with chemically vapor deposited tungsten (CVD W) 0.762 and 0.1016 cm (0.030-and 0.040-in.) thick, respectively. The second specimen with 0.1016-cm (0.040-in.) thick cladding was irradiated at temperature for 2607 hours, corresponding to an average burnup of 1.516 x 10 to the 20th power fissions/cu cm. Postirradiation examination showed distortion in the bottom end cap, failure of the weld joint, and fracture of the central vent tube. Diametral growth was 1.3 percent. No evidence of gross interaction between CVD tungsten or arc-cast tungsten cladding and the UO2 fuel was observed. Some of the fission gases passed from the fuel cavity to the gas surrounding the fuel specimen via the vent tube and possibly the end-cap weld failure. Whether the UO2 loss rates through the vent tube were within acceptable limits could not be determined in view of the end-cap weld failure.

NASA's Kilopower Reactor Development and the Path to Higher Power Missions

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Oleson, Steven R.; Poston, David I.; McClure, Patrick

2017-01-01

The development of NASAs Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

NASA's Kilopower Reactor Development and the Path to Higher Power Missions

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Oleson, Steven R.; Poston, Dave I.; McClure, Patrick

2017-01-01

The development of NASA's Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

Neutron-rich rare-isotope production from projectile fission of heavy nuclei near 20 MeV/nucleon beam energy

NASA Astrophysics Data System (ADS)

Vonta, N.; Souliotis, G. A.; Loveland, W.; Kwon, Y. K.; Tshoo, K.; Jeong, S. C.; Veselsky, M.; Bonasera, A.; Botvina, A.

2016-12-01

We investigate the possibilities of producing neutron-rich nuclides in projectile fission of heavy beams in the energy range of 20 MeV/nucleon expected from low-energy facilities. We report our efforts to theoretically describe the reaction mechanism of projectile fission following a multinucleon transfer collision at this energy range. Our calculations are mainly based on a two-step approach: The dynamical stage of the collision is described with either the phenomenological deep-inelastic transfer model (DIT) or with the microscopic constrained molecular dynamics model (CoMD). The de-excitation or fission of the hot heavy projectile fragments is performed with the statistical multifragmentation model (SMM). We compared our model calculations with our previous experimental projectile-fission data of 238U (20 MeV/nucleon) + 208Pb and 197Au (20 MeV/nucleon) + 197Au and found an overall reasonable agreement. Our study suggests that projectile fission following peripheral heavy-ion collisions at this energy range offers an effective route to access very neutron-rich rare isotopes toward and beyond the astrophysical r-process path.

Testing actinide fission yield treatment in CINDER90 for use in MCNP6 burnup calculations

DOE PAGES

Fensin, Michael Lorne; Umbel, Marissa

2015-09-18

Most of the development of the MCNPX/6 burnup capability focused on features that were applied to the Boltzman transport or used to prepare coefficients for use in CINDER90, with little change to CINDER90 or the CINDER90 data. Though a scheme exists for best solving the coupled Boltzman and Bateman equations, the most significant approximation is that the employed nuclear data are correct and complete. Thus, the CINDER90 library file contains 60 different actinide fission yields encompassing 36 fissionable actinides (thermal, fast, high energy and spontaneous fission). Fission reaction data exists for more than 60 actinides and as a result, fissionmore » yield data must be approximated for actinides that do not possess fission yield information. Several types of approximations are used for estimating fission yields for actinides which do not possess explicit fission yield data. The objective of this study is to test whether or not certain approximations of fission yield selection have any impact on predictability of major actinides and fission products. Further we assess which other fission products, available in MCNP6 Tier 3, result in the largest difference in production. Because the CINDER90 library file is in ASCII format and therefore easily amendable, we assess reasons for choosing, as well as compare actinide and major fission product prediction for the H. B. Robinson benchmark for, three separate fission yield selection methods: (1) the current CINDER90 library file method (Base); (2) the element method (Element); and (3) the isobar method (Isobar). Results show that the three methods tested result in similar prediction of major actinides, Tc-99 and Cs-137; however, certain fission products resulted in significantly different production depending on the method of choice.« less

Demonstrating the Viability and Affordability of Nuclear Surface Power Systems

NASA Technical Reports Server (NTRS)

Vandyke, Melissa K.

2006-01-01

A set of tasks have been identified to help demonstrate the viability, performance, and affordability of surface fission systems. Completion of these tasks will move surface fission systems closer to reality by demonstrating affordability and performance potential. Tasks include fabrication and test of a 19-pin section of a Surface Power Unit Demonstrator (SPUD); design, fabrication, and utilization of thermal simulators optimized for surface fission' applications; design, fabrication, and utilization of GPHS module thermal simulators; design, fabrication, and test of a fission surface power system shield; and work related to potential fission surface power fuel/clad systems. Work on the SPUD will feed directly into joint NASA MSFC/NASA GRC fabrication and test of a surface power plant Engineering Development Unit (EDU). The goal of the EDU will be to perform highly realistic thermal, structural, and electrical testing on an integrated fission surface power system. Fission thermal simulator work will help enable high fidelity non-nuclear testing of pumped NaK surface fission power systems. Radioisotope thermal simulator work will help enable design and development of higher power radioisotope systems (power ultimately limited by Pu-238 availability). Shield work is designed to assess the potential of using a water neutron shield on the surface of the moon. Fuels work is geared toward assessing the current potential of using fuels that have already flown in space.

SPES and the neutron facilities at Laboratori Nazionali di Legnaro

NASA Astrophysics Data System (ADS)

Silvestrin, L.; Bisello, D.; Esposito, J.; Mastinu, P.; Prete, G.; Wyss, J.

2016-03-01

The SPES Radioactive Ion Beam (RIB) facility, now in the construction phase at INFN-LNL, has the aim to provide high-intensity and high-quality beams of neutron-rich nuclei for nuclear physics research as well as to develop an interdisciplinary research center based on the cyclotron proton beam. The SPES system is based on a dual-exit high-current cyclotron, with tunable proton beam energy 35-70MeV and 0.20-0.75mA. The first exit is used as proton driver to supply an ISOL system with an UCx Direct Target able to sustain a power of 10kW. The expected fission rate in the target is of the order of 10^{13} fissions per second. The exotic isotopes will be re-accelerated by the ALPI superconducting LINAC at energies of 10 A MeV and higher, for masses around A=130 amu, with an expected beam intensity of 10^7 - 10^9 pps. The second exit will be used for applied physics: radioisotope production for medicine and neutrons for material studies. Fast neutron spectra will be produced by the proton beam interaction with a conversion target. A production rate in excess of 10^{14} n/s can be achieved: this opens up the prospect of a high-flux neutron irradiation facility (NEPIR) to produce both discrete and continuous energy neutrons. A direct proton beam line is also envisaged. NEPIR and the direct proton line would dramatically increase the wide range of irradiation facilities presently available at LNL. We also present LENOS, a proposed project dedicated to accurate neutron cross-sections measurements using intense, well-characterized, broad energy neutron beams. Other activities already in operation at LNL are briefly reviewed: the SIRAD facility for proton and heavy-ion irradiation at the TANDEM-ALPI accelerator and the BELINA test facility at CN van de Graaff accelerator.

Rocky Flats CAAS System Recalibrated, Retested, and Analyzed to Install in the Criticality Experiments Facility at the Nevada Test Site

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, S; Heinrichs, D; Biswas, D

2009-05-27

Neutron detectors and control panels transferred from the Rocky Flats Plant (RFP) were recalibrated and retested for redeployment to the CEF. Testing and calibration were successful with no failure to any equipment. Detector sensitivity was tested at a TRIGA reactor, and the response to thermal neutron flux was satisfactory. MCNP calculated minimum fission yield ({approx} 2 x 10{sup 15} fissions) was applied to determine the thermal flux at selected detector positions at the CEF. Thermal flux levels were greater than 6.39 x 10{sup 6} (n/cm{sup 2}-sec), which was about four orders of magnitude greater than the minimum alarm flux. Calculationsmore » of detector survivable distances indicate that, to be out of lethal area, a detector needs to be placed greater than 15 ft away from a maximum credible source. MCNP calculated flux/dose results were independently verified by COG. CAAS calibration and the testing confirmed that the RFP CAAS system is performing its functions as expected. New criteria for the CAAS detector placement and 12-rad zone boundaries at the CEF are established. All of the CAAS related documents and hardware have been transferred from LLNL to NSTec for installation at the CEF high bay areas.« less

Design and Test Plans for a Non-Nuclear Fission Power System Technology Demonstration Unit

NASA Technical Reports Server (NTRS)

Mason, Lee; Palac, Donald; Gibson, Marc; Houts, Michael; Warren, John; Werner, James; Poston, David; Qualls, Arthur Lou; Radel, Ross; Harlow, Scott

2012-01-01

A joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) team is developing concepts and technologies for affordable nuclear Fission Power Systems (FPSs) to support future exploration missions. A key deliverable is the Technology Demonstration Unit (TDU). The TDU will assemble the major elements of a notional FPS with a non-nuclear reactor simulator (Rx Sim) and demonstrate system-level performance in thermal vacuum. The Rx Sim includes an electrical resistance heat source and a liquid metal heat transport loop that simulates the reactor thermal interface and expected dynamic response. A power conversion unit (PCU) generates electric power utilizing the liquid metal heat source and rejects waste heat to a heat rejection system (HRS). The HRS includes a pumped water heat removal loop coupled to radiator panels suspended in the thermal-vacuum facility. The basic test plan is to subject the system to realistic operating conditions and gather data to evaluate performance sensitivity, control stability, and response characteristics. Upon completion of the testing, the technology is expected to satisfy the requirements for Technology Readiness Level 6 (System Demonstration in an Operational and Relevant Environment) based on the use of high-fidelity hardware and prototypic software tested under realistic conditions and correlated with analytical predictions.

Design and Test Plans for a Non-Nuclear Fission Power System Technology Demonstration Unit

NASA Astrophysics Data System (ADS)

Mason, L.; Palac, D.; Gibson, M.; Houts, M.; Warren, J.; Werner, J.; Poston, D.; Qualls, L.; Radel, R.; Harlow, S.

A joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) team is developing concepts and technologies for affordable nuclear Fission Power Systems (FPSs) to support future exploration missions. A key deliverable is the Technology Demonstration Unit (TDU). The TDU will assemble the major elements of a notional FPS with a non-nuclear reactor simulator (Rx Sim) and demonstrate system-level performance in thermal vacuum. The Rx Sim includes an electrical resistance heat source and a liquid metal heat transport loop that simulates the reactor thermal interface and expected dynamic response. A power conversion unit (PCU) generates electric power utilizing the liquid metal heat source and rejects waste heat to a heat rejection system (HRS). The HRS includes a pumped water heat removal loop coupled to radiator panels suspended in the thermal-vacuum facility. The basic test plan is to subject the system to realistic operating conditions and gather data to evaluate performance sensitivity, control stability, and response characteristics. Upon completion of the testing, the technology is expected to satisfy the requirements for Technology Readiness Level 6 (System Demonstration in an Operational and Relevant Environment) based on the use of high-fidelity hardware and prototypic software tested under realistic conditions and correlated with analytical predictions.

New infrastructure for studies of transmutation and fast systems concepts

NASA Astrophysics Data System (ADS)

Panza, Fabio; Firpo, Gabriele; Lomonaco, Guglielmo; Osipenko, Mikhail; Ricco, Giovanni; Ripani, Marco; Saracco, Paolo; Viberti, Carlo Maria

2017-09-01

In this work we report initial studies on a low power Accelerator-Driven System as a possible experimental facility for the measurement of relevant integral nuclear quantities. In particular, we performed Monte Carlo simulations of minor actinides and fission products irradiation and estimated the fission rate within fission chambers in the reactor core and the reflector, in order to evaluate the transmutation rates and the measurement sensitivity. We also performed a photo-peak analysis of available experimental data from a research reactor, in order to estimate the expected sensitivity of this analysis method on the irradiation of samples in the ADS considered.

A low power ADS for transmutation studies in fast systems

NASA Astrophysics Data System (ADS)

Panza, Fabio; Firpo, Gabriele; Lomonaco, Guglielmo; Osipenko, Mikhail; Ricco, Giovanni; Ripani, Marco; Saracco, Paolo; Viberti, Carlo Maria

2017-12-01

In this work, we report studies on a fast low power accelerator driven system model as a possible experimental facility, focusing on its capabilities in terms of measurement of relevant integral nuclear quantities. In particular, we performed Monte Carlo simulations of minor actinides and fission products irradiation and estimated the fission rate within fission chambers in the reactor core and the reflector, in order to evaluate the transmutation rates and the measurement sensitivity. We also performed a photo-peak analysis of available experimental data from a research reactor, in order to estimate the expected sensitivity of this analysis method on the irradiation of samples in the ADS considered.

Design study of 10 kW direct fission target for RISP project

NASA Astrophysics Data System (ADS)

Tshoo, K.; Jang, D. Y.; Woo, H. J.; Kang, B. H.; Kim, G. D.; Hwang, W.; Kim, Y. K.

2014-03-01

We are developing Isotope Separation On-Line (ISOL) target system, which consists of 1.3 mm-thick uranium-carbide multi-disks and cylindrical tantalum heater, to be installed in new facility for Rare Isotope Science Project in Korea. The intense neutron-rich nuclei are produced via the fission process using the uranium carbide targets with a 70 MeV proton beam. The fission rate was estimated to be ˜1.5 × 1013/sec for 10 kW proton beam. The target system has been designed to be operated at a temperature of ˜2000 °C so as to improve the release effciency.

Operational Results From a High Power Alternator Test Bed

NASA Technical Reports Server (NTRS)

Birchenough, Arthur; Hervol, David

2007-01-01

The Alternator Test Unit (ATU) in the Lunar Power System Facility (LPSF) located at the NASA Glenn Research Center (GRC) in Cleveland, Ohio was used to simulate the operating conditions and evaluate the performance of the ATU and its interaction with various LPSF components in accordance with the current Fission Surface Power System (FSPS) requirements. The testing was carried out at the breadboard development level. These results successfully demonstrated excellent ATU power bus characteristics and rectified user load power quality during steady state and transient conditions. Information gained from this work could be used to assist the design and primary power quality considerations for a possible future FSPS. This paper describes the LPSF components and some preliminary test results.

10 CFR 140.3 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... atomic weapon, designed or used to sustain nuclear fission in a self-supporting chain reaction. (g... experiments; or (ii) A liquid fuel loading; or (iii) An experimental facility in the core in excess of 16... in the isotope 235, except laboratory scale facilities designed or used for experimental or...

10 CFR 140.3 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... in the isotope 235, except laboratory scale facilities designed or used for experimental or... atomic weapon, designed or used to sustain nuclear fission in a self-supporting chain reaction. (g... experiments; or (ii) A liquid fuel loading; or (iii) An experimental facility in the core in excess of 16...

10 CFR 140.3 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... in the isotope 235, except laboratory scale facilities designed or used for experimental or... atomic weapon, designed or used to sustain nuclear fission in a self-supporting chain reaction. (g... experiments; or (ii) A liquid fuel loading; or (iii) An experimental facility in the core in excess of 16...

Diversification of 99Mo/99mTc separation: non–fission reactor production of 99Mo as a strategy for enhancing 99mTc availability.

PubMed

Pillai, Maroor R A; Dash, Ashutosh; Knapp, Furn F Russ

2015-01-01

This paper discusses the benefits of obtaining (99m)Tc from non-fission reactor-produced low-specific-activity (99)Mo. This scenario is based on establishing a diversified chain of facilities for the distribution of (99m)Tc separated from reactor-produced (99)Mo by (n,γ) activation of natural or enriched Mo. Such facilities have expected lower investments than required for the proposed chain of cyclotrons for the production of (99m)Tc. Facilities can receive and process reactor-irradiated Mo targets then used for extraction of (99m)Tc over a period of 2 wk, with 3 extractions on the same day. Estimates suggest that a center receiving 1.85 TBq (50 Ci) of (99)Mo once every 4 d can provide 1.48-3.33 TBq (40-90 Ci) of (99m)Tc daily. This model can use research reactors operating in the United States to supply current (99)Mo needs by applying natural (nat)Mo targets. (99)Mo production capacity can be enhanced by using (98)Mo-enriched targets. The proposed model reduces the loss of (99)Mo by decay and avoids proliferation as well as waste management issues associated with fission-produced (99)Mo.

Transport calculation of neutrons leaked to the surroundings of the facilities by the JCO criticality accident in Tokai-mura.

PubMed

Imanaka, T

2001-09-01

A transport calculation of the neutrons leaked to the environment by the JCO criticality accident was carried out based on three-dimensional geometrical models of the buildings within the JCO territory. Our work started from an initial step to simulate the leakage process of neutrons from the precipitation tank, and proceeded to a step to calculate the neutron propagation throughout the JCO facilities. The total fission number during the accident in the precipitation tank was evaluated to be 2.5 x 10(18) by comparing the calculated neutron-induced activities per 235U fission with the measured values in a stainless-steel net sample taken 2 m from the precipitation tank. Shield effects by various structures within the JCO facilities were evaluated by comparing the present results with a previous calculation using two-dimensional models which suppose a point source of the fission spectrum in the air above the ground without any shield structures. The shield effect by the precipitation tank, itself, was obtained to be a factor of 3. The shield factor by the conversion building varied between 1.1 and 2, depending on the direction from the building. The shield effect by the surrounding buildings within the JCO territory was between I and 5, also depending on the direction.

An Overview of Facilities and Capabilities to Support the Development of Nuclear Thermal Propulsion

DOE Office of Scientific and Technical Information (OSTI.GOV)

James Werner; Sam Bhattacharyya; Mike Houts

Abstract. The future of American space exploration depends on the ability to rapidly and economically access locations of interest throughout the solar system. There is a large body of work (both in the US and the Former Soviet Union) that show that Nuclear Thermal Propulsion (NTP) is the most technically mature, advanced propulsion system that can enable this rapid and economical access by its ability to provide a step increase above what is a feasible using a traditional chemical rocket system. For an NTP system to be deployed, the earlier measurements and recent predictions of the performance of the fuelmore » and the reactor system need to be confirmed experimentally prior to launch. Major fuel and reactor system issues to be addressed include fuel performance at temperature, hydrogen compatibility, fission product retention, and restart capability. The prime issue to be addressed for reactor system performance testing involves finding an affordable and environmentally acceptable method to test a range of engine sizes using a combination of nuclear and non-nuclear test facilities. This paper provides an assessment of some of the capabilities and facilities that are available or will be needed to develop and test the nuclear fuel, and reactor components. It will also address briefly options to take advantage of the greatly improvement in computation/simulation and materials processing capabilities that would contribute to making the development of an NTP system more affordable. Keywords: Nuclear Thermal Propulsion (NTP), Fuel fabrication, nuclear testing, test facilities.« less

Nuclear fission: a review of experimental advances and phenomenology

NASA Astrophysics Data System (ADS)

Andreyev, A. N.; Nishio, K.; Schmidt, K.-H.

2018-01-01

In the last two decades, through technological, experimental and theoretical advances, the situation in experimental fission studies has changed dramatically. With the use of advanced production and detection techniques both much more detailed and precise information can now be obtained for the traditional regions of fission research and, crucially, new regions of nuclei have become routinely accessible for fission studies. This work first of all reviews the recent developments in experimental fission techniques, in particular the resurgence of transfer-induced fission reactions with light and heavy ions, the emerging use of inverse-kinematic approaches, both at Coulomb and relativistic energies, and of fission studies with radioactive beams. The emphasis on the fission-fragment mass and charge distributions will be made in this work, though some of the other fission observables, such as prompt neutron and γ-ray emission will also be reviewed. A particular attention will be given to the low-energy fission in the so far scarcely explored nuclei in the very neutron-deficient lead region. They recently became the focus for several complementary experimental studies, such as β-delayed fission with radioactive beams at ISOLDE(CERN), Coulex-induced fission of relativistic secondary beams at FRS(GSI), and several prompt fusion–fission studies. The synergy of these approaches allows a unique insight in the new region of asymmetric fission around {\\hspace{0pt}}180 Hg, recently discovered at ISOLDE. Recent extensive theoretical efforts in this region will also be outlined. The unprecedented high-quality data for fission fragments, completely identified in Z and A, by means of reactions in inverse kinematics at FRS(GSI) and VAMOS(GANIL) will be also reviewed. These experiments explored an extended range of mercury-to-californium elements, spanning from the neutron-deficient to neutron-rich nuclides, and covering both asymmetric, symmetric and transitional fission regions. Some aspects of heavy-ion induced fusion–fission and quasifission reactions will be also discussed, which reveal their dynamical features, such as the fission time scale. The crucial role of the multi-chance fission, probed by means of multinucleon-transfer induced fission reactions, will be highlighted. The review will conclude with the discussion of the new experimental fission facilities which are presently being brought into operation, along with promising ‘next-generation’ fission approaches, which might become available within the next decade.

Neutronics and Transient Calculations for the Conversion of the Transient Reactor Rest Facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Papadias, Dionissios D.

2015-01-01

The Transient Reactor Test Facility (TREAT) is a graphite-reflected, graphitemoderated, and air-cooled reactor fueled with 93.1% enriched UO2 particles dispersed in graphite, with a carbon-to-235U ratio of ~10000:1. TREAT was used to simulate accident conditions by subjecting fuel test samples placed at the center of the core to high energy transient pulses. The transient pulse production is based on the core’s selflimiting nature due to the negative reactivity feedback provided by the fuel graphite as the core temperature rises. The analysis of the conversion of TREAT to low enriched uranium (LEU) is currently underway. This paper presents the analytical methodsmore » used to calculate the transient performance of TREAT in terms of power pulse production and resulting peak core temperatures. The validation of the HEU neutronics TREAT model, the calculation of the temperature distribution and the temperature reactivity feedback as well as the number of fissions generated inside fuel test samples are discussed.« less

Thermal-electric coupled-field finite element modeling and experimental testing of high-temperature ion sources for the production of radioactive ion beams

NASA Astrophysics Data System (ADS)

Manzolaro, M.; Meneghetti, G.; Andrighetto, A.; Vivian, G.; D'Agostini, F.

2016-02-01

In isotope separation on line facilities the target system and the related ion source are two of the most critical components. In the context of the selective production of exotic species (SPES) project, a 40 MeV 200 μA proton beam directly impinges a uranium carbide target, generating approximately 1013 fissions per second. The radioactive isotopes produced in this way are then directed to the ion source, where they can be ionized and finally accelerated to the subsequent areas of the facility. In this work both the surface ion source and the plasma ion source adopted for the SPES facility are presented and studied by means of numerical thermal-electric models. Then, numerical results are compared with temperature and electric potential difference measurements, and finally the main advantages of the proposed simulation approach are discussed.

Gamma rays as probe of fission and quasi-fission dynamics in the reaction 32S + 197Au near the Coulomb barrier

NASA Astrophysics Data System (ADS)

Pulcini, A.; Vardaci, E.; Kozulin, E.; Ashaduzzaman, M.; Borcea, C.; Bracco, A.; Brambilla, S.; Calinescu, S.; Camera, F.; Ciemala, M.; de Canditiis, B.; Dorvaux, O.; Harca, I. M.; Itkis, I.; Kirakosyan, V. V.; Knyazheva, G.; Kozulina, N.; Kolesov, I. V.; La Rana, G.; Maj, A.; Matea, I.; Novikov, K.; Petrone, C.; Quero, D.; Rath, P.; Saveleva, E.; Schmitt, C.; Sposito, G.; Stezowski, O.; Trzaska, W. H.; Wilson, J.

2018-05-01

Compound nucleus fission and quasi-fission are both binary decay channels whose common properties make the experimental separation between them difficult. A way to achieve this separation could be to probe the angular momentum of the binary fragments. This can be done detecting gamma rays in coincidence with the two fragments. As a case study, the reaction 32S + 197Au near the Coulomb barrier has been performed at the Tandem ALTO facility at IPN ORSAY. ORGAM and PARIS, two different gamma detectors arrays, are coupled with the CORSET detector, a two-arm time-of-flight spectrometer. TOF-TOF data were analyzed to reconstruct the mass-energy distribution of the primary fragments coupled with gamma multiplicity and spectroscopic analysis. Preliminary results of will be shown.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youker, Amanda J.; Krebs, John F.; Quigley, Kevin J.

With funding from the National Nuclear Security Administrations Material Management and Minimization Office, Argonne National Laboratory (Argonne) is providing technical assistance to help accelerate the U.S. production of Mo-99 using a non-highly enriched uranium (non-HEU) source. A potential Mo-99 production pathway is by accelerator-initiated fissioning in a subcritical uranyl sulfate solution containing low enriched uranium (LEU). As part of the Argonne development effort, we are undertaking the AMORE (Argonne Molybdenum Research Experiment) project, which is essentially a pilot facility for all phases of Mo-99 production, recovery, and purification. Production of Mo-99 and other fission products in the subcritical target solutionmore » is initiated by putting an electron beam on a depleted uranium (DU) target; the fast neutrons produced in the DU target are thermalized and lead to fissioning of U-235. At the end of irradiation, Mo is recovered from the target solution and separated from uranium and most of the fission products by using a titania column. The Mo is stripped from the column with an alkaline solution. After acidification of the Mo product solution from the recovery column, the Mo is concentrated (and further purified) in a second titania column. The strip solution from the concentration column is then purified with the LEU Modified Cintichem process. A full description of the process can be found elsewhere [1–3]. The initial commissioning steps for the AMORE project include performing a Mo-99 spike test with pH 1 sulfuric acid in the target vessel without a beam on the target to demonstrate the initial Mo separation-and-recovery process, followed by the concentration column process. All glovebox operations were tested with cold solutions prior to performing the Mo-99 spike tests. Two Mo-99 spike tests with pH 1 sulfuric acid have been performed to date. Figure 1 shows the flow diagram for the remotely operated Mo-recovery system for the AMORE project. There are two separate pumps and flow paths for the acid and base operations. The system contains three sample ladders with eight sample loops per ladder for target mixing; column loading, including acid and water washes; and column stripping, including the final water wash.« less

Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron Activated Shield Wall

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael R. Kruzic

2007-09-16

Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility was used in the early to mid-1960s for the testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles in the immediate area. Identified as Corrective Action Unit 115, the TCA facility was decontaminated and decommissioned (D&D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the ''Federal Facilitymore » Agreement and Consent Order''. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously provided technical decisions are made by an experienced decision maker within the site conceptual site model, identified in the Data Quality Objective process. Facility closure involved a seven-step decommissioning strategy. Key lessons learned from the project included: (1) Targeted preliminary investigation activities provided a more solid technical approach, reduced surprises and scope creep, and made the working environment safer for the D&D worker. (2) Early identification of risks and uncertainties provided opportunities for risk management and mitigation planning to address challenges and unanticipated conditions. (3) Team reviews provided an excellent mechanism to consider all aspects of the task, integrated safety into activity performance, increase team unity and ''buy-in'' and promoted innovative and time saving ideas. (4) Development of CED protocols ensured safety and control. (5) The same proven D&D strategy is now being employed on the larger ''sister'' facility, Test Cell C.« less

Prompt neutron emission and energy balance in 235U(n,f)

NASA Astrophysics Data System (ADS)

Göök, Alf; Hambsch, Franz-Josef; Oberstedt, Stephan

2017-09-01

Investigations of prompt fission neutron (PFN) emission are of importance in understanding the fission process in general and the sharing of excitation energy among the fission fragments in particular. Experimental activities at JRC-Geel on PFN emission in response to OECD/NEA nuclear data requests is presented in this contribution. The focus lies on on-going investigations of PFN emission from the reaction 235U(n,f) in the region of the resolved resonances taking place at the GELINA facility. For this reaction strong fluctuations of fission fragment mass distributions and mean total kinetic energy have been observed as a function of incident neutron energy in the resonance region. In addition, fluctuations of prompt neutron multiplicities have also been observed. The goal of the present study is to verify the current knowledge of PFN multiplicity fluctuations and to study correlations with fission fragment properties. The experiment employs a scintillation detector array for neutron detection, while fission fragment properties are determined via the double kinetic energy technique using a position sensitive twin ionization chamber. Results on PFN multiplicity correlations with fission fragment properties from the present study show significant differences compared to earlier studies on this reaction, induced by thermal neutrons. Specifically, the total kinetic energy dependence of the neutron multiplicity per fission shows an inverse slope FX1TKE/FX2ν approximately 35% weaker than observed in earlier studies of thermal neutron induced fission on 235U. The inverse slope is related to the energy carried away per emitted neutron and is, thereby, closely connected to the energy balance of the fission reaction. The present result should have strong impact on the modeling of both prompt neutron and prompt γ-ray emission in fission of the 236U compound nucleus.

Technical Bases to Aid in the Decision of Conducting Full Power Ground Nuclear Tests for Space Fission Reactors

NASA Astrophysics Data System (ADS)

Hixson, Laurie L.; Houts, Michael G.; Clement, Steven D.

2004-02-01

The extent to which, if any, full power ground nuclear testing of space reactors should be performed has been a point of discussion within the industry for decades. Do the benefits outweigh the risks? Are there equivalent alternatives? Can a test facility be constructed (or modified) in a reasonable amount of time? Is the test article an accurate representation of the flight system? Are the costs too restrictive? The obvious benefits of full power ground nuclear testing; obtaining systems integrated reliability data on a full-scale, complete end-to-end system; come at some programmatic risk. Safety related information is not obtained from a full-power ground nuclear test. This paper will discuss and assess these and other technical considerations essential in the decision to conduct full power ground nuclear-or alternative-tests.

Recent MELCOR and VICTORIA Fission Product Research at the NRC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bixler, N.E.; Cole, R.K.; Gauntt, R.O.

1999-01-21

The MELCOR and VICTORIA severe accident analysis codes, which were developed at Sandia National Laboratories for the U. S. Nuclear Regulatory Commission, are designed to estimate fission product releases during nuclear reactor accidents in light water reactors. MELCOR is an integrated plant-assessment code that models the key phenomena in adequate detail for risk-assessment purposes. VICTORIA is a more specialized fission- product code that provides detailed modeling of chemical reactions and aerosol processes under the high-temperature conditions encountered in the reactor coolant system during a severe reactor accident. This paper focuses on recent enhancements and assessments of the two codes inmore » the area of fission product chemistry modeling. Recently, a model for iodine chemistry in aqueous pools in the containment building was incorporated into the MELCOR code. The model calculates dissolution of iodine into the pool and releases of organic and inorganic iodine vapors from the pool into the containment atmosphere. The main purpose of this model is to evaluate the effect of long-term revolatilization of dissolved iodine. Inputs to the model include dose rate in the pool, the amount of chloride-containing polymer, such as Hypalon, and the amount of buffering agents in the containment. Model predictions are compared against the Radioiodine Test Facility (RTF) experiments conduced by Atomic Energy of Canada Limited (AECL), specifically International Standard Problem 41. Improvements to VICTORIA's chemical reactions models were implemented as a result of recommendations from a peer review of VICTORIA that was completed last year. Specifically, an option is now included to model aerosols and deposited fission products as three condensed phases in addition to the original option of a single condensed phase. The three-condensed-phase model results in somewhat higher predicted fission product volatilities than does the single-condensed-phase model. Modeling of U02 thermochemistry was also improved, and results in better prediction of vaporization of uranium from fuel, which can react with released fission products to affect their volatility. This model also improves the prediction of fission product release rates from fuel. Finally, recent comparisons of MELCOR and VICTORIA with International Standard Problem 40 (STORM) data are presented. These comparisons focus on predicted therrnophoretic deposition, which is the dominant deposition mechanism. Sensitivity studies were performed with the codes to examine experimental and modeling uncertainties.« less

Radioactive ion beams produced by neutron-induced fission at ISOLDE

NASA Astrophysics Data System (ADS)

Catherall, R.; Lettry, J.; Gilardoni, S.; Köster, U.; Isolde Collaboration

2003-05-01

The production rates of neutron-rich fission products for the next-generation radioactive beam facility EURISOL [EU-RTD Project EURISOL (HPRI-CT-1999-50001)] are mainly limited by the maximum amount of power deposited by protons in the target. An alternative approach is to use neutron beams to induce fission in actinide targets. This has the advantage of reducing: the energy deposited by the proton beam in the target; contamination from neutron-deficient isobars that would be produced by spallation; and mechanical stress on the target. At ISOLDE CERN [E. Kugler, Hyperfine Interact. 129 (2000) 23], tests have been made on standard ISOLDE actinide targets using fast-neutron bunches produced by bombarding thick, high- Z metal converters with 1 and 1.4 GeV proton pulses. This paper reviews the first applications of converters used at ISOLDE. It highlights the different geometries and the techniques used to compare fission yields produced by the proton beam directly on the target with neutron-induced fission. Results from the six targets already tested, namely UC 2/graphite and ThO 2 targets with tungsten and tantalum converters, are presented. To gain further knowledge for the design of a dedicated target as required by the TARGISOL project [EU-RTD Project TARGISOL (HPRI-CT-2001-50033)], the results are compared to simulations, using the MARS [N.V. Mokhov, S.I. Striganov, A. Van Ginneken, S.G. Mashnik, A.J. Sierk, J. Ranft, MARS code developments, in: 4th Workshop on Simulating Accelerator Radiation Environments, SARE-4, Knoxville, USA, 14-15.9.1998, FERMILAB-PUB-98-379, nucl-th/9812038; N.V. Mokhov, The Mars Code System User's Guide, Fermilab-FN-628, 1995; N.V. Mokhov, MARS Code Developments, Benchmarking and Applications, Fermilab-Conf-00-066, 2000; O.E. Krivosheev, N.V. Mokhov, A New MARS and its Applications, Fermilab-Conf-98/43, 1998] code interfaced with MCNP [J.S. Hendrics, MCNP4C LANL Memo X-5; JSH-2000-3; J.F. Briemesteir (Ed.), MCNP - A General Montecarlo N-Particle Transport Code, Version 4C, LA-13709-M] libraries, of the neutron flux from the converters interacting with the actinide targets.

Radioactive ion beams produced by neutron-induced fission at ISOLDE

NASA Astrophysics Data System (ADS)

Isolde Collaboration; Catherall, R.; Lettry, J.; Gilardoni, S.; Köster, U.

2003-05-01

The production rates of neutron-rich fission products for the next-generation radioactive beam facility EURISOL [EU-RTD Project EURISOL (HPRI-CT-1999-50001)] are mainly limited by the maximum amount of power deposited by protons in the target. An alternative approach is to use neutron beams to induce fission in actinide targets. This has the advantage of reducing: the energy deposited by the proton beam in the target; contamination from neutron-deficient isobars that would be produced by spallation; and mechanical stress on the target. At ISOLDE CERN [E. Kugler, Hyperfine Interact. 129 (2000) 23], tests have been made on standard ISOLDE actinide targets using fast-neutron bunches produced by bombarding thick, high-/Z metal converters with 1 and 1.4 GeV proton pulses. This paper reviews the first applications of converters used at ISOLDE. It highlights the different geometries and the techniques used to compare fission yields produced by the proton beam directly on the target with neutron-induced fission. Results from the six targets already tested, namely UC2/graphite and ThO2 targets with tungsten and tantalum converters, are presented. To gain further knowledge for the design of a dedicated target as required by the TARGISOL project [EU-RTD Project TARGISOL (HPRI-CT-2001-50033)], the results are compared to simulations, using the MARS [N.V. Mokhov, S.I. Striganov, A. Van Ginneken, S.G. Mashnik, A.J. Sierk, J. Ranft, MARS code developments, in: 4th Workshop on Simulating Accelerator Radiation Environments, SARE-4, Knoxville, USA, 14-15.9.1998, FERMILAB-PUB-98-379, nucl-th/9812038; N.V. Mokhov, The Mars Code System User's Guide, Fermilab-FN-628, 1995; N.V. Mokhov, MARS Code Developments, Benchmarking and Applications, Fermilab-Conf-00-066, 2000; O.E. Krivosheev, N.V. Mokhov, A New MARS and its Applications, Fermilab-Conf-98/43, 1998] code interfaced with MCNP [J.S. Hendrics, MCNP4C LANL Memo X-5; JSH-2000-3; J.F. Briemesteir (Ed.), MCNP - A General Montecarlo N-Particle Transport Code, Version 4C, LA-13709-M] libraries, of the neutron flux from the converters interacting with the actinide targets.

Design of a high-flux epithermal neutron beam using 235U fission plates at the Brookhaven Medical Research Reactor.

PubMed

Liu, H B; Brugger, R M; Rorer, D C; Tichler, P R; Hu, J P

1994-10-01

Beams of epithermal neutrons are being used in the development of boron neutron capture therapy for cancer. This report describes a design study in which 235U fission plates and moderators are used to produce an epithermal neutron beam with higher intensity and better quality than the beam currently in use at the Brookhaven Medical Research Reactor (BMRR). Monte Carlo calculations are used to predict the neutron and gamma fluxes and absorbed doses produced by the proposed design. Neutron flux measurements at the present epithermal treatment facility (ETF) were made to verify and compare with the computed results where feasible. The calculations indicate that an epithermal neutron beam produced by a fission-plate converter could have an epithermal neutron intensity of 1.2 x 10(10) n/cm2.s and a fast neutron dose per epithermal neutron of 2.8 x 10(-11) cGy.cm2/nepi plus being forward directed. This beam would be built into the beam shutter of the ETF at the BMRR. The feasibility of remodeling the facility is discussed.

Final Report - Assessment of Testing Options for the NTR at the INL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Howe, Steven D; McLing, Travis L; McCurry, Michael

One of the main technologies that can be developed to dramatically enhance the human exploration of space is the nuclear thermal rocket (NTR). Several studies over the past thirty years have shown that the NTR can reduce the cost of a lunar outpost, reduce the risk of a human mission to Mars, enable fast transits for most missions throughout the solar system, and reduce the cost and time for robotic probes to deep space. Three separate committees of the National Research Council of the National Academy of Sciences have recommended that NASA develop the NTR. One of the primary issuesmore » in development of the NTR is the ability to verify a flight ready unit. Three main methods can be used to validate safe operation of a NTR: 1) Full power, full duration test in an above ground facility that scrubs the rocket exhaust clean of any fission products; 2) Full power , full duration test using the Subsurface Active Filtering of Exhaust (SAFE) technique to capture the exhaust in subsurface strata; 3) Test of the reactor fuel at temperature and power density in a driver reactor with subsequent first test of the fully integrated NTR in space. The first method, the above ground facility, has been studied in the past. The second method, SAFE, has been examined for application at the Nevada Test Site. The third method relies on the fact that the Nuclear Furnace series of tests in 1971 showed that the radioactive exhaust coming from graphite based fuel for the NTR could be completely scrubbed of fission products and the clean hydrogen flared into the atmosphere. Under funding from the MSFC, the Center for Space Nuclear Research (CSNR) at the Idaho National laboratory (INL) has completed a reexamination of Methods 2 and 3 for implementation at the INL site. In short, the effort performed the following: 1) Assess the geology of the INL site and determine a location suitable SAFE testing; 2) Perform calculations of gas transport throughout the geology; 3) Produce a cost estimate of a non-nuclear , sub-scale test using gas injection to validate the computational models; 4) Produce a preliminary cost estimate to build a nuclear furnace equivalent facility to test NTR fuel on a green field location on the INL site. The results show that the INL geology is substantially better suited to the SAFE testing method than the NTS site. The existence of impermeable interbeds just above the sub-surface aquifer ensure that no material from the test, radioactive or not, can enter the water table. Similar beds located just below the surface will prevent any gaseous products from reaching the surface for dispersion. The extremely high permeability of the strata between the interbeds allows rapid dispersion of the rocket exhaust. In addition, the high permeability suggests that a lower back-pressure may develop in the hole against the rocket thrust, which increases safety of operations. Finally, the cost of performing a sub-scale, non-nuclear verification experiment was determined to be $3M. The third method was assessed through discussions with INL staff resident at the site. In essence, any new Category I facility on any DOE site will cost in excess of $250M. Based on the results of this study, a cost estimate for testing a nuclear rocket at the INL site appears to be warranted. Given the fact that a new nuclear fuel may be possible that does not release any fission products, the SAFE testing option appears to be the most affordable.« less

Quantitative NDA measurements of advanced reprocessing product materials containing uranium, neptunium, plutonium, and americium

NASA Astrophysics Data System (ADS)

Goddard, Braden

The ability of inspection agencies and facility operators to measure powders containing several actinides is increasingly necessary as new reprocessing techniques and fuel forms are being developed. These powders are difficult to measure with nondestructive assay (NDA) techniques because neutrons emitted from induced and spontaneous fission of different nuclides are very similar. A neutron multiplicity technique based on first principle methods was developed to measure these powders by exploiting isotope-specific nuclear properties, such as the energy-dependent fission cross sections and the neutron induced fission neutron multiplicity. This technique was tested through extensive simulations using the Monte Carlo N-Particle eXtended (MCNPX) code and by one measurement campaign using the Active Well Coincidence Counter (AWCC) and two measurement campaigns using the Epithermal Neutron Multiplicity Counter (ENMC) with various (alpha,n) sources and actinide materials. Four potential applications of this first principle technique have been identified: (1) quantitative measurement of uranium, neptunium, plutonium, and americium materials; (2) quantitative measurement of mixed oxide (MOX) materials; (3) quantitative measurement of uranium materials; and (4) weapons verification in arms control agreements. This technique still has several challenges which need to be overcome, the largest of these being the challenge of having high-precision active and passive measurements to produce results with acceptably small uncertainties.

Dual-fission chamber and neutron beam characterization for fission product yield measurements using monoenergetic neutrons

NASA Astrophysics Data System (ADS)

Bhatia, C.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.; Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rundberg, R. S.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Macri, R.; Ryan, C.; Sheets, S. A.; Stoyer, M. A.; Tonchev, A. P.

2014-09-01

A program has been initiated to measure the energy dependence of selected high-yield fission products used in the analysis of nuclear test data. We present out initial work of neutron activation using a dual-fission chamber with quasi-monoenergetic neutrons and gamma-counting method. Quasi-monoenergetic neutrons of energies from 0.5 to 15 MeV using the TUNL 10 MV FM tandem to provide high-precision and self-consistent measurements of fission product yields (FPY). The final FPY results will be coupled with theoretical analysis to provide a more fundamental understanding of the fission process. To accomplish this goal, we have developed and tested a set of dual-fission ionization chambers to provide an accurate determination of the number of fissions occurring in a thick target located in the middle plane of the chamber assembly. Details of the fission chamber and its performance are presented along with neutron beam production and characterization. Also presented are studies on the background issues associated with room-return and off-energy neutron production. We show that the off-energy neutron contribution can be significant, but correctable, while room-return neutron background levels contribute less than <1% to the fission signal.

On the conversion of infrared radiation from fission reactor-based photon engine into parallel beam

NASA Astrophysics Data System (ADS)

Gulevich, Andrey V.; Levchenko, Vladislav E.; Loginov, Nicolay I.; Kukharchuk, Oleg F.; Evtodiev, Denis A.; Zrodnikov, Anatoly V.

2002-01-01

The efficiency of infrared radiation conversion from photon engine based on fission reactor into parallel photon beam is discussed. Two different ways of doing that are considered. One of them is to use the parabolic mirror to convert of infrared radiation into parallel photon beam. The another one is based on the use of special lattice consisting of numerous light conductors. The experimental facility and some results are described. .

A generalized method for characterization of 235U and 239Pu content using short-lived fission product gamma spectroscopy

NASA Astrophysics Data System (ADS)

Knowles, Justin; Skutnik, Steven; Glasgow, David; Kapsimalis, Roger

2016-10-01

Rapid nondestructive assay methods for trace fissile material analysis are needed in both nuclear forensics and safeguards communities. To address these needs, research at the Oak Ridge National Laboratory High Flux Isotope Reactor Neutron Activation Analysis facility has developed a generalized nondestructive assay method to characterize materials containing fissile isotopes. This method relies on gamma-ray emissions from short-lived fission products and makes use of differences in fission product yields to identify fissile compositions of trace material samples. Although prior work has explored the use of short-lived fission product gamma-ray measurements, the proposed method is the first to provide a complete characterization of isotopic identification, mass ratios, and absolute mass determination. Successful single fissile isotope mass recoveries of less than 6% recovery bias have been conducted on standards of 235U and 239Pu as low as 12 ng in less than 10 minutes. Additionally, mixtures of fissile isotope standards containing 235U and 239Pu have been characterized as low as 198 ng of fissile mass with less than 7% recovery bias. The generalizability of this method is illustrated by evaluating different fissile isotopes, mixtures of fissile isotopes, and two different irradiation positions in the reactor. It is anticipated that this method will be expanded to characterize additional fissile nuclides, utilize various irradiation facilities, and account for increasingly complex sample matrices.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bustraan, M.; Coehoorn, J.; Veenema, J.J.

This report is a collection of separate contributions on some aspects of the work done on STEK up to February 1970. A description is given of STEK together with the philosophy of its design, i.e. integral measurements of fission product cross sections by a sample oscillator technique in fast reactor spectra. The influences fission products may have on fast breeder reactors are briefly demonstrated by an example. A description of the facility and of the sample oscillator and sample exchange mechanism is given. Some preliminary results of measurements of reactor parameters and the neutron spectrum in the first fast zonemore » in STEK are given. For the use of lead as material for the buffer an argumentation is given. The proposed program for the measurements of the integral fission product cross sections is outlined. The procurement of some, highly active, samples of actual fission products is briefly sketched. (auth)« less

Lithium Circuit Test Section Design and Fabrication

NASA Technical Reports Server (NTRS)

Godfroy, Thomas; Garber, Anne

2006-01-01

The Early Flight Fission - Test Facilities (EFF-TF) team has designed and built an actively pumped lithium flow circuit. Modifications were made to a circuit originally designed for NaK to enable the use of lithium that included application specific instrumentation and hardware. Component scale freeze/thaw tests were conducted to both gain experience with handling and behavior of lithium in solid and liquid form and to supply anchor data for a Generalized Fluid System Simulation Program (GFSSP) model that was modified to include the physics for freeze/thaw transitions. Void formation was investigated. The basic circuit components include: reactor segment, lithium to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and trace heaters. This paper will discuss the overall system design and build and the component testing findings.

Lithium Circuit Test Section Design and Fabrication

NASA Astrophysics Data System (ADS)

Godfroy, Thomas; Garber, Anne; Martin, James

2006-01-01

The Early Flight Fission - Test Facilities (EFF-TF) team has designed and built an actively pumped lithium flow circuit. Modifications were made to a circuit originally designed for NaK to enable the use of lithium that included application specific instrumentation and hardware. Component scale freeze/thaw tests were conducted to both gain experience with handling and behavior of lithium in solid and liquid form and to supply anchor data for a Generalized Fluid System Simulation Program (GFSSP) model that was modified to include the physics for freeze/thaw transitions. Void formation was investigated. The basic circuit components include: reactor segment, lithium to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and trace heaters. This paper discusses the overall system design and build and the component testing findings.

Energy Production Demonstrator for Megawatt Proton Beams

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pronskikh, Vitaly S.; Mokhov, Nikolai V.; Novitski, Igor

2014-07-16

A preliminary study of the Energy Production Demonstrator (EPD) concept - a solid heavy metal target irradiated by GeV-range intense proton beams and producing more energy than consuming - is carried out. Neutron production, fission, energy deposition, energy gain, testing volume and helium production are simulated with the MARS15 code for tungsten, thorium, and natural uranium targets in the proton energy range 0.5 to 120 GeV. This study shows that the proton energy range of 2 to 4 GeV is optimal for both a natU EPD and the tungsten-based testing station that would be the most suitable for proton acceleratormore » facilities. Conservative estimates, not including breeding and fission of plutonium, based on the simulations suggest that the proton beam current of 1 mA will be sufficient to produce 1 GW of thermal output power with the natU EPD while supplying < 8% of that power to operate the accelerator. The thermal analysis shows that the concept considered has a problem due to a possible core meltdown; however, a number of approaches (a beam rastering, in first place) are suggested to mitigate the issue. The efficiency of the considered EPD as a Materials Test Station (MTS) is also evaluated in this study.« less

Characterization of 618-11 solid waste burial ground, disposed waste, and description of the waste generating facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hladek, K.L.

1997-10-07

The 618-11 (Wye or 318-11) burial ground received transuranic (TRTJ) and mixed fission solid waste from March 9, 1962, through October 2, 1962. It was then closed for 11 months so additional burial facilities could be added. The burial ground was reopened on September 16, 1963, and continued operating until it was closed permanently on December 31, 1967. The burial ground received wastes from all of the 300 Area radioactive material handling facilities. The purpose of this document is to characterize the 618-11 solid waste burial ground by describing the site, burial practices, the disposed wastes, and the waste generatingmore » facilities. This document provides information showing that kilogram quantities of plutonium were disposed to the drum storage units and caissons, making them transuranic (TRU). Also, kilogram quantities of plutonium and other TRU wastes were disposed to the three trenches, which were previously thought to contain non-TRU wastes. The site burial facilities (trenches, caissons, and drum storage units) should be classified as TRU and the site plutonium inventory maintained at five kilograms. Other fissile wastes were also disposed to the site. Additionally, thousands of curies of mixed fission products were also disposed to the trenches, caissons, and drum storage units. Most of the fission products have decayed over several half-lives, and are at more tolerable levels. Of greater concern, because of their release potential, are TRU radionuclides, Pu-238, Pu-240, and Np-237. TRU radionuclides also included slightly enriched 0.95 and 1.25% U-231 from N-Reactor fuel, which add to the fissile content. The 618-11 burial ground is located approximately 100 meters due west of Washington Nuclear Plant No. 2. The burial ground consists of three trenches, approximately 900 feet long, 25 feet deep, and 50 feet wide, running east-west. The trenches constitute 75% of the site area. There are 50 drum storage units (five 55-gallon steel drums welded together) buried in three rows in the northeast comer. In addition, five eight-foot diameter caissons are located at the west end of the center row of the drum storage units. Initially, wastes disposed to the caissons and drum storage units were from the 325 and 327 building hot cells. Later, a small amount of remote-handled (RH) waste from the 309 building Plutonium Recycle Test Reactor (PRTR) cells, and the newly built 324 building hot cells, was disposed at the site.« less

DOSE PROFILE MODELING OF IDAHO NATIONAL LABORATORY’S ACTIVE NEUTRON INTERROGATION TEST FACILITY

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. L. Chichester; E. H. Seabury; J. M. Zabriskie

2009-06-01

A new research and development laboratory has been commissioned at Idaho National Laboratory for performing active neutron interrogation research and development. The facility is designed to provide radiation shielding for DT fusion (14.1 MeV) neutron generators (2 x 108 neutrons per second), DD fusion (2.5 MeV) neutron generators (up to 2 x 106 neutrons per second), and 252Cf spontaneous fission neutron sources (6.7 x 107 neutrons per second, 30 micrograms). Shielding at the laboratory is comprised of modular concrete shield blocks 0.76 m thick with tongue-in-groove features to prevent radiation streaming, arranged into one small and one large test vault.more » The larger vault is designed to allow operation of the DT generator and has walls 3.8 m tall, an entrance maze, and a fully integrated electrical interlock system; the smaller test vault is designed for 252Cf and DD neutron sources and has walls 1.9 m tall and a simple entrance maze. Both analytical calculations and numerical simulations were used in the design process for the building to assess the performance of the shielding walls and to ensure external dose rates are within required facility limits. Dose rate contour plots have been generated for the facility to visualize the effectiveness of the shield wall and entrance maze and to illustrate the spatial profile of the radiation dose field above the facility and the effects of skyshine around the vaults.« less

Test Facilities and Experience on Space Nuclear System Developments at the Kurchatov Institute

NASA Astrophysics Data System (ADS)

Ponomarev-Stepnoi, Nikolai N.; Garin, Vladimir P.; Glushkov, Evgeny S.; Kompaniets, George V.; Kukharkin, Nikolai E.; Madeev, Vicktor G.; Papin, Vladimir K.; Polyakov, Dmitry N.; Stepennov, Boris S.; Tchuniyaev, Yevgeny I.; Tikhonov, Lev Ya.; Uksusov, Yevgeny I.

2004-02-01

The complexity of space fission systems and rigidity of requirement on minimization of weight and dimension characteristics along with the wish to decrease expenditures on their development demand implementation of experimental works which results shall be used in designing, safety substantiation, and licensing procedures. Experimental facilities are intended to solve the following tasks: obtainment of benchmark data for computer code validations, substantiation of design solutions when computational efforts are too expensive, quality control in a production process, and ``iron'' substantiation of criticality safety design solutions for licensing and public relations. The NARCISS and ISKRA critical facilities and unique ORM facility on shielding investigations at the operating OR nuclear research reactor were created in the Kurchatov Institute to solve the mentioned tasks. The range of activities performed at these facilities within the implementation of the previous Russian nuclear power system programs is briefly described in the paper. This experience shall be analyzed in terms of methodological approach to development of future space nuclear systems (this analysis is beyond this paper). Because of the availability of these facilities for experiments, the brief description of their critical assemblies and characteristics is given in this paper.

Transient Approximation of SAFE-100 Heat Pipe Operation

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Reid, Robert S.

2005-01-01

Engineers at Los Alamos National Laboratory (LANL) have designed several heat pipe cooled reactor concepts, ranging in power from 15 kWt to 800 kWt, for both surface power systems and nuclear electric propulsion systems. The Safe, Affordable Fission Engine (SAFE) is now being developed in a collaborative effort between LANL and NASA Marshall Space Flight Center (NASA/MSFC). NASA is responsible for fabrication and testing of non-nuclear, electrically heated modules in the Early Flight Fission Test Facility (EFF-TF) at MSFC. In-core heat pipes must be properly thawed as the reactor power starts. Computational models have been developed to assess the expected operation of a specific heat pipe design during start-up, steady state operation, and shutdown. While computationally intensive codes provide complete, detailed analyses of heat pipe thaw, a relatively simple. concise routine can also be applied to approximate the response of a heat pipe to changes in the evaporator heat transfer rate during start-up and power transients (e.g., modification of reactor power level) with reasonably accurate results. This paper describes a simplified model of heat pipe start-up that extends previous work and compares the results to experimental measurements for a SAFE-100 type heat pipe design.

Thermal-electric numerical simulation of a surface ion source for the production of radioactive ion beams

NASA Astrophysics Data System (ADS)

Manzolaro, Mattia; Meneghetti, Giovanni; Andrighetto, Alberto

2010-11-01

In a facility for the production of radioactive ion beams (RIBs), the target system and the ion source are the most critical objects. In the context of the Selective Production of Exotic Species (SPES) project, a proton beam directly impinges a Uranium Carbide production target, generating approximately 10 13 fissions per second. The radioactive isotopes produced by the 238U fissions are then directed to the ion source to acquire a charge state. After that, the radioactive ions obtained are transported electrostatically to the subsequent areas of the facility. In this work the surface ion source at present adopted for the SPES project is studied by means of both analytical and numerical thermal-electric models. The theoretical results are compared with temperature and electric potential difference measurements.

Four-point Bend Testing of Irradiated Monolithic U-10Mo Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rabin, B. H.; Lloyd, W. R.; Schulthess, J. L.

2015-03-01

This paper presents results of recently completed studies aimed at characterizing the mechanical properties of irradiated U-10Mo fuel in support of monolithic base fuel qualification. Mechanical properties were evaluated in four-point bending. Specimens were taken from fuel plates irradiated in the RERTR-12 and AFIP-6 Mk. II irradiation campaigns, and tests were conducted in the Hot Fuel Examination Facility (HFEF) at Idaho National Laboratory (INL). The monolithic fuel plates consist of a U-10Mo fuel meat covered with a Zr diffusion barrier layer fabricated by co-rolling, clad in 6061 Al using a hot isostatic press (HIP) bonding process. Specimens exhibited nominal (fresh)more » fuel meat thickness ranging from 0.25 mm to 0.64 mm, and fuel plate average burnup ranged from approximately 0.4 x 1021 fissions/cm 3 to 6.0 x 1021 fissions/cm 3. After sectioning the fuel plates, the 6061 Al cladding was removed by dissolution in concentrated NaOH. Pre- and post-dissolution dimensional inspections were conducted on test specimens to facilitate accurate analysis of bend test results. Four-point bend testing was conducted on the HFEF Remote Load Frame at a crosshead speed of 0.1 mm/min using custom-designed test fixtures and calibrated load cells. All specimens exhibited substantially linear elastic behavior and failed in a brittle manner. The influence of burnup on the observed slope of the stress-strain curve and the calculated fracture strength is discussed.« less

Precision Gamma-Ray Branching Ratios for Long-Lived Radioactive Nuclei

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tonchev, Anton

Many properties of the high-energy-density environments in nuclear weapons tests, advanced laser-fusion experiments, the interior of stars, and other astrophysical bodies must be inferred from the resulting long-lived radioactive nuclei that are produced. These radioactive nuclei are most easily and sensitively identified by studying the characteristic gamma rays emitted during decay. Measuring a number of decays via detection of the characteristic gamma-rays emitted during the gamma-decay (the gamma-ray branching ratio) of the long-lived fission products is one of the most straightforward and reliable ways to determine the number of fissions that occurred in a nuclear weapon test. The fission productsmore » 147Nd, 144Ce, 156Eu, and certain other long-lived isotopes play a crucial role in science-based stockpile stewardship, however, the large uncertainties (about 8%) on the branching ratios measured for these isotopes are currently limiting the usefulness of the existing data [1,2]. We performed highly accurate gamma-ray branching-ratio measurements for a group of high-atomic-number rare earth isotopes to greatly improve the precision and reliability with which the fission yield and reaction products in high-energy-density environments can be determined. We have developed techniques that take advantage of new radioactive-beam facilities, such as DOE's CARIBU located at Argonne National Laboratory, to produce radioactive samples and perform decay spectroscopy measurements. The absolute gamma-ray branching ratios for 147Nd and 144Ce are reduced <2% precision. In addition, high-energy monoenergetic neutron beams from the FN Tandem accelerator in TUNL at Duke University was used to produce 167Tm using the 169Tm(n,3n) reaction. Fourtime improved branching ratio of 167Tm is used now to measure reaction-in-flight (RIF) neutrons from a burning DT capsule at NIF [10]. This represents the first measurement of RIF neutrons in any laboratory fusion system, and the magnitude of the signal has important implications for fundamental plasma science and for weapons physics.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Remec, Igor; Ronningen, Reginald Martin

The research studied one-step and two-step Isotope Separation on Line (ISOL) targets for future radioactive beam facilities with high driver-beam power through advanced computer simulations. As a target material uranium carbide in the form of foils was used because of increasing demand for actinide targets in rare-isotope beam facilities and because such material was under development in ISAC at TRIUMF when this project started. Simulations of effusion were performed for one-step and two step targets and the effects of target dimensions and foil matrix were studied. Diffusion simulations were limited by availability of diffusion parameters for UC x material atmore » reduced density; however, the viability of the combined diffusion?effusion simulation methodology was demonstrated and could be used to extract physical parameters such as diffusion coefficients and effusion delay times from experimental isotope release curves. Dissipation of the heat from the isotope-producing targets is the limiting factor for high-power beam operation both for the direct and two-step targets. Detailed target models were used to simulate proton beam interactions with the targets to obtain the fission rates and power deposition distributions, which were then applied in the heat transfer calculations to study the performance of the targets. Results indicate that a direct target, with specification matching ISAC TRIUMF target, could operate in 500-MeV proton beam at beam powers up to ~40 kW, producing ~8 10 13 fission/s with maximum temperature in UCx below 2200 C. Targets with larger radius allow higher beam powers and fission rates. For the target radius in the range 9-mm to 30-mm the achievable fission rate increases almost linearly with target radius, however, the effusion delay time also increases linearly with target radius.« less

Assembly and Thermal Hydraulic Test of a Stainless Steel Sodium-Potassium Circuit

NASA Technical Reports Server (NTRS)

Garber, A.; Godfroy, T.; Webster, K.

2007-01-01

Early Flight Fission Test Facilities (EFF-TF) team has been tasked by the NASA Marshall Space Flight Center Nuclear Systems Office to design, fabricate, and test an actively pumped alkali metal flow circuit. The system was originally built for use with lithium, but due to a shift in focus, it was redesigned for use with a eutectic mixture of sodium potassium (NaK). Basic circuit components include: reactor segment, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and a spill reservoir. A 37-pin partial-array core (pin and flow path dimensions are the same as those in a full design) was selected for fabrication and test. This paper summarizes the first fill and checkout testing of the Stainless Steel NaK-Cooled Circuit (SNaKC).

PBF Reactor Building (PER620) basement, inside cubicle 13. Lead bricks ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620) basement, inside cubicle 13. Lead bricks shield the fission product detection system (FPDS). The system detected fission products in pressure loop from in-pile tube. shielding was to prevent other radiation in cubicle from interfering. Assembly of bricks in foreground will slide back to enclose and shield equipment in the three chambers. Date: 1982. INEEL negative no. 82-6376 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

NECTAR—A fission neutron radiography and tomography facility

NASA Astrophysics Data System (ADS)

Bücherl, T.; Lierse von Gostomski, Ch.; Breitkreutz, H.; Jungwirth, M.; Wagner, F. M.

2011-09-01

NECTAR (Neutron Computerized Tomography and Radiography) is a versatile facility for radiographic and tomographic investigations as well as for neutron activation experiments using fission neutrons. The radiation sources for this facility are two plates of highly enriched uranium situated in the moderator vessel in FRM II. Thermal neutrons originating from the main fuel element of the reactor generate in these plates fast neutrons. These can escape through a horizontal beam tube without moderation. The beam can be filtered and manipulated in order to reduce the accompanying gamma radiation and to match the specific experimental tasks. A summary of the main parameters required for experimental set-up and (quantitative) data evaluation is presented. The (measured) spectra of the neutron and gamma radiations are shown along with the effect of different filters on their behavior. The neutron and gamma fluxes, dose rates, L/ D-ratios, etc. and the main parameters of the actually used detection systems for neutron imaging are given, too.

Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nigg, David W.; Nielsen, Joseph W.; Norman, Daren R.

The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be wellmore » outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.« less

Neutron-Induced Fission Cross Sections of 240Pu, 243Am, and natW in the Energy Range 1-200 MeV

NASA Astrophysics Data System (ADS)

Laptev, A. B.; Donets, A. Yu.; Dushin, V. N.; Fomichev, A. V.; Fomichev, A. A.; Haight, R. C.; Shcherbakov, O. A.; Soloviev, S. M.; Tuboltsev, Yu. V.; Vorobyev, A. S.

2005-05-01

A long-range research program devoted to measurements of neutron-induced fission cross-sections of actinides and stable isotopes is under way at the GNEIS facility. By now the new series of experiments for measurements of fission cross-section ratios relative to 235U has been completed for 240Pu, 243Am, and natW in a wide energy range of incident neutrons from 1 MeV to 200 MeV in the frame of the ISTC Project ♯1971. The measurements were performed using the multiplate ionization chamber and time-of-flight techniques. The results obtained in this measurement are presented in comparison with the other data.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Astrophysics Data System (ADS)

Emrich, William J.

2008-01-01

To support a potential future development of a nuclear thermal rocket engine, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components could be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowing hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2008-01-01

To support the eventual development of a nuclear thermal rocket engine, a state-of-the-art experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowing hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts.

Nuclear pursuits

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1993-05-01

This table lists quantities of warheads (in stockpile, peak number per year, total number built, number of known test explosions), weapon development milestones (developers of the atomic bomb and hydrogen bomb, date of first operational ICBM, first nuclear-powered naval SSN in service, first MIRVed missile deployed), and testing milestones (first fission test, type of boosted fission weapon, multistage thermonuclear test, number of months from fission bomb to multistage thermonuclear bomb, etc.), and nuclear infrastructure (assembly plants, plutonium production reactors, uranium enrichment plants, etc.). Countries included in the tally are the United States, Soviet Union, Britain, France, and China.

Initial Operation and Shakedown of the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2014-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Prototypical fuel elements mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission in addition to being exposed to flowing hydrogen. Recent upgrades to NTREES now allow power levels 24 times greater than those achievable in the previous facility configuration. This higher power operation will allow near prototypical power densities and flows to finally be achieved in most prototypical fuel elements.

Total Kinetic Energy and Fragment Mass Distribution of Neutron-Induced Fission of U-233

DOE Office of Scientific and Technical Information (OSTI.GOV)

Higgins, Daniel James; Schmitt, Kyle Thomas; Mosby, Shea Morgan

Properties of fission in U-233 were studied at the Los Alamos Neutron Science Center (LANSCE) at incident neutron energies from thermal to 40 MeV at both the Lujan Neutron Scattering Center flight path 12 and at WNR flight path 90-Left from Dec 2016 to Jan 2017. Fission fragments are observed in coincidence using a twin ionization chamber with Frisch grids. The average total kinetic energy (TKE) released from fission and fragment mass distributions are calculated from observations of energy deposited in the detector and conservation of mass and momentum. Accurate experimental measurements of these parameters are necessary to better understandmore » the fission process and obtain data necessary for calculating criticality. The average TKE released from fission has been well characterized for several isotopes at thermal neutron energy, however, few measurements have been made at fast neutron energies. This experiment expands on previous successful experiments using an ionization chamber to measure TKE and fragment mass distributions of U-235, U-238, and Pu-239. This experiment requires the full spectrum of neutron energies and can therefore only be performed at a small number of facilities in the world. The required full neutron energy spectrum is obtained by combining measurements from WNR 90L and Lujan FP12 at LANSCE.« less

Comparison of the Recently proposed Super Marx Generator Approach to Thermonuclear Ignition with the DT Laser Fusion-Fission Hybrid Concept (LIFE) by the Lawrence Livermore National Laboratory.

NASA Astrophysics Data System (ADS)

Winterberg, Friedwardt

2009-05-01

The recently proposed Super Marx pure deuterium micro-detonation ignition concept [1] is compared to the Lawrence Livermore National Ignition Facility (NIF) laser DT fusion-fission hybrid concept (LIFE) [2]. A typical example of the LIFE concept is a fusion gain 30, and a fission gain of 10, making up for a total gain of 300, with about 10 times more energy released into fission as compared to fusion. This means a substantial release of fission products, as in fusion-less pure fission reactors. In the Super Marx approach for the ignition of a pure deuterium micro-detonation gains of the same magnitude can in theory be reached. If the theoretical prediction can be supported by more elaborate calculations, the Super Marx approach is likely to make lasers obsolete as a means for the ignition of thermonuclear micro-explosions. [1] ``Ignition of a Deuterium Micro-Detonation with a Gigavolt Super Marx Generator,'' Winterberg, F., Journal of Fusion Energy, Springer, 2008. http://www.springerlink.com/content/r2j046177j331241/fulltext.pdf. [2] ``LIFE: Clean Energy from Nuclear Waste,'' https://lasers.llnl.gov/missions/energy&_slash;for&_slash;the&_slash;future/life/

Measuring Light-ion Production and Fission Cross Sections Normalised to H(n,p) Scattering at the Upcoming NFS Facility

NASA Astrophysics Data System (ADS)

Jansson, K.; Gustavsson, C.; Pomp, S.; Prokofiev, A. V.; Scian, G.; Tarrío, D.

2014-05-01

The Medley detector setup is planned to be moved to and used at the new neutron facility NFS where measurements of light-ion production and fission cross-sections are planned at 1-40 MeV. Medley has eight detector telescopes providing ΔE-ΔE-E data, each consisting of two silicon detectors and a CsI(Tl) detector at the back. The telescope setup can be rotated and arranged to cover any angle. Medley has previously been used in many measurements at The Svedberg Laboratory (TSL) in Uppsala mainly with a quasi-mono-energetic neutron beam at 96 and 175 MeV. To be able to do measurements at NFS, which will have a white neutron beam, Medley needs to detect the reaction products with a high timing resolution providing the ToF of the primary neutron. In this paper we discuss the design of the Medley upgrade along with simulations of the setup. We explore the use of Parallel Plate Avalanche Counters (PPACs) which work very well for detecting fission fragments but require more consideration for detecting deeply penetrating particles.

Improved charge breeding efficiency of light ions with an electron cyclotron resonance ion source

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondrasek, R.; Kutsaev, Sergey; Delahaye, P.

2012-11-15

The Californium Rare Isotope Breeder Upgrade is a new radioactive beam facility for the Argonne Tandem Linac Accelerator System (ATLAS). The facility utilizes a {sup 252}Cf fission source coupled with an electron cyclotron resonance ion source to provide radioactive beam species for the ATLAS experimental program. The californium fission fragment distribution provides nuclei in the mid-mass range which are difficult to extract from production targets using the isotope separation on line technique and are not well populated by low-energy fission of uranium. To date the charge breeding program has focused on optimizing these mid-mass beams, achieving high charge breeding efficienciesmore » of both gaseous and solid species including 14.7% for the radioactive species {sup 143}Ba{sup 27+}. In an effort to better understand the charge breeding mechanism, we have recently focused on the low-mass species sodium and potassium which up to present have been difficult to charge breed efficiently. Unprecedented charge breeding efficiencies of 10.1% for {sup 23}Na{sup 7+} and 17.9% for {sup 39}K{sup 10+} were obtained injecting stable Na{sup +} and K{sup +} beams from a surface ionization source.« less

Improved charge breeding efficiency of light ions with an electron cyclotron resonance ion source

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondrasek, R.; Delahaye, P.; Kutsaev, Sergey

2012-11-01

The Californium Rare Isotope Breeder Upgrade is a new radioactive beam facility for the Argonne Tandem Linac Accelerator System (ATLAS). The facility utilizes a 252Cf fission source coupled with an electron cyclotron resonance ion source to provide radioactive beam species for the ATLAS experimental program. The californium fission fragment distribution provides nuclei in the mid-mass range which are difficult to extract from production targets using the isotope separation on line technique and are not well populated by low-energy fission of uranium. To date the charge breeding program has focused on optimizing these mid-mass beams, achieving high charge breeding efficiencies ofmore » both gaseous and solid species including 14.7% for the radioactive species 143Ba27+. In an effort to better understand the charge breeding mechanism, we have recently focused on the low-mass species sodium and potassium which up to present have been difficult to charge breed efficiently. Unprecedented charge breeding efficiencies of 10.1% for 23Na7+ and 17.9% for 39K10+ were obtained injecting stable Na+ and K+ beams from a surface ionization source.« less

Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

NASA Astrophysics Data System (ADS)

Porter, D. L.; Tsai, Hanchung

2012-08-01

The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 °C, cooling to 522 °C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ˜0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher ΔT between fuel center and cladding than at core center, together providing more rare earths at the cladding and more FCCI. This behavior could actually help extend the life of a fuel pin in a "long pin" reactor design to a higher peak fuel burnup.

The Neutrons for Science Facility at SPIRAL-2

NASA Astrophysics Data System (ADS)

Ledoux, X.; Aïche, M.; Avrigeanu, M.; Avrigeanu, V.; Audouin, L.; Balanzat, E.; Ban-d'Etat, B.; Ban, G.; Barreau, G.; Bauge, E.; Bélier, G.; Bem, P.; Blideanu, V.; Blomgren, J.; Borcea, C.; Bouffard, S.; Caillaud, T.; Chatillon, A.; Czajkowski, S.; Dessagne, P.; Doré, D.; Fallot, M.; Farget, F.; Fischer, U.; Giot, L.; Granier, T.; Guillous, S.; Gunsing, F.; Gustavsson, C.; Herber, S.; Jacquot, B.; Jurado, B.; Kerveno, M.; Klix, A.; Landoas, O.; Lecolley, F. R.; Lecolley, J. F.; Lecouey, J. L.; Majerle, M.; Marie, N.; Materna, T.; Mrazek, J.; Negoita, F.; Novak, J.; Oberstedt, S.; Oberstedt, A.; Panebianco, S.; Perrot, L.; Petrascu, M.; Plompen, A. J. M.; Pomp, S.; Ramillon, J. M.; Ridikas, D.; Rossé, B.; Rudolf, G.; Serot, O.; Shcherbakov, O.; Simakov, S. P.; Simeckova, E.; Smith, A. G.; Steckmeyer, J. C.; Sublet, J. C.; Taïeb, J.; Tassan-Got, L.; Takibayev, A.; Tungborn, E.; Thfoin, I.; Tsekhanovich, I.; Varignon, C.; Wieleczko, J. P.

2011-12-01

The "Neutrons for Science" (NFS) facility will be a component of SPIRAL-2, the future accelerator dedicated to the production of very intense radioactive ion beams, under construction at GANIL in Caen (France). NFS will be composed of a pulsed neutron beam for in-flight measurements and irradiation stations for cross-section measurements and material studies. Continuous and quasi-monokinetic energy spectra will be available at NFS respectively produced by the interaction of deuteron beam on thick a Be converter and by the 7Li(p,n) reaction on a thin converter. The flux at NFS will be up to 2 orders of magnitude higher than those of other existing time-of-flight facilities in the 1 MeV to 40 MeV range. NFS will be a very powerful tool for physics and fundamental research as well as applications like the transmutation of nuclear waste, design of future fission and fusion reactors, nuclear medicine or test and development of new detectors.

An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility

NASA Astrophysics Data System (ADS)

Calderoni, P.; Sharpe, J.; Shimada, M.; Denny, B.; Pawelko, B.; Schuetz, S.; Longhurst, G.; Hatano, Y.; Hara, M.; Oya, Y.; Otsuka, T.; Katayama, K.; Konishi, S.; Noborio, K.; Yamamoto, Y.

2011-10-01

The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.

Operators in the Plum Brook Reactor Facility Control Room

NASA Image and Video Library

1970-03-21

Donald Rhodes, left, and Clyde Greer, right, monitor the operation of the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility from the control room. The 60-megawatt test reactor, NASA’s only reactor, was the eighth largest test reactor in the world. The facility was built by the Lewis Research Center in the late 1950s to study the effects of radiation on different materials that could be used to construct nuclear propulsion systems for aircraft or rockets. The reactor went critical for the first time in 1961. For the next two years, two operators were on duty 24 hours per day working on the fission process until the reactor reached its full-power level in 1963. Reactor Operators were responsible for monitoring and controlling the reactor systems. Once the reactor was running under normal operating conditions, the work was relatively uneventful. Normally the reactor was kept at a designated power level within certain limits. Occasionally the operators had to increase the power for a certain test. The shift supervisor and several different people would get together and discuss the change before boosting the power. All operators were required to maintain a Reactor Operator License from the Atomic Energy Commission. The license included six months of training, an eight-hour written exam, a four-hour walkaround, and testing on the reactor controls.

Space Fission Propulsion System Development Status

NASA Astrophysics Data System (ADS)

Houts, M.; Van Dyke, M. K.; Godfroy, T. J.; Pedersen, K. W.; Martin, J. J.; Dickens, R.; Williams, E.; Harper, R.; Salvail, P.; Hrbud, I.

2001-01-01

The world's first man-made self-sustaining fission reaction was achieved in 1942. Since then fission has been used to propel submarines, generate tremendous amounts of electricity, produce medical isotopes, and provide numerous other benefits to society. Fission systems operate independently of solar proximity or orientation, and are thus well suited for deep space or planetary surface missions. In addition, the fuel for fission systems (enriched uranium) is virtually non-radioactive. The primary safety issue with fission systems is avoiding inadvertent system start. Addressing this issue through proper system design is straight-forward. Despite the relative simplicity and tremendous potential of space fission systems, the development and utilization of these systems has proven elusive. The first use of fission technology in space occurred 3 April 1965 with the US launch of the SNAP-10A reactor. There have been no additional US uses of space fission systems. While space fission systems were used extensively by the former Soviet Union, their application was limited to earth-orbital missions. Early space fission systems must be safely and affordably utilized if we are to reap the benefits of advanced space fission systems. NASA's Marshall Space Flight Center, working with Los Alamos National Laboratory (LANL), Sandia National Laboratories, and others, has conducted preliminary research related to a Safe Affordable Fission Engine (SAFE). An unfueled core has been fabricated by LANL, and resistance heaters used to verify predicted core thermal performance by closely mimicking heat from fission. The core is designed to use only established nuclear technology and be highly testable. In FY01 an energy conversion system and thruster will be coupled to the core, resulting in an 'end-to-end' nuclear electric propulsion demonstrator being tested using resistance heaters to closely mimic heat from fission. Results of the SAFE test program will be presented. The applicability of a SAFE-powered electric propulsion system to outer planet science missions will also be discussed.

The Prompt Fission Neutron Spectrum of 235U(n,f) below 2.5 MeV for Incident Neutrons from 0.7 to 20 MeV

NASA Astrophysics Data System (ADS)

Devlin, M.; Gomez, J. A.; Kelly, K. J.; Haight, R. C.; O'Donnell, J. M.; Taddeucci, T. N.; Lee, H. Y.; Mosby, S. M.; Perdue, B. A.; Fotiades, N.; Ullmann, J. L.; Wu, C. Y.; Bucher, B.; Buckner, M. Q.; Henderson, R. A.; Neudecker, D.; White, M. C.; Talou, P.; Rising, M. E.; Solomon, C. J.

2018-02-01

New prompt fission neutron spectrum measurements are reported for 235U(n , f) reactions induced by neutrons with energies from 0.7 to 20 MeV. These measurements cover outgoing neutron energies from 2.5 MeV down to 10 keV, using an array of 6Li-glass scintillators for neutron detection and a double time-of-flight technique. The neutrons were produced at the Weapons Neutron Research facility of the Los Alamos Neutron Science Center. A detailed MCNP® model of the experimental equipment and the surrounding room was used to interpret the experimental results. Backgrounds were measured in situ, making use of the time-dependent singles rates of the various detectors with asynchronous readout from waveform digitizers. The results presented here have been included in a re-evaluation of the fission neutron spectra for this fissioning system, a description of which is presented elsewhere in this issue.

The Prompt Fission Neutron Spectrum of 235U(n,f) below 2.5 MeV for Incident Neutrons from 0.7 to 20 MeV

DOE Office of Scientific and Technical Information (OSTI.GOV)

Devlin, Matthew James; Gomez, Jaime A.; Kelly, Keegan John

New prompt fission neutron spectrum measurements are reported for 235U(n,f) reactions induced by neutrons with energies from 0.7 to 20 MeV. These measurements cover outgoing neutron energies from 2.5 MeV down to 10 keV, using an array of 6Li-glass scintillators for neutron detection and a double time-of-flight technique. The neutrons were produced at the Weapons Neutron Research facility of the Los Alamos Neutron Science Center. A detailed MCNP ® model of the experimental equipment and the surrounding room was used to interpret the experimental results. Backgrounds were measured in situ, making use of the time-dependent singles rates of the variousmore » detectors with asynchronous readout from waveform digitizers. The results presented here have been included in a re-evaluation of the fission neutron spectra for this fissioning system, a description of which is presented elsewhere in this issue.« less

Production of Sn and Sb isotopes in high-energy neutron-induced fission of natU

NASA Astrophysics Data System (ADS)

Mattera, A.; Pomp, S.; Lantz, M.; Rakopoulos, V.; Solders, A.; Al-Adili, A.; Penttilä, H.; Moore, I. D.; Rinta-Antila, S.; Eronen, T.; Kankainen, A.; Pohjalainen, I.; Gorelov, D.; Canete, L.; Nesterenko, D.; Vilén, M.; Äystö, J.

2018-03-01

The first systematic measurement of neutron-induced fission yields has been performed at the upgraded IGISOL-4 facility at the University of Jyväskylä, Finland. The fission products from high-energy neutron-induced fission of nat U were stopped in a gas cell filled with helium buffer gas, and were online separated with a dipole magnet. The isobars, with masses in the range A = 128-133 , were transported to a tape-implantation station and identified using γ -spectroscopy. We report here the relative cumulative isotopic yields of tin ( Z = 50) and the relative independent isotopic yields of antimony ( Z = 51) . Isomeric yield ratios were also obtained for five nuclides. The yields of tin show a staggered behaviour around A = 131 , not observed in the ENDF/B-VII.1 evaluation. The yields of antimony also contradict the trend from the evaluation, but are in agreement with a calculation performed using the GEF model that shows the yield increasing with mass in the range A = 128-133.

The Prompt Fission Neutron Spectrum of 235U(n,f) below 2.5 MeV for Incident Neutrons from 0.7 to 20 MeV

DOE PAGES

Devlin, Matthew James; Gomez, Jaime A.; Kelly, Keegan John; ...

2018-02-01

New prompt fission neutron spectrum measurements are reported for 235U(n,f) reactions induced by neutrons with energies from 0.7 to 20 MeV. These measurements cover outgoing neutron energies from 2.5 MeV down to 10 keV, using an array of 6Li-glass scintillators for neutron detection and a double time-of-flight technique. The neutrons were produced at the Weapons Neutron Research facility of the Los Alamos Neutron Science Center. A detailed MCNP ® model of the experimental equipment and the surrounding room was used to interpret the experimental results. Backgrounds were measured in situ, making use of the time-dependent singles rates of the variousmore » detectors with asynchronous readout from waveform digitizers. The results presented here have been included in a re-evaluation of the fission neutron spectra for this fissioning system, a description of which is presented elsewhere in this issue.« less

FALSTAFF: A new tool for fission studies

NASA Astrophysics Data System (ADS)

Dore, D.; Farget, F.; Lecolley, F.-R.; Lehaut, G.; Materna, T.; Pancin, J.; Panebianco, S.; Papaevangelou, Th.

2013-12-01

The future NFS installation will produce high intensity neutron beams from hundreds of keV up to 40 MeV. Taking advantage of this facility, data of particular interest for the nuclear community in view of the development of the fast reactor technology will be measured. The development of an experimental setup called FALSTAFF for a full characterization of actinide fission fragments has been undertaken. Fission fragment isotopic yields and associated neutron multiplicities will be measured as a function of the neutron energy. Based on time-of-flight and residual energy technique, the setup will allow the simultaneous measurement of the complementary fragments velocity and energy. The performances of TOF detectors of FALSTAFF will be presented and expected resolutions for fragment masses and neutron multiplicities, based on realistic simulations, will be shown.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cassata, W. S.; Velsko, C. A.; Stoeffl, W.

We determined fission yields of xenon ( 133mXe, 135Xe, 135mXe, 137Xe, 138Xe, and 139Xe) resulting from 14 MeV neutron induced fission of depleted uranium at the National Ignition Facility. Measurements begin approximately 20 s after shot time, and yields have been determined for nuclides with half-lives as short as tens of seconds. We determined the relative independent yields of 133mXe, 135Xe, and 135mXe to significantly higher precision than previously reported. The relative fission yields of all nuclides are statistically indistinguishable from values reported by England and Rider (ENDF-349. LA-UR-94-3106, 1994), with exception of the cumulative yield of 139Xe. Furthermore, considerablemore » differences exist between our measured yields and the JEFF-3.1 database values.« less

Plasma core reactor simulations using RF uranium seeded argon discharges

NASA Technical Reports Server (NTRS)

Roman, W. C.

1975-01-01

An experimental investigation was conducted using the United Technologies Research Center (UTRC) 80 kW and 1.2 MW RF induction heater systems to aid in developing the technology necessary for designing a self-critical fissioning uranium plasma core reactor (PCR). A nonfissioning, steady-state RF-heated argon plasma seeded with pure uranium hexafluoride (UF6) was used. An overall objective was to achieve maximum confinement of uranium vapor within the plasma while simultaneously minimizing the uranium compound wall deposition. Exploratory tests were conducted using the 80 kW RF induction heater with the test chamber at approximately atmospheric pressure and discharge power levels on the order of 10 kW. Four different test chamber flow configurations were tested to permit selection of the configuration offering the best confinement characteristics for subsequent tests at higher pressure and power in the 1.2 MW RF induction heater facility.

Performance Testing of a Liquid Metal Pump for In-Space Power Systems

NASA Technical Reports Server (NTRS)

Polzin, Kurt

2011-01-01

Fission surface power (FSP) systems could be used to provide power on the surface of the moon, Mars, or other planets and moons of our solar system. Fission power systems could provide excellent performance at any location, including those near the poles or other permanently shaded regions, and offer the capability to provide on demand power at any time, even at large distances from the sun. Fission-based systems also offer the potential for outposts, crew and science instruments to operate in a power-rich environment. NASA has been exploring technologies with the goal of reducing the cost and technical risk of employing FSP systems. A reference 40 kWe option has been devised that is cost-competitive with alternatives while providing more power for less mass anywhere on the lunar surface. The reference FSP system is also readily extensible for use on Mars, where it would be capable of operating through global dust storms and providing year-round power at any Martian latitude. Detailed development of the FSP concept and the reference mission are documented in various other reports. The development discussed in this paper prepares the way for testing of the Technology Demonstration Unit (TDU), which is a 10 kWe end-to-end test of FSP technologies intended to raise the entire FSP system to technology readiness level (TRL) 6. The Early Flight Fission Test Facility (EFF-TF) was established by NASA s Marshall Space Flight Center (MSFC) to provide a capability for performing hardware-directed activities to support multiple in-space nuclear reactor concepts by using a nonnuclear test methodology. This includes fabrication and testing at both the module/component level and at near prototypic reactor components and configurations allowing for realistic thermal-hydraulic evaluations of systems. The liquid-metal pump associated with the FSP system must be compatible with the liquid NaK coolant and have adequate performance to enable a viable flight system. Idaho National Laboratory (INL) was tasked with the modeling, design, and fabrication of an ALIP suitable for the FSP reference mission. A prototypic ALIP was fabricated under the direction of INL and shipped to MSFC for inclusion in the Technology Demonstration Unit (TDU), a quarter-scale end-to-end reactor simulator system that is scheduled for testing at NASA-GRC. Before inclusion in the TDU, the ALIP was tested in the ALIP test circuit (ATC), which is a rig developed and operated at MSFC for the specific purpose of providing accurate quantification of liquid metal pump performance. Data showing the pump performance curves (pressure, flowrate, and pump efficiency) are presented for various operating power levels, demonstrating the full performance envelope of the pump.

Yields of short-lived fission products produced following {sup 235}U(n{sub th},f)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tipnis, S.V.; Campbell, J.M.; Couchell, G.P.

1998-08-01

Measurements of gamma-ray spectra, following the thermal neutron fission of {sup 235}U have been made using a high purity germanium detector at the University of Massachusetts Lowell (UML) Van de Graaff facility. The gamma spectra were measured at delay times ranging from 0.2 s to nearly 10thinsp000 s following the rapid transfer of the fission fragments with a helium-jet system. On the basis of the known gamma transitions, forty isotopes have been identified and studied. By measuring the relative intensities of these transitions, the relative yields of the various precursor nuclides have been calculated. The results are compared with themore » recommended values listed in the ENDF/B-VI fission product data base (for the lifetimes and the relative yields) and those published in the Nuclear Data Sheets (for the beta branching ratios). This information is particularly useful for the cases of short-lived fission products with lifetimes of the order of fractions of a second or a few seconds. Independent yields of many of these isotopes have rather large uncertainties, some of which have been reduced by the present study. {copyright} {ital 1998} {ital The American Physical Society}« less

Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

NASA Astrophysics Data System (ADS)

Tsibulskiy, V. F.; Andrianova, E. A.; Davidenko, V. D.; Rodionova, E. V.; Tsibulskiy, S. V.

2017-12-01

A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium-plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

Comparison of the recently proposed super-Marx generator approach to thermonuclear ignition with the deuterium-tritium laser fusion-fission hybrid concept by the Lawrence Livermore National Laboratory

DOE PAGES

Winterberg, F.

2009-01-01

The recently proposed super-Marx generator pure deuterium microdetonation ignition concept is compared to the Lawrence Livermore National Ignition Facility (NIF) Laser deuterium-tritium fusion-fission hybrid concept (LIFE). In a super-Marx generator, a large number of ordinary Marx generators charge up a much larger second stage ultrahigh voltage Marx generator from which for the ignition of a pure deuterium microexplosion an intense GeV ion beam can be extracted. Typical examples of the LIFE concept are a fusion gain of 30 and a fission gain of 10, making up a total gain of 300, with about ten times more energy released into fissionmore » as compared to fusion. This means the substantial release of fission products, as in fissionless pure fission reactors. In the super-Marx approach for the ignition of pure deuterium microdetonation, a gain of the same magnitude can, in theory, be reached. If feasible, the super-Marx generator deuterium ignition approach would make lasers obsolete as a means for the ignition of thermonuclear microexplosions.« less

Neutron-induced fission cross section measurement of 233U, 241Am and 243Am in the energy range 0.5 MeV En 20 MeV at n TOF at CERN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belloni, F.; Milazzo, P. M.; Calviani, M.

2012-01-01

Neutron-induced fission cross section measurements of 233U, 243Am and 241Am relative to 235U have been carried out at the neutron time-of-flight facility n TOF at CERN. A fast ionization chamber has been employed. All samples were located in the same detector; therefore the studied elements and the reference 235U target are subject to the same neutron beam.

FY-2015 Methyl Iodide Deep-Bed Adsorption Test Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soelberg, Nicholas Ray; Watson, Tony Leroy

2015-09-30

Nuclear fission produces fission and activation products, including iodine-129, which could evolve into used fuel reprocessing facility off-gas systems, and could require off-gas control to limit air emissions to levels within acceptable emission limits. Deep-bed methyl iodide adsorption testing has continued in Fiscal Year 2015 according to a multi-laboratory methyl iodide adsorption test plan. Updates to the deep-bed test system have also been performed to enable the inclusion of evaporated HNO 3 and increased NO 2 concentrations in future tests. This report summarizes the result of those activities. Test results showed that iodine adsorption from gaseous methyl iodide using reducedmore » silver zeolite (AgZ) resulted in initial iodine decontamination factors (DFs, ratios of uncontrolled and controlled total iodine levels) under 1,000 for the conditions of the long-duration test performed this year (45 ppm CH3I, 1,000 ppm each NO and NO 2, very low H 2O levels [3 ppm] in balance air). The mass transfer zone depth exceeded the cumulative 5-inch depth of 4 bed segments, which is deeper than the 2-4 inch depth estimated for the mass transfer zone for adsorbing I 2 using AgZ in prior deep-bed tests. The maximum iodine adsorption capacity for the AgZ under the conditions of this test was 6.2% (6.2 g adsorbed I per 100 g sorbent). The maximum Ag utilization was 51%. Additional deep-bed testing and analyses are recommended to (a) expand the data base for methyl iodide adsorption and (b) provide more data for evaluating organic iodide reactions and reaction byproducts for different potential adsorption conditions.« less

Fission Product Yield Study of 235U, 238U and 239Pu Using Dual-Fission Ionization Chambers

NASA Astrophysics Data System (ADS)

Bhatia, C.; Fallin, B.; Howell, C.; Tornow, W.; Gooden, M.; Kelley, J.; Arnold, C.; Bond, E.; Bredeweg, T.; Fowler, M.; Moody, W.; Rundberg, R.; Rusev, G.; Vieira, D.; Wilhelmy, J.; Becker, J.; Macri, R.; Ryan, C.; Sheets, S.; Stoyer, M.; Tonchev, A.

2014-05-01

To resolve long-standing differences between LANL and LLNL regarding the correct fission basis for analysis of nuclear test data [M.B. Chadwick et al., Nucl. Data Sheets 111, 2891 (2010); H. Selby et al., Nucl. Data Sheets 111, 2891 (2010)], a collaboration between TUNL/LANL/LLNL has been established to perform high-precision measurements of neutron induced fission product yields. The main goal is to make a definitive statement about the energy dependence of the fission yields to an accuracy better than 2-3% between 1 and 15 MeV, where experimental data are very scarce. At TUNL, we have completed the design, fabrication and testing of three dual-fission chambers dedicated to 235U, 238U, and 239Pu. The dual-fission chambers were used to make measurements of the fission product activity relative to the total fission rate, as well as for high-precision absolute fission yield measurements. The activation method was employed, utilizing the mono-energetic neutron beams available at TUNL. Neutrons of 4.6, 9.0, and 14.5 MeV were produced via the 2H(d,n)3He reaction, and for neutrons at 14.8 MeV, the 3H(d,n)4He reaction was used. After activation, the induced γ-ray activity of the fission products was measured for two months using high-resolution HPGe detectors in a low-background environment. Results for the yield of seven fission fragments of 235U, 238U, and 239Pu and a comparison to available data at other energies are reported. For the first time results are available for neutron energies between 2 and 14 MeV.

Development status of the heatpipe power and bimodal systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poston, David I.; Houts, Michael G.

1999-01-01

Space fission power systems can potentially enhance or enable ambitious lunar and Martian surface missions. Research into space fission power systems has been ongoing (at various levels) since the 1950s, but to date the United States (US) has flown only one space fission system, SNAP-10A, in 1965. Cost and development time have been significant reasons why space fission systems have not been used by the US. High cost and long development time are not inherent to the use of space fission power. However, high cost and long development time are inherent to any program that tries to do too muchmore » at once. Nearly all US space fission power programs have attempted to field systems capable of high power, even though more modest systems had not yet been flown. All of these programs have failed to fly a space fission system. Relatively low power (10 to 100 kWe) fission systems may be useful for near-term lunar and Martian surface missions, including missions in which in situ resource utilization is a priority. Such systems may also be useful for deep-space science missions and other missions. These systems can be significantly less expensive to develop than high power systems. Experience gained in the development of low-power space fission systems can then be used to enable cost-effective development of high-power ({gt}1000 kWe) fission systems. The Heatpipe Power System (HPS) is a potential, near-term, low-cost space fission power system. The Heatpipe Bimodal System (HBS) is a potential, near-term, low-cost space fission power and/or propulsion system. Both systems will be composed of independent modules, and all components use existing technology and operate within the existing database. The HPS and HBS have relatively few system integration issues; thus, the successful development of a module is a significant step toward verifying system feasibility and performance estimates. A prototypic HPS module was fabricated, and initial testing was completed in April 1997. All test objectives were accomplished, demonstrating the basic feasibility of the HPS. Fabrication of an HBS module is under way, and testing should begin in 1999. {copyright} {ital 1999 American Institute of Physics.}« less

Development status of the heatpipe power and bimodal systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poston, David I.; Houts, Michael G.; Emrich, William J. Jr.

1999-01-22

Space fission power systems can potentially enhance or enable ambitious lunar and Martian surface missions. Research into space fission power systems has been ongoing (at various levels) since the 1950s, but to date the United States (US) has flown only one space fission system, SNAP-10A, in 1965. Cost and development time have been significant reasons why space fission systems have not been used by the US. High cost and long development time are not inherent to the use of space fission power. However, high cost and long development time are inherent to any program that tries to do too muchmore » at once. Nearly all US space fission power programs have attempted to field systems capable of high power, even though more modest systems had not yet been flown. All of these programs have failed to fly a space fission system. Relatively low power (10 to 100 kWe) fission systems may be useful for near-term lunar and Martian surface missions, including missions in which in situ resource utilization is a priority. Such systems may also be useful for deep-space science missions and other missions. These systems can be significantly less expensive to develop than high power systems. Experience gained in the development of low-power space fission systems can then be used to enable cost-effective development of high-power (>1000 kWe) fission systems. The Heatpipe Power System (HPS) is a potential, near-term, low-cost space fission power system. The Heatpipe Bimodal System (HBS) is a potential, near-term, low-cost space fission power and/or propulsion system. Both systems will be composed of independent modules, and all components use existing technology and operate within the existing database. The HPS and HBS have relatively few system integration issues; thus, the successful development of a module is a significant step toward verifying system feasibility and performance estimates. A prototypic HPS module was fabricated, and initial testing was completed in April 1997. All test objectives were accomplished, demonstrating the basic feasibility of the HPS. Fabrication of an HBS module is under way, and testing should begin in 1999.« less

Development status of the heatpipe power and bimodal systems

NASA Astrophysics Data System (ADS)

Poston, David I.; Houts, Michael G.; Emrich, William J.

1999-01-01

Space fission power systems can potentially enhance or enable ambitious lunar and Martian surface missions. Research into space fission power systems has been ongoing (at various levels) since the 1950s, but to date the United States (US) has flown only one space fission system, SNAP-10A, in 1965. Cost and development time have been significant reasons why space fission systems have not been used by the US. High cost and long development time are not inherent to the use of space fission power. However, high cost and long development time are inherent to any program that tries to do too much at once. Nearly all US space fission power programs have attempted to field systems capable of high power, even though more modest systems had not yet been flown. All of these programs have failed to fly a space fission system. Relatively low power (10 to 100 kWe) fission systems may be useful for near-term lunar and Martian surface missions, including missions in which in situ resource utilization is a priority. Such systems may also be useful for deep-space science missions and other missions. These systems can be significantly less expensive to develop than high power systems. Experience gained in the development of low-power space fission systems can then be used to enable cost-effective development of high-power (>1000 kWe) fission systems. The Heatpipe Power System (HPS) is a potential, near-term, low-cost space fission power system. The Heatpipe Bimodal System (HBS) is a potential, near-term, low-cost space fission power and/or propulsion system. Both systems will be composed of independent modules, and all components use existing technology and operate within the existing database. The HPS and HBS have relatively few system integration issues; thus, the successful development of a module is a significant step toward verifying system feasibility and performance estimates. A prototypic HPS module was fabricated, and initial testing was completed in April 1997. All test objectives were accomplished, demonstrating the basic feasibility of the HPS. Fabrication of an HBS module is under way, and testing should begin in 1999.

Fission Surface Power Systems (FSPS) Project Final Report for the Exploration Technology Development Program (ETDP): Fission Surface Power, Transition Face to Face

NASA Technical Reports Server (NTRS)

Palac, Donald T.

2011-01-01

The Fission Surface Power Systems Project became part of the ETDP on October 1, 2008. Its goal was to demonstrate fission power system technology readiness in an operationally relevant environment, while providing data on fission system characteristics pertinent to the use of a fission power system on planetary surfaces. During fiscal years 08 to 10, the FSPS project activities were dominated by hardware demonstrations of component technologies, to verify their readiness for inclusion in the fission surface power system. These Pathfinders demonstrated multi-kWe Stirling power conversion operating with heat delivered via liquid metal NaK, composite Ti/H2O heat pipe radiator panel operations at 400 K input water temperature, no-moving-part electromagnetic liquid metal pump operation with NaK at flight-like temperatures, and subscale performance of an electric resistance reactor simulator capable of reproducing characteristics of a nuclear reactor for the purpose of system-level testing, and a longer list of component technologies included in the attached report. Based on the successful conclusion of Pathfinder testing, work began in 2010 on design and development of the Technology Demonstration Unit (TDU), a full-scale 1/4 power system-level non-nuclear assembly of a reactor simulator, power conversion, heat rejection, instrumentation and controls, and power management and distribution. The TDU will be developed and fabricated during fiscal years 11 and 12, culminating in initial testing with water cooling replacing the heat rejection system in 2012, and complete testing of the full TDU by the end of 2014. Due to its importance for Mars exploration, potential applicability to missions preceding Mars missions, and readiness for an early system-level demonstration, the Enabling Technology Development and Demonstration program is currently planning to continue the project as the Fission Power Systems project, including emphasis on the TDU completion and testing.

Overview of Non-nuclear Testing of the Safe, Affordable 30-kW Fission Engine, Including End-to-End Demonstrator Testing

NASA Technical Reports Server (NTRS)

VanDyke, M. K.; Martin, J. J.; Houts, M. G.

2003-01-01

Successful development of space fission systems will require an extensive program of affordable and realistic testing. In addition to tests related to design/development of the fission system, realistic testing of the actual flight unit must also be performed. At the power levels under consideration (3-300 kW electric power), almost all technical issues are thermal or stress related and will not be strongly affected by the radiation environment. These issues can be resolved more thoroughly, less expensively, and in a more timely fashing with nonnuclear testing, provided it is prototypic of the system in question. This approach was used for the safe, affordable fission engine test article development program and accomplished viz cooperative efforts with Department of Energy labs, industry, universiites, and other NASA centers. This Technical Memorandum covers the analysis, testing, and data reduction of a 30-kW simulated reactor as well as an end-to-end demonstrator, including a power conversion system and an electric propulsion engine, the first of its kind in the United States.

Diffusive deposition of aerosols in Phebus containment during FPT-2 test

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontautas, A.; Urbonavicius, E.

2012-07-01

At present the lumped-parameter codes is the main tool to investigate the complex response of the containment of Nuclear Power Plant in case of an accident. Continuous development and validation of the codes is required to perform realistic investigation of the processes that determine the possible source term of radioactive products to the environment. Validation of the codes is based on the comparison of the calculated results with the measurements performed in experimental facilities. The most extensive experimental program to investigate fission product release from the molten fuel, transport through the cooling circuit and deposition in the containment is performedmore » in PHEBUS test facility. Test FPT-2 performed in this facility is considered for analysis of processes taking place in containment. Earlier performed investigations using COCOSYS code showed that the code could be successfully used for analysis of thermal-hydraulic processes and deposition of aerosols, but there was also noticed that diffusive deposition on the vertical walls does not fit well with the measured results. In the CPA module of ASTEC code there is implemented different model for diffusive deposition, therefore the PHEBUS containment model was transferred from COCOSYS code to ASTEC-CPA to investigate the influence of the diffusive deposition modelling. Analysis was performed using PHEBUS containment model of 16 nodes. The calculated thermal-hydraulic parameters are in good agreement with measured results, which gives basis for realistic simulation of aerosol transport and deposition processes. Performed investigations showed that diffusive deposition model has influence on the aerosol deposition distribution on different surfaces in the test facility. (authors)« less

Method of fission heat flux determination from experimental data

DOEpatents

Paxton, Frank A.

1999-01-01

A method is provided for determining the fission heat flux of a prime specimen inserted into a specimen of a test reactor. A pair of thermocouple test specimens are positioned at the same level in the holder and a determination is made of various experimental data including the temperature of the thermocouple test specimens, the temperature of bulk water channels located in the test holder, the gamma scan count ratios for the thermocouple test specimens and the prime specimen, and the thicknesses of the outer clads, the fuel fillers, and the backclad of the thermocouple test specimen. Using this experimental data, the absolute value of the fission heat flux for the thermocouple test specimens and prime specimen can be calculated.

Phase 2 Methyl Iodide Deep-Bed Adsorption Tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soelberg, Nick; Watson, Tony

2014-09-01

Nuclear fission produces fission products (FPs) and activation products, including iodine-129, which could evolve into used fuel reprocessing facility off-gas systems, and could require off-gas control to limit air emissions to levels within acceptable emission limits. Research, demonstrations, and some reprocessing plant experience have indicated that diatomic iodine can be captured with efficiencies high enough to meet regulatory requirements. Research on the capture of organic iodides has also been performed, but to a lesser extent. Several questions remain open regarding the capture of iodine bound in organic compounds. Deep-bed methyl iodide adsorption testing has progressed according to a multi-laboratory methylmore » iodide adsorption test plan. This report summarizes the second phase of methyl iodide adsorption work performed according to this test plan using the deep-bed iodine adsorption test system at the Idaho National Laboratory (INL), performed during the second half of Fiscal Year (FY) 2014. Test results continue to show that methyl iodide adsorption using AgZ can achieve total iodine decontamination factors (DFs, ratios of uncontrolled and controlled total iodine levels) above 1,000, until breakthrough occurred. However, mass transfer zone depths are deeper for methyl iodide adsorption compared to diatomic iodine (I2) adsorption. Methyl iodide DFs for the Ag Aerogel test adsorption efficiencies were less than 1,000, and the methyl iodide mass transfer zone depth exceeded 8 inches. Additional deep-bed testing and analyses are recommended to (a) expand the data base for methyl iodide adsorption under various conditions specified in the methyl iodide test plan, and (b) provide more data for evaluating organic iodide reactions and reaction byproducts for different potential adsorption conditions.« less

Atmospheric Transport Modelling confining potential source location of East-Asian radionuclide detections in May 2010

NASA Astrophysics Data System (ADS)

Ross, J. Ole; Ceranna, Lars

2016-04-01

The radionuclide component of the International Monitoring System (IMS) to verify compliance with the Comprehensive Nuclear-Test-Ban Treaty (CTBT) is in place to detect tiny traces of fission products from nuclear explosions in the atmosphere. The challenge for the interpretation of IMS radionuclide data is to discriminate radionuclide sources of CTBT relevance against emissions from nuclear facilities. Remarkable activity concentrations of Ba/La-140 occurred at the IMS radionuclide stations RN 37 (Okinawa) and RN 58 (Ussurysk) mid of May 2010. In those days also an elevated Xe-133 level was measured at RN 38 (Takasaki). Additional regional measurements of radioxenon were reported in the press and further analyzed in various publications. The radionuclide analysis gives evidence for the presence of a nuclear fission source between 10 and 12 May 2010. Backward Atmospheric Transport Modelling (ATM) with HYSPLIT driven by 0.2° ECMWF meteorological data for the IMS samples indicates that, assuming a single source, a wide range of source regions is possible including the Korean Peninsula, the Sea of Japan (East Sea), and parts of China and Russia. Further confinement of the possible source location can be provided by atmospheric backtracking for the assumed sampling periods of the reported regional xenon measurements. New studies indicate a very weak seismic event at the DPRK test site on early 12 May 2010. Forward ATM for a pulse release caused by this event shows fairly good agreement with the observed radionuclide signature. Nevertheless, the underlying nuclear fission scenario remains quite unclear and speculative even if assuming a connection between the waveform and the radionuclide event.

Measurement of Fission Product Yields from Fast-Neutron Fission

NASA Astrophysics Data System (ADS)

Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Henderson, R.; Kenneally, J.; Macri, R.; McNabb, D.; Ryan, C.; Sheets, S.; Stoyer, M. A.; Tonchev, A. P.; Bhatia, C.; Bhike, M.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.

2014-09-01

One of the aims of the Stockpile Stewardship Program is a reduction of the uncertainties on fission data used for analyzing nuclear test data [1,2]. Fission products such as 147Nd are convenient for determining fission yields because of their relatively high yield per fission (about 2%) and long half-life (10.98 days). A scientific program for measuring fission product yields from 235U,238U and 239Pu targets as a function of bombarding neutron energy (0.1 to 15 MeV) is currently underway using monoenergetic neutron beams produced at the 10 MV Tandem Accelerator at TUNL. Dual-fission chambers are used to determine the rate of fission in targets during activation. Activated targets are counted in highly shielded HPGe detectors over a period of several weeks to identify decaying fission products. To date, data have been collected at neutron bombarding energies 4.6, 9.0, 14.5 and 14.8 MeV. Experimental methods and data reduction techniques are discussed, and some preliminary results are presented.

The SPIDER fission fragment spectrometer for fission product yield measurements

DOE PAGES

Meierbachtol, K.; Tovesson, F.; Shields, D.; ...

2015-04-01

We developed the SPectrometer for Ion DEtermination in fission Research (SPIDER) for measuring mass yield distributions of fission products from spontaneous and neutron-induced fission. The 2E–2v method of measuring the kinetic energy (E) and velocity (v) of both outgoing fission products has been utilized, with the goal of measuring the mass of the fission products with an average resolution of 1 atomic mass unit (amu). Moreover, the SPIDER instrument, consisting of detector components for time-of-flight, trajectory, and energy measurements, has been assembled and tested using 229Th and 252Cf radioactive decay sources. For commissioning, the fully assembled system measured fission productsmore » from spontaneous fission of 252Cf. Individual measurement resolutions were met for time-of-flight (250 ps FWHM), spacial resolution (2 mm FHWM), and energy (92 keV FWHM for 8.376 MeV). Finally, these mass yield results measured from 252Cf spontaneous fission products are reported from an E–v measurement.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harker, Y.D.

On August 3-4, 1994, an INEL team made measurements related to a real-time monitoring system for use on the epithermal beam facility at the BMRR. BNL has installed two fission chambers in front of the beam collimator, which are to monitor the beam coming from the reactor. These two monitors are located with one just above the 16-cm dia. front aperture and the other is just below. The fission chambers contain depleted uranium, but because of the small amount of U-235 present, they respond to thermal and near thermal neutrons rather than fast neutrons. This feature combined with their relativelymore » small size (0.6 cm dia x 4 cm long) makes them very good monitors in the BMRR epithermal neutron beam. The INEL team worked with H.B. Lui (BNL) in performing initial tests of these monitors and established the settings to achieve stable operation. The main purpose of the measurement studies was to establish a basis for a monitoring method that tracks the dose the patient is receiving rather than the neutron fluence being delivered down the beam line.« less

Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Gibson, Marc A.; Sanzi, James

2017-01-01

The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one to be used at the Device Assembly Facility (DAF) at the Nevada National Security Site. Fabrication, assembly, and testing of the depleted uranium core has allowed for higher fidelity system level testing at GRC, and has validated the fabrication methods to be used on the highly enriched uranium core that will supply heat for the DAF KRUSTY demonstration.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Emrich, William J. Jr.

2008-01-21

To support a potential future development of a nuclear thermal rocket engine, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components could be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowingmore » hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts.« less

Neutron-rich isotope production using a uranium carbide - carbon nanotubes SPES target prototype

NASA Astrophysics Data System (ADS)

Corradetti, S.; Biasetto, L.; Manzolaro, M.; Scarpa, D.; Carturan, S.; Andrighetto, A.; Prete, G.; Vasquez, J.; Zanonato, P.; Colombo, P.; Jost, C. U.; Stracener, D. W.

2013-05-01

The SPES (Selective Production of Exotic Species) project, under development at the Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali di Legnaro (INFN-LNL), is a new-generation Isotope Separation On-Line (ISOL) facility for the production of radioactive ion beams by means of the proton-induced fission of uranium. In the framework of the research on the SPES target, seven uranium carbide discs, obtained by reacting uranium oxide with graphite and carbon nanotubes, were irradiated with protons at the Holifield Radioactive Ion Beam Facility (HRIBF) of Oak Ridge National Laboratory (ORNL). In the following, the yields of several fission products obtained during the experiment are presented and discussed. The experimental results are then compared to those obtained using a standard uranium carbide target. The reported data highlights the capability of the new type of SPES target to produce and release isotopes of interest for the nuclear physics community.

Fission products and nuclear fuel behaviour under severe accident conditions part 2: Fuel behaviour in the VERDON-1 sample

NASA Astrophysics Data System (ADS)

Geiger, E.; Le Gall, C.; Gallais-During, A.; Pontillon, Y.; Lamontagne, J.; Hanus, E.; Ducros, G.

2017-11-01

Within the framework of the International Source Term Programme (ISTP), the VERDON programme aims at quantifying the source term of radioactive materials in case of a hypothetical severe accident in a light water reactor (LWR). Tests were performed in a new experimental laboratory (VERDON) built in the LECA-STAR facility (CEA Cadarache). The VERDON-1 test was devoted to the study of a high burn-up UO2 fuel and FP releases at very high temperature (≈2873 K) in a reducing atmosphere. Post-test qualitative and quantitative characterisations of the VERDON-1 sample led to the proposal of a scenario explaining the phenomena occurring during the experimental sequence. Hence, the fuel and the cladding may have interacted which led to the melting of UO2-ZrO2 alloy. Although no relocation was observed during the test, it may have been imminent.

Fission and Properties of Neutron-Rich Nuclei

NASA Astrophysics Data System (ADS)

Hamilton, Joseph H.; Ramayya, A. V.; Carter, H. K.

2008-08-01

Opening session. Nuclear processes in stellar explosions / M. Wiescher. In-beam [symbol]-ray spectroscopy of neutron-rich nuclei at NSCL / A. Gade -- Nuclear structure I. Shell-model structure of neutron-rich nuclei beyond [symbol]Sn / A. Covello ... [et al.]. Shell structure and evolution of collectivity in nuclei above the [symbol]Sn core / S. Sarkar and M. S. Sarkar. Heavy-ion fusion using density-constrained TDHF / A. S. Umar and V. E. Oberacker. Towards an extended microscopic theory for upper-fp shell nuclei / K. P. Drumev. Properties of the Zr and Pb isotopes near the drip-line / V. N. Tarasov ... [et al.]. Identification of high spin states in [symbol] Cs nuclei and shell model calculations / K. Li ... [et al.]. Recent measurements of spherical and deformed isomers using the Lohengrin fission-fragment spectrometer / G. S. Simpson ... [et al.] -- Nuclear structure II. Nuclear structure investigation with rare isotope spectroscopic investigations at GSI / P. Boutachkov. Exploring the evolution of the shell structures by means of deep inelastic reactions / G. de Anaelis. Probing shell closures in neutron-rich nuclei / R. Krücken for the S277 and REX-ISOLDEMINIBALL collaborations. Structure of Fe isotopes at the limits of the pf-shell / N. Hoteling ... [et al.]. Spectroscopy of K isomers in shell-stabilized trans-fermium nuclei / S. K. Tandel ... [et al.] -- Radioactive ion beam facilities. SPIRAL2 at GANIL: a world leading ISOL facility for the next decade / S. Gales. New physics at the International Facility for Antiproton and Ion Research (FAIR) next to GSI / I. Augustin ... [et al.]. Radioactive beams from a high powered ISOL system / A. C. Shotter. RlKEN RT beam factory / T. Motobayashi. NSCL - ongoing activities and future perspectives / C. K. Gelbke. Rare isotope beams at Argonne / W. F. Henning. HRIBF: scientific highlights and future prospects / J. R. Beene. Radioactive ion beam research done in Dubna / G. M. Ter-Akopian ... [et al.] -- Fission I. Fission-fragment spectroscopy with STEFF / A. G. Smith ... [et al.]. Gamma ray multiplicity of [symbol]Cf spontaneous fission using LiBerACE / D. L. Bleuel ... [et al.]. Excitation energy dependence of fragment mass and total kinetic energy distributions in proton-induced fission of light actinides / I. Nishinaka ... [et al.]. A dynamical calculation of multi-modal nuclear fission / T. Wada and T. Asano. Structure of fission potential energy surfaces in ten-dimensional spaces / V. V. Pashkevich, Y. K Pyatkov and A. V. Unzhakova. A possible enhancement of nuclear fission in scattering with low energy charged particles / V. Gudkov. Dynamical multi-break processes in the [symbol]Sn + [symbol]Ni system at 35 MeV/Nucleon / M. Papa and ISOSPIN-RE VERSE collaboration -- New experimental techniques. MTOF - a high resolution isobar separator for studies of exotic decays / A. Piechaczek ... [et al.]. Development of ORRUBA: a silicon array for the measurement of transfer reactions in inverse kinematics / S. D. Pain ... [et al.]. Indian national gamma array: present & future / R. K. Bhowmik. Absolute intensities of [symbol] rays emitted in the decay of [symbol]U / H. C. Griffin -- Superheavy elements theory and experiments / M. G. Itkis ... [et al.]. Study of superheavy elements at SHIP / S. Hofinann. Heaviest nuclei from [symbol]Ca-induced reactions / Yu. Ts. Oaanessian. Superheavy nuclei and giant nuclear systems / W. Greiner and V. Zagrebaev. Fission approach to alpha-decay of superheavy nuclei / D.N. Poenaru and W. Greiner. Superheavy elements in the Magic Islands / C. Samanta. Relativistic mean field studies of superheavy nuclei / A. V. Afanas jev. Understanding the synthesis of the heaviest nuclei / W. Loveland -- Mass measurements and g-factors. G factor measurements in neutron-rich [symbol]Cf fission fragments, measured using the gammasphere array / R. Orlandi ... [et al.]. Technique for measuring angular correlations and g-factors in neutron rich nuclei produced by the spontaneous fission of [symbol]Cf / A. V. Daniel ... [et al.]. Magnetic moment measurements in a radioactive beam environment / N. Benczer-Koller and G. Kumbartzki. g-Factor measurements of picosecond states: opportunities and limitations of the recoil-in-vacuum method / N. J. Stone ... [et al.]. Precision mass measurements and trap-assisted spectroscopy of fission products from Ni to Pd / A. Jokinen -- Fission II. Fission research at IRMM / F.-J. Hambsch. Fission yield measurements at the IGISOL facility with JYFLTRAP / H. Penttilä ... [et al.]. Fission of radioactive beams and dissipation in nuclear matter / A. Heinz (for the CHARMS collaboration). Fission of [symbol]U at 80 MeVlu and search for new neutron-rich isotopes / C.M. Folden, III ... [et al.]. Measurement of the average energy and multiplicity of prompt-fission neutrons and gamma rays from [symbol], [symbol], and [symbol] for incident neutron energies of 1 to 200 MeV / R. C. Haight ... [et al.]. Fission measurements with DANCE / M. Jandel ... [et al.]. Measured and calculated neutron-induced fission cross sections of [symbol]Pu / F. Tovesson and T. S. Hill. The fission barrier landscape / L. Phair and L. G. Moretto. Fast neutron-induced fission of some actinides and sub-actinides / A. B. Lautev ... [et al.] -- Fission III/Nuclear structure III. Complex structure in even-odd staggering of fission fragment yields / M. Caamāno and F. Rejmund. The surrogate method: past, present and future / S. R. Lesher ... [et al]. Effects of nuclear incompressibility on heavy-ion fusion / H. Esbensen and Ş. Mişicu. High spin states in [symbol]Pm / A. Dhal ... [et al]. Structure of [symbol]Sm, spherical vibrator versus softly deformed rotor / J. B. Gupta -- Astrophysics. Measuring the astrophysical S-factor in plasmas / A. Bonasera ... [et al.]. Is there shell quenching or shape coexistence in Cd isotopes near N = 82? / J. K. Hwang, A. V. Ramayya and J. H. Hamilton. Spectroscopy of neutron-rich palladium and cadmium isostopes near A= 120 / M. A. Stoyer and W. B. Walters -- Nuclear structure IV. First observation of new neutron-rich magnesium, aluminum and silicon isotopes / A. Stolz ... [et al.]. Spectroscopy of [symbol]Na revolution of shell structure with isospin / V. Tripathi ... [et al.]. Rearrangement of proton single particle orbitals in neutron-rich potassium isotopes - spectroscopy of [symbol]K / W. Królas ... [et al.]. Laser spectroscopy and the nature of the shape transition at N [symbol] 60 / B. Cheal ... [et al.]. Study of nuclei near stability as fission fragments following heavy-ion reactions / N. Fotiadis. [symbol]C and [symbol]N: lifetime measurements of their first-excited states / M. Wiedeking ... [et al.] -- Nuclear astrophysics. Isomer spectroscopy near [symbol]Sn - first observation of excited states in [symbol]Cd / M. Pfitzner ... [et al.]. Nuclear masses and what they imply for the structures of neutron rich nuclei / A. Awahamian and A. Teymurazyan. Multiple nucleosynthesis processes in the early universe / F. Montes. Single-neutron structure of neutron-rich nuclei near N = 50 and N = 82 / J. A. Cizewski ... [et al.]. [symbol]Cadmium: ugly duckling or young swan / W. B. Walters ... [et al.] -- Nuclear structure V. Evidence for chiral doublet bands in [symbol]Ru / Y. X. Luo ... [et al.]. Unusual octupole shape deformation terms and K-mixing / J. O. Rasmussen ... [et al.]. Spin assignments, mixing ratios, and g-factors in neutron rich [symbol]Cf fission products / C. Goodin ... [et al.]. Level structures and double [symbol]-bands in [symbol]Mo, [symbol]Mo and [symbol]Ru / S. J. Zhu ... [et al.] -- Nuclear theory. Microscopic dynamics of shape coexistence phenomena around [symbol]Se and [symbol]Kr / N. Hinohara ... [et al.]. Nuclear structure, double beta decay and test of physics beyond the standard model / A. Faessler. Collective modes in elastic nuclear matter / Ş. Mişicu and S. Bastrukov. From N = Z to neutron rich: magnetic moments of Cu isotopes at and above the [symbol]Ni and [symbol]Ni double shell closures - what next? / N. J. Stone, J. R. Stone and U. Köster -- Nuclear structure VI. Decay studies of nuclei near [symbol]Ni / R. Grzywacz. Weakening of the [symbol]Ni core for Z > 28, N > 50? / J. A. Winger ... [et al.]. Coulomb excitation of the odd-A [symbol]Cu isotopes with MINIBALL and REX-ISOLDE / I. Stefanescu ... [et al.]. Neutron single particle states and isomers in odd mass nickel isotopes near [symbol]Ni / M. M. Raiabali ... [et al.]. [symbol] and [symbol]-delayed neutron decay studies of [symbol]Ch at the HRIBF / S. V. Ilvushkin ... [et al.] -- Posters. Properties of Fe, Ni and Zn isotope chains near the drip-line / V. N. Tarasov ... [et al.]. Probing nuclear structure of [symbol]Xe / J. B. Gupta. Shape coexistence in [symbol]Zr and large deformation in [symbol]Zr / J. K. Hwang ... [et al.]. Digital electronics and their application to beta decay spectroscopy / S. N. Liddick, S. Padgett and R. Grzywacz. Nuclear shape and structure in neutron-rich [symbol]Tc / Y. X. Luo ... [et al.]. Speeding up the r-process. Investigation of first forbidden [symbol] decays in N > 50 isotopes near [symbol]Ni / S. Padgett ... [et al.]. Yields of fission products from various actinide targets / E. H. Sveiewski ... [et al.].

Studies of neutron-rich nuclei far from stability at TRISTAN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gill, R.L.

The ISOL facility, TRISTAN, is a user facility located at Brookhaven National Laboratory's High Flux Beam Reactor. Short-lived, neutron-rich nuclei, far from stability, are produced by thermal neutron fission of /sup 235/U. An extensive array of experimental end stations are available for nuclear structure studies. These studies are augmented by a variety of long-lived ion sources suitable for use at a reactor facility. Some recent results at TRISTAN are presented as examples of using an ISOL facility to study series of nuclei, whereby an effective means of conducting nuclear structure investigations is available.

The Thermal Neutron Beam Option for NECTAR at MLZ

NASA Astrophysics Data System (ADS)

Mühlbauer, M. J.; Bücherl, T.; Genreith, C.; Knapp, M.; Schulz, M.; Söllradl, S.; Wagner, F. M.; Ehrenberg, H.

The beam port SR10 at the neutron source FRM II of Heinz Maier-Leibnitz Zentrum (MLZ) is equipped with a moveable assembly of two uranium plates, which can be placed in front of the entrance window of the beam tube via remote control. With these plates placed in their operating position the thermal neutron spectrum produced by the neutron source FRM II is converted to fission neutrons with 1.9 MeV of mean energy. This fission neutron spectrum is routinely used for medical applications at the irradiation facility MEDAPP, for neutron radiography and tomography experiments at the facility NECTAR and for materials testing. If, however, the uranium plates are in their stand-by position far off the tip of the beam tube and the so-called permanent filter for thermal neutrons is removed, thermal neutrons originating from the moderator tank enter the beam tube and a thermal spectrum becomes available for irradiation or activation of samples. By installing a temporary flight tube the beam may be used for thermal neutron radiography and tomography experiments at NECTAR. The thermal neutron beam option not only adds a pure thermal neutron spectrum to the energy ranges available for neutron imaging at MLZ instruments but it also is an unique possibility to combine two quite different neutron energy ranges at a single instrument including their respective advantages. The thermal neutron beam option for NECTAR is funded by BMBF in frame of research project 05K16VK3.

Precise ruthenium fission product isotopic analysis using dynamic reaction cell inductively coupled plasma mass spectrometry (DRC-ICP-MS)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Christopher F.; Dresel, P. Evan; Geiszler, Keith N.

2006-05-09

99Tc is a subsurface contaminant of interest at numerous federal, industrial, and international facilities. However, as a mono-isotopic fission product, 99Tc lacks the ability to be used as a signature to differentiate between the different waste disposal pathways that could have contributed to subsurface contamination at these facilities. Ruthenium fission-product isotopes are attractive analogues for the characterization of 99Tc sources because of their direct similarity to technetium with regard to subsurface mobility, and their large fission yields and low natural background concentrations. We developed an inductively coupled plasma mass spectrometry (ICP-MS) method capable of measuring ruthenium isotopes in groundwater samplesmore » and extracts of vadose zone sediments. Samples were analyzed directly on a Perkin Elmer ELAN DRC II ICP-MS after a single pass through a 1-ml bed volume of Dowex AG 50W-X8 100-200 mesh cation exchange resin. Precise ruthenium isotopic ratio measurements were achieved using a low-flow Meinhard-type nebulizer and long sample acquisition times (150,000 ms). Relative standard deviations of triplicate replicates were maintained at less than 0.5% when the total ruthenium solution concentration was 0.1 ng/ml or higher. Further work was performed to minimize the impact caused by mass interferences using the dynamic reaction cell (DRC) with O2 as the reaction gas. The aqueous concentrations of 96Mo and 96Zr were reduced by more than 99.7% in the reaction cell prior to injection of the sample into the mass analyzer quadrupole. The DRC was used in combination with stable-mass correction to quantitatively analyze samples containing up to 2-orders of magnitude more zirconium and molybdenum than ruthenium. The analytical approach documented herein provides an efficient and cost-effective way to precisely measure ruthenium isotopes and quantitate total ruthenium (natural vs. fission-product) in aqueous matrixes.« less

Burning plasma regime for Fussion-Fission Research Facility

NASA Astrophysics Data System (ADS)

Zakharov, Leonid E.

2010-11-01

The basic aspects of burning plasma regimes of Fusion-Fission Research Facility (FFRF, R/a=4/1 m/m, Ipl=5 MA, Btor=4-6 T, P^DT=50-100 MW, P^fission=80-4000 MW, 1 m thick blanket), which is suggested as the next step device for Chinese fusion program, are presented. The mission of FFRF is to advance magnetic fusion to the level of a stationary neutron source and to create a technical, scientific, and technology basis for the utilization of high-energy fusion neutrons for the needs of nuclear energy and technology. FFRF will rely as much as possible on ITER design. Thus, the magnetic system, especially TFC, will take advantage of ITER experience. TFC will use the same superconductor as ITER. The plasma regimes will represent an extension of the stationary plasma regimes on HT-7 and EAST tokamaks at ASIPP. Both inductive discharges and stationary non-inductive Lower Hybrid Current Drive (LHCD) will be possible. FFRF strongly relies on new, Lithium Wall Fusion (LiWF) plasma regimes, the development of which will be done on NSTX, HT-7, EAST in parallel with the design work. This regime will eliminate a number of uncertainties, still remaining unresolved in the ITER project. Well controlled, hours long inductive current drive operation at P^DT=50-100 MW is predicted.

METHOD OF TESTING THERMAL NEUTRON FISSIONABLE MATERIAL FOR PURITY

DOEpatents

Fermi, E.; Anderson, H.L.

1961-01-24

A process is given for determining the neutronic purity of fissionable material by the so-called shotgun test. The effect of a standard neutron absorber of known characteristics and amounts on a neutronic field also of known characteristics is measured and compared with the effect which the impurities derived from a known quantity of fissionable material has on the same neutronic field. The two readings are then made the basis of calculation from which the amount of impurities can be computed.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meierbachtol, K.; Tovesson, F.; Shields, D.

We developed the SPectrometer for Ion DEtermination in fission Research (SPIDER) for measuring mass yield distributions of fission products from spontaneous and neutron-induced fission. The 2E–2v method of measuring the kinetic energy (E) and velocity (v) of both outgoing fission products has been utilized, with the goal of measuring the mass of the fission products with an average resolution of 1 atomic mass unit (amu). Moreover, the SPIDER instrument, consisting of detector components for time-of-flight, trajectory, and energy measurements, has been assembled and tested using 229Th and 252Cf radioactive decay sources. For commissioning, the fully assembled system measured fission productsmore » from spontaneous fission of 252Cf. Individual measurement resolutions were met for time-of-flight (250 ps FWHM), spacial resolution (2 mm FHWM), and energy (92 keV FWHM for 8.376 MeV). Finally, these mass yield results measured from 252Cf spontaneous fission products are reported from an E–v measurement.« less

Utilization of the Differential Die-Away Self-Interrogation Technique for Characterization and Verification of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Trahan, Alexis Chanel

New nondestructive assay techniques are sought to better characterize spent nuclear fuel. One of the NDA instruments selected for possible deployment is differential die-away self-interrogation (DDSI). The proposed DDSI approach for spent fuel assembly assay utilizes primarily the spontaneous fission and (alpha, n) neutrons in the assemblies as an internal interrogating radiation source. The neutrons released in spontaneous fission or (alpha,n) reactions are thermalized in the surrounding water and induce fission in fissile isotopes, thereby creating a measurable signal from isotopes of interest that would be otherwise difficult to measure. The DDSI instrument employs neutron coincidence counting with 3He tubes and list-mode-based data acquisition to allow for production of Rossi-alpha distributions (RADs) in post-processing. The list-mode approach to data collection and subsequent construction of RADs has expanded the analytical possibilities, as will be demonstrated throughout this thesis. One of the primary advantages is that the measured signal in the form of a RAD can be analyzed in its entirety including determination of die-away times in different time domains. This capability led to the development of the early die-away method, a novel leakage multiplication determination method which is tested throughout the thesis on different sources in simulation space and fresh fuel experiments. The early die-away method is a robust, accurate, improved method of determining multiplication without the need for knowledge of the (alpha,n) source term. The DDSI technique and instrument are presented along with the many novel capabilities enabled by and discovered through RAD analysis. Among the new capabilities presented are the early die-away method, total plutonium content determination, and highly sensitive missing pin detection. Simulation of hundreds of different spent and fresh fuel assemblies were used to develop the analysis algorithms and the techniques were tested on a variety of spontaneous fission-driven fresh fuel assemblies at Los Alamos National Laboratory and the BeRP ball at the Nevada National Security Site. The development of the new, improved analysis and characterization methods with the DDSI instrument makes it a viable technique for implementation in a facility to meet material control and safeguards needs.

Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K.; Ross Finlay, M.

2012-06-01

The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

New fission-fragment detector for experiments at DANCE

NASA Astrophysics Data System (ADS)

Rusev, G.; Roman, A. R.; Daum, J. K.; Springs, R. K.; Bond, E. M.; Jandel, M.; Baramsai, B.; Bredeweg, T. A.; Couture, A.; Favalli, A.; Ianakiev, K. D.; Iliev, M. L.; Mosby, S.; Ullmann, J. L.; Walker, C. L.

2015-10-01

A fission-fragment detector based on thin scintillating films has been built to serve as a veto/trigger detector in neutron-induced fission measurements at DANCE. The fissile material is surrounded by scintillating films providing a 4 π detection of the fission fragments. The scintillation events caused by the fission fragment interactions in the films are registered with silicon photomultipliers. Design of the detector and test measurements are described. Work supported by the U.S. Department of Energy through the LANL/LDRD Program and the U.S. Department of Energy, Office of Science, Nuclear Physics under the Early Career Award No. LANL20135009.

DOE Office of Scientific and Technical Information (OSTI.GOV)

None, None

This report is concerned with the nature and scope of the technical services to be rendered and the general plan proposed for operation of Building 3525, High Radiation Level Examination Laboratory (HRLEL). The role of postirradiation examination in implementing the over- all task of irradiation testing for various programs under way at the Oak Ridge National Laboratory (ORNL) and the importance of this effort to the United Stat es reactor development program are stressed . The shielded-cell complex with provisions for remote decontamination, hot-equipment storage, and maintenance is described, as well as other supporting activities which are incorporated into themore » facility. The proposed technical functions include general observation, mensuration, nondestructive testing, burnup and induced-activity measurements, fission-gas sampling and analysis, corrosion evaluation and related measurements, disassembly and cutup, metallographic examination, mechanical-property determinations , and x -ray diffraction analyses. Equipment design and operational features to improve detection and measurement of selected properties in radioactive material s are described, also. The current status on design, procurement, construction, and preoperational testing of in- cell equipment in the mockup is presented along with a forecast of future needs. The mode of operation, manpower requirements, and management of the facility are discussed.« less

Effect of reactor radiation on the thermal conductivity of TREAT fuel

NASA Astrophysics Data System (ADS)

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Wright, Arthur E.; Yacout, Abdellatif M.

2017-04-01

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO2 particle size on fission-fragment damage. The proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

DHS Summary Report -- Robert Weldon

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weldon, Robert A.

This summer I worked on benchmarking the Lawrence Livermore National Laboratory fission multiplicity capability used in the Monte Carlo particle transport code MCNPX. This work involved running simulations and then comparing the simulation results with experimental experiments. Outlined in this paper is a brief description of the work completed this summer, skills and knowledge gained, and how the internship has impacted my planning for the future. Neutron multiplicity counting is a neutron detection technique that leverages the multiplicity emissions of neutrons from fission to identify various actinides in a lump of material. The identification of individual actinides in lumps ofmore » material crossing our boarders, especially U-235 and Pu-239, is a key component for maintaining the safety of the country from nuclear threats. Several multiplicity emission options from spontaneous and induced fission already existed in MCNPX 2.4.0. These options can be accessed through use of the 6th entry on the PHYS:N card. Lawrence Livermore National Laboratory (LLNL) developed a physics model for the simulation of neutron and gamma ray emission from fission and photofission that was included in MCNPX 2.7.B as an undocumented feature and then was documented in MCNPX 2.7.C. The LLNL multiplicity capability provided a different means for MCNPX to simulate neutron and gamma-ray distributions for neutron induced, spontaneous and photonuclear fission reactions. The original testing on the model for implementation into MCNPX was conducted by Gregg McKinney and John Hendricks. The model is an encapsulation of measured data of neutron multiplicity distributions from Gwin, Spencer, and Ingle, along with the data from Zucker and Holden. One of the founding principles of MCNPX was that it would have several redundant capabilities, providing the means of testing and including various physics packages. Though several multiplicity sampling methodologies already existed within MCNPX, the LLNL fission multiplicity was included to provide a separate capability for computing multiplicity as well as including several new features not already included in MCNPX. These new features include: (1) prompt gamma emission/multiplicity from neutron-induced fission; (2) neutron multiplicity and gamma emission/multiplicity from photofission; and (3) an option to enforce energy correlation for gamma neutron multiplicity emission. These new capabilities allow correlated signal detection for identifying presence of special nuclear material (SNM). Therefore, these new capabilities help meet the missions of the Domestic Nuclear Detection Office (DNDO), which is tasked with developing nuclear detection strategies for identifying potential radiological and nuclear threats, by providing new simulation capability for detection strategies that leverage the new available physics in the LLNL multiplicity capability. Two types of tests were accomplished this summer to test the default LLNL neutron multiplicity capability: neutron-induced fission tests and spontaneous fission tests. Both cases set the 6th entry on the PHYS:N card to 5 (i.e. use LLNL multiplicity). The neutron-induced fission tests utilized a simple 0.001 cm radius sphere where 0.0253 eV neutrons were released at the sphere center. Neutrons were forced to immediately collide in the sphere and release all progeny from the sphere, without further collision, using the LCA card, LCA 7j -2 (therefore density and size of the sphere were irrelevant). Enough particles were run to ensure that the average error of any specific multiplicity did not exceed 0.36%. Neutron-induced fission multiplicities were computed for U-233, U-235, Pu-239, and Pu-241. The spontaneous fission tests also used the same spherical geometry, except: (1) the LCA card was removed; (2) the density of the sphere was set to 0.001 g/cm3; and (3) instead of emitting a thermal neutron, the PAR keyword was set to PAR=SF. The purpose of the small density was to ensure that the spontaneous fission neutrons would not further interact and induce fissions (i.e. the mean free path greatly exceeded the size of the sphere). Enough particles were run to ensure that the average error of any specific spontaneous multiplicity did not exceed 0.23%. Spontaneous fission multiplicities were computed for U-238, Pu-238, Pu-240, Pu-242, Cm-242, and Cm-244. All of the computed results were compared against experimental results compiled by Holden at Brookhaven National Laboratory.« less

Measurements of extinct fission products in nuclear bomb debris: Determination of the yield of the Trinity nuclear test 70 y later

DOE PAGES

Hanson, Susan Kloek; Pollington, Anthony Douglas; Waidmann, Christopher Russell; ...

2016-07-05

This study describes an approach to measuring extinct fission products that would allow for the characterization of a nuclear test at any time. The isotopic composition of molybdenum in five samples of glassy debris from the 1945 Trinity nuclear test has been measured. Nonnatural molybdenum isotopic compositions were observed, reflecting an input from the decay of the short-lived fission products 95Zr and 97Zr. By measuring both the perturbation of the 95Mo/ 96Mo and 97Mo/ 96Mo isotopic ratios and the total amount of molybdenum in the Trinity nuclear debris samples, it is possible to calculate the original concentrations of the 95Zrmore » and 97Zr isotopes formed in the nuclear detonation. Together with a determination of the amount of plutonium in the debris, these measurements of extinct fission products allow for new estimates of the efficiency and yield of the historic Trinity test.« less

Measurements of extinct fission products in nuclear bomb debris: Determination of the yield of the Trinity nuclear test 70 y later

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hanson, Susan Kloek; Pollington, Anthony Douglas; Waidmann, Christopher Russell

This study describes an approach to measuring extinct fission products that would allow for the characterization of a nuclear test at any time. The isotopic composition of molybdenum in five samples of glassy debris from the 1945 Trinity nuclear test has been measured. Nonnatural molybdenum isotopic compositions were observed, reflecting an input from the decay of the short-lived fission products 95Zr and 97Zr. By measuring both the perturbation of the 95Mo/ 96Mo and 97Mo/ 96Mo isotopic ratios and the total amount of molybdenum in the Trinity nuclear debris samples, it is possible to calculate the original concentrations of the 95Zrmore » and 97Zr isotopes formed in the nuclear detonation. Together with a determination of the amount of plutonium in the debris, these measurements of extinct fission products allow for new estimates of the efficiency and yield of the historic Trinity test.« less

Measurements of extinct fission products in nuclear bomb debris: Determination of the yield of the Trinity nuclear test 70 y later

PubMed Central

Hanson, Susan K.; Pollington, Anthony D.; Waidmann, Christopher R.; Kinman, William S.; Wende, Allison M.; Miller, Jeffrey L.; Berger, Jennifer A.; Oldham, Warren J.; Selby, Hugh D.

2016-01-01

This paper describes an approach to measuring extinct fission products that would allow for the characterization of a nuclear test at any time. The isotopic composition of molybdenum in five samples of glassy debris from the 1945 Trinity nuclear test has been measured. Nonnatural molybdenum isotopic compositions were observed, reflecting an input from the decay of the short-lived fission products 95Zr and 97Zr. By measuring both the perturbation of the 95Mo/96Mo and 97Mo/96Mo isotopic ratios and the total amount of molybdenum in the Trinity nuclear debris samples, it is possible to calculate the original concentrations of the 95Zr and 97Zr isotopes formed in the nuclear detonation. Together with a determination of the amount of plutonium in the debris, these measurements of extinct fission products allow for new estimates of the efficiency and yield of the historic Trinity test. PMID:27382169

Implementing and testing theoretical fission fragment yields in a Hauser-Feshbach statistical decay framework

NASA Astrophysics Data System (ADS)

Jaffke, Patrick; Möller, Peter; Stetcu, Ionel; Talou, Patrick; Schmitt, Christelle

2018-03-01

We implement fission fragment yields, calculated using Brownian shape-motion on a macroscopic-microscopic potential energy surface in six dimensions, into the Hauser-Feshbach statistical decay code CGMF. This combination allows us to test the impact of utilizing theoretically-calculated fission fragment yields on the subsequent prompt neutron and γ-ray emission. We draw connections between the fragment yields and the total kinetic energy TKE of the fission fragments and demonstrate that the use of calculated yields can introduce a difference in the 〈TKE〉 and, thus, the prompt neutron multiplicity v, as compared with experimental fragment yields. We deduce the uncertainty on the 〈TKE〉 and v from this procedure and identify possible applications.

Laboratory-scale uranium RF plasma confinement experiments

NASA Technical Reports Server (NTRS)

Roman, W. C.

1976-01-01

An experimental investigation was conducted using 80 kW and 1.2 MW RF induction heater facilities to aid in developing the technology necessary for designing a self-critical fissioning uranium plasma core reactor. Pure uranium hexafluoride (UF6) was injected into argon-confined, steady-state, RF-heated plasmas in different uranium plasma confinement tests to investigate the characteristics of plamas core nuclear reactors. The objectives were: (1) to confine as high a density of uranium vapor as possible within the plasma while simultaneously minimizing the uranium compound wall deposition; (2) to develop and test materials and handling techniques suitable for use with high-temperature, high-pressure gaseous UF6; and (3) to develop complementary diagnostic instrumentation and measurement techniques to characterize the uranium plasma and residue deposited on the test chamber components. In all tests, the plasma was a fluid-mechanically-confined vortex-type contained within a fused-silica cylindrical test chamber. The test chamber peripheral wall was 5.7 cm ID by 10 cm long.

Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

2016-06-01

Low-enriched (U-235 <20 pct) U-Mo dispersion fuel is being developed for use in research and test reactors. In most cases, fuel plates with Al or Al-Si alloy matrices have been tested in the Advanced Test Reactor to support this development. In addition, fuel plates with Mg as the matrix have also been tested. The benefit of using Mg as the matrix is that it potentially will not chemically interact with the U-Mo fuel particles during fabrication or irradiation, whereas with Al and Al-Si alloys such interactions will occur. Fuel plate R9R010 is a Mg matrix fuel plate that was aggressively irradiated in ATR. This fuel plate was irradiated as part of the RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

NASA Astrophysics Data System (ADS)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Roberts, Jeremy A.; Unruh, Troy C.; McGregor, Douglas S.

2018-01-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of neutron flux in the reactor core.

Development of a thin scintillation films fission-fragment detector and a novel neutron source

NASA Astrophysics Data System (ADS)

Rusev, G.; Jandel, M.; Baramsai, B.; Bond, E. M.; Bredeweg, T. A.; Couture, A.; Daum, J. K.; Favalli, A.; Ianakiev, K. D.; Iliev, M. L.; Mosby, S.; Roman, A. R.; Springs, R. K.; Ullmann, J. L.; Walker, C. L.

2015-08-01

Investigation of prompt fission and neutron-capture Υ rays from fissile actinide samples at the Detector for Advanced Neutron Capture Experiments (DANCE) requires use of a fission-fragment detector to provide a trigger or a veto signal. A fission-fragment detector based on thin scintillating films and silicon photomultipliers has been built to serve as a trigger/veto detector in neutron-induced fission measurements at DANCE. The fissile material is surrounded by scintillating films providing a 4π detection of the fission fragments. The scintillations were registered with silicon photomultipliers. A measurement of the 235U(n,f) reaction with this detector at DANCE revealed a correct time-of-flight spectrum and provided an estimate for the efficiency of the prototype detector of 11.6(7)%. Design and test measurements with the detector are described. A neutron source with fast timing has been built to help with detector-response measurements. The source is based on the neutron emission from the spontaneous fission of 252Cf and the same type of scintillating films and silicon photomultipliers. Overall time resolution of the source is 0.3 ns. Design of the source and test measurements with it are described. An example application of the source for determining the neutron/gamma pulse-shape discrimination by a stilbene crystal is given.

Simulating the venting of radioactivity from a soviet nuclear test

NASA Astrophysics Data System (ADS)

Rodriguez, Daniel J.; Peterson, Kendall R.

Fresh fission products were found in several routine air samples in Europe during the second and third weeks of March 1987. Initially, it was suspected that the radionuclides, principally 133Xe and 131I, had been accidentally released from a European facility handling nuclear materials. However, the announcement of an underground nuclear test at Semipalatinsk, U.S.S.R. on 26 February 1987 suggested that the elevated amounts of radioactivity may, instead, have been caused by a venting episode. Upon learning of these events, we simulated the transport and diffusion of 133Xe with our Hemispheric MEDIC and ADPIC models, assuming Semipalatinsk to be the source of the radioactive emissions. The correspondence between the calculated concentrations and the daily average 133Xe measurements made by the Federal Office for Civil Protection in F.R.G. was excellent. While this agreement does not, in itself, prove that an atmospheric venting of radioactive material occurred at Semipalatinsk, a body of circumstantial evidence exists which, when added together, strongly supports this conclusion. Our calculations suggested a total fission yield of about 40 kt, which is within the 20-150 kt range of tests acknowledged by the U.S.S.R. Finally, dose calculations indicated that no health or environmental impact occurred outside of the U.S.S.R. due to the suspected venting of 133Xe. However, the inhalation dose resulting from 133I, an unmodeled component of the radioactive cloud, represented a greater potential risk to public health.

Real-time radiography at the NECTAR facility

NASA Astrophysics Data System (ADS)

Bücherl, T.; Lierse von Gostomski, Ch.

2011-09-01

A feasibility study has shown that real-time radiography using fission neutrons is possible at the NECTAR facility, when using an improved detection system for fast variations (Bücherl et al., 2009 [1]). Continuing this study, real-time measurements of slowly varying processes like the water uptake in medium sized trunks (diameter about 12 cm) and of slow periodic processes (e.g. a slowly rotating iron disk) are investigated successfully using the existing detection system.

Fissile interrogation using gamma rays from oxygen

DOEpatents

Smith, Donald; Micklich, Bradley J.; Fessler, Andreas

2004-04-20

The subject apparatus provides a means to identify the presence of fissionable material or other nuclear material contained within an item to be tested. The system employs a portable accelerator to accelerate and direct protons to a fluorine-compound target. The interaction of the protons with the fluorine-compound target produces gamma rays which are directed at the item to be tested. If the item to be tested contains either a fissionable material or other nuclear material the interaction of the gamma rays with the material contained within the test item with result in the production of neutrons. A system of neutron detectors is positioned to intercept any neutrons generated by the test item. The results from the neutron detectors are analyzed to determine the presence of a fissionable material or other nuclear material.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marques, J.G.; Ramos, A.R.; Fernandes, A.C.

The behavior of electronic components and circuits under radiation is a concern shared by the nuclear industry, the space community and the high-energy physics community. Standard commercial components are used as much as possible instead of radiation hard components, since they are easier to obtain and allow a significant reduction of costs. However, these standard components need to be tested in order to determine their radiation tolerance. The Portuguese Research Reactor (RPI) is a 1 MW pool-type reactor, operating since 1961. The irradiation of electronic components and circuits is one area where a 1 MW reactor can be competitive, sincemore » the fast neutron fluences required for testing are in most cases well below 10{sup 16} n/cm{sup 2}. A program was started in 1999 to test electronics components and circuits for the LHC facility at CERN, initially using a dedicated in-pool irradiation device and later a beam line with tailored neutron and gamma filters. Neutron filters are essential to reduce the intensity of the thermal neutron flux, which does not produce significant defects in electronic components but produces unwanted radiation from activation of contacts and packages of integrated circuits and also of the printed circuit boards. In irradiations performed within the line-of-sight of the core of a fission reactor there is simultaneous gamma radiation which complicates testing in some cases. Filters can be used to reduce its importance and separate testing with a pure gamma radiation source can contribute to clarify some irradiation results. Practice has shown the need to introduce several improvements to the procedures and facilities over the years. We will review improvements done in the following areas: - Optimization of neutron and gamma filters; - Dosimetry procedures in mixed neutron / gamma fields; - Determination of hardness parameter and 1 MeV-equivalent neutron fluence; - Temperature measurement and control during irradiation; - Follow-up of reactor power operational fluctuations; - Study of gamma radiation effects only. The fission neutron spectrum can be limitative for some of the tests, as most neutrons are in the 1-2 MeV energy range. Significant progress has been made lately in compact neutron generators using D-D and D-T fusion reactions, achieving higher neutron fluxes and longer lifetime than previously available. The advantages of using compact neutron generators for testing of electronic components and circuits will be also discussed. (authors)« less

Validation of Cross Sections with Criticality Experiment and Reaction Rates: the Neptunium Case

NASA Astrophysics Data System (ADS)

Leong, L. S.; Tassan-Got, L.; Audouin, L.; Berthier, B.; Le Naour, C.; Stéphan, C.; Paradela, C.; Tarrío, D.; Duran, I.

2014-04-01

The 237Np neutron-induced fission cross section has been recently measured in a large energy range (from eV to GeV) at the n_TOF facility at CERN. When compared to previous measurements the n_TOF fission cross section appears to be higher by 5-7% beyond the fission threshold. To check the relevance of the n_TOF data, we considered a criticality experiment performed at Los Alamos with a 6 kg sphere of 237Np, surrounded by uranium highly enriched in 235U so as to approach criticality with fast neutrons. The multiplication factor keff of the calculation is in better agreement with the experiment when we replace the ENDF/B-VII.0 evaluation of the 237Np fission cross section by the n_TOF data. We also explored the hypothesis of deficiencies of the inelastic cross section in 235U which has been invoked by some authors to explain the deviation of 750 pcm. The large modification needed to reduce the deviation seems to be incompatible with existing inelastic cross section measurements. Also we show that the νbar of 237Np can hardly be incriminated because of the high accuracy of the existing data. Fission rate ratios or averaged fission cross sections measured in several fast neutron fields seem to give contradictory results on the validation of the 237Np cross section but at least one of the benchmark experiments, where the active deposits have been well calibrated for the number of atoms, favors the n_TOF data set. These outcomes support the hypothesis of a higher fission cross section of 237Np.

Criticality experiments and benchmarks for cross section evaluation: the neptunium case

NASA Astrophysics Data System (ADS)

Leong, L. S.; Tassan-Got, L.; Audouin, L.; Paradela, C.; Wilson, J. N.; Tarrio, D.; Berthier, B.; Duran, I.; Le Naour, C.; Stéphan, C.

2013-03-01

The 237Np neutron-induced fission cross section has been recently measured in a large energy range (from eV to GeV) at the n_TOF facility at CERN. When compared to previous measurement the n_TOF fission cross section appears to be higher by 5-7% beyond the fission threshold. To check the relevance of n_TOF data, we apply a criticality experiment performed at Los Alamos with a 6 kg sphere of 237Np, surrounded by enriched uranium 235U so as to approach criticality with fast neutrons. The multiplication factor ke f f of the calculation is in better agreement with the experiment (the deviation of 750 pcm is reduced to 250 pcm) when we replace the ENDF/B-VII.0 evaluation of the 237Np fission cross section by the n_TOF data. We also explore the hypothesis of deficiencies of the inelastic cross section in 235U which has been invoked by some authors to explain the deviation of 750 pcm. With compare to inelastic large distortion calculation, it is incompatible with existing measurements. Also we show that the v of 237Np can hardly be incriminated because of the high accuracy of the existing data. Fission rate ratios or averaged fission cross sections measured in several fast neutron fields seem to give contradictory results on the validation of the 237Np cross section but at least one of the benchmark experiments, where the active deposits have been well calibrated for the number of atoms, favors the n_TOF data set. These outcomes support the hypothesis of a higher fission cross section of 237Np.

Iodine-129 measurements in soil samples from Dolon village near the Semipalatinsk nuclear test site.

PubMed

Endo, Satoru; Tomita, Junpei; Tanaka, Kenichi; Yamamoto, Masayoshi; Fukutani, Satoshi; Imanaka, Tetsuji; Sakaguchi, Aya; Amano, Hikaru; Kawamura, Hidehisa; Kawamura, Hisao; Apsalikov, Kazbek N; Gusev, Boris I; Whitehead, Neil E; Shinkarev, Sergey; Hoshi, Masaharu

2008-07-01

Dolon village, located about 60 km from the border of the Semipalatinsk nuclear test site, is known to be heavily contaminated by the first USSR atomic bomb test in August 1949. Soil samples around Dolon were taken in October 2005 in an attempt to evaluate internal thyroid dose arising from incorporation of radioiodine isotopes (mainly (131)I). Iodine-129 in soil was measured by using the technique of accelerator mass spectrometry. The (129)I/(127)I atom ratios measured were in the range from 3.3 x 10(-9) to 3.3 x 10(-7). These values were within the range of the current background level ( approximately 10(-9) to 10(-7)) in the environment, including contributions from the global fallout of atmospheric nuclear tests and local fallout of nuclear facilities. The (129)I atom accumulated level in soil ranged from 1.28 x 10(13) to 1.59 x 10(14) atoms m(-2), the average (8.0 x 10(13)) of which was higher than the background level of (2-5) x 10(13). From the relationship between (129)I and( 137)Cs (corrected for background and decay from 1949 to 2005) accumulated levels, the background level of (129)I and the (129)I/(137)Cs ratio around Dolon were estimated to be (6.4 +/- 0.4) x 10(13) atoms m(-2) and 0.25 +/- 0.16, respectively. This (129)I/(137)Cs ratio is almost similar to the fission yield ratio for (239)Pu fast fission (0.24).

Maximal design basis accident of fusion neutron source DEMO-TIN

NASA Astrophysics Data System (ADS)

Kolbasov, B. N.

2015-12-01

When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission-fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or the first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.

Radionuclide Basics: Iodine

MedlinePlus

... the release of a relatively large amount of energy. Fissioning that occurs without any outside cause is called "spontaneous fission." reaction from either nuclear weapons testing or nuclear power plants . Some forms ...

Isolation and Purification of the Xenon Fraction of 252Cf Spontaneous Fission Products for the Production of Radio Xenon Calibration Standards

DOE Office of Scientific and Technical Information (OSTI.GOV)

McGrath, Christopher A.

2015-04-01

The presence of radioactive xenon isotopes indicates that fission events have occurred, and is used to help enforce the Comprehensive Test Ban Treaty. Idaho National Laboratory (INL) produces 135Xe, 133mXe, 133Xe, and 131mXe standards used for the calibration and testing of collection equipment and analytical techniques used to monitor radio xenon emissions. At INL, xenon is produced and collected as one of several spontaneous fission products from a 252Cf source. Further chromatographic purification of the fission gases ensures the separations of the xenon fraction for selective collection. An explanation of the fission gas collection, separation and purification is presented. Additionally,more » the range of 135Xe to 133Xe ratio that can be isolated is explained. This is an operational update on the work introduced previously, now that it is in operation and has been recharged with a second 252Cf source.« less

HIGH-TEMPERATURE SAFETY TESTING OF IRRADIATED AGR-1 TRISO FUEL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stempien, John D.; Demkowicz, Paul A.; Reber, Edward L.

High-Temperature Safety Testing of Irradiated AGR-1 TRISO Fuel John D. Stempien, Paul A. Demkowicz, Edward L. Reber, and Cad L. Christensen Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415, USA Corresponding Author: john.stempien@inl.gov, +1-208-526-8410 Two new safety tests of irradiated tristructural isotropic (TRISO) coated particle fuel have been completed in the Fuel Accident Condition Simulator (FACS) furnace at the Idaho National Laboratory (INL). In the first test, three fuel compacts from the first Advanced Gas Reactor irradiation experiment (AGR-1) were simultaneously heated in the FACS furnace. Prior to safety testing, each compact was irradiated in the Advanced Testmore » Reactor to a burnup of approximately 15 % fissions per initial metal atom (FIMA), a fast fluence of 3×1025 n/m2 (E > 0.18 MeV), and a time-average volume-average (TAVA) irradiation temperature of about 1020 °C. In order to simulate a core-conduction cool-down event, a temperature-versus-time profile having a peak temperature of 1700 °C was programmed into the FACS furnace controllers. Gaseous fission products (i.e., Kr-85) were carried to the Fission Gas Monitoring System (FGMS) by a helium sweep gas and captured in cold traps featuring online gamma counting. By the end of the test, a total of 3.9% of an average particle’s inventory of Kr-85 was detected in the FGMS traps. Such a low Kr-85 activity indicates that no TRISO failures (failure of all three TRISO layers) occurred during the test. If released from the compacts, condensable fission products (e.g., Ag-110m, Cs-134, Cs-137, Eu-154, Eu-155, and Sr-90) were collected on condensation plates fitted to the end of the cold finger in the FACS furnace. These condensation plates were then analyzed for fission products. In the second test, five loose UCO fuel kernels, obtained from deconsolidated particles from an irradiated AGR-1 compact, were heated in the FACS furnace to a peak temperature of 1600 °C. This test had two primary goals. First, the test was intended to assess the retention of fission products in loose kernels without the effects of the other TRISO layers (buffer, IPyC, SiC, and OPyC) or the graphitic matrix material comprising the compact. Second, this test served as an evaluation of the FACS fission product condensation plate collection efficiency.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

L. Zakharov, J. Li and Y. Wu

The project of ASIPP (with PPPL participation), called FFRF, (R/a=4/1 m/m, Ipl=5 MA, Btor=4-6 T, PDT=50-100 MW, Pfission=80-4000 MW, 1 m thick blanket) is outlined. FFRF stands for the Fusion-Fission Research Facility with a unique fusion mission and a pioneering mission of merging fusion and fission for accumulation of design, experimental, and operational data for future hybrid applications. The design of FFRF will use as much as possible the EAST and ITER design experience. On the other hand, FFRF strongly relies on new, Lithium Wall Fusion plasma regimes, the development of which has already started in the US and China.

Dose profile modeling of Idaho National Laboratory's active neutron interrogation laboratory.

PubMed

Chichester, D L; Seabury, E H; Zabriskie, J M; Wharton, J; Caffrey, A J

2009-06-01

A new laboratory has been commissioned at Idaho National Laboratory for performing active neutron interrogation research and development. The facility is designed to provide radiation shielding for deuterium-tritium (DT) fusion (14.1 MeV) neutron generators (2 x 10(8) n/s), deuterium-deuterium (DD) fusion (2.5 MeV) neutron generators (1 x 10(7) n/s), and (252)Cf spontaneous fission neutron sources (6.96 x 10(7) n/s, 30 microg). Shielding at the laboratory is comprised of modular concrete shield blocks 0.76 m thick with tongue-in-groove features to prevent radiation streaming, arranged into one small and one large test vault. The larger vault is designed to allow operation of the DT generator and has walls 3.8m tall, an entrance maze, and a fully integrated electrical interlock system; the smaller test vault is designed for (252)Cf and DD neutron sources and has walls 1.9 m tall and a simple entrance maze. Both analytical calculations and numerical simulations were used in the design process for the building to assess the performance of the shielding walls and to ensure external dose rates are within required facility limits. Dose rate contour plots have been generated for the facility to visualize the effectiveness of the shield walls and entrance mazes and to illustrate the spatial profile of the radiation dose field above the facility and the effects of skyshine around the vaults.

Experimental Measurements at the MASURCA Facility

NASA Astrophysics Data System (ADS)

Assal, W.; Bosq, J. C.; Mellier, F.

2012-12-01

Dedicated to the neutronics studies of fast and semi-fast reactor lattices, MASURCA (meaning “mock-up facility for fast breeder reactor studies at CADARACHE”) is an airflow cooled fast reactor operating at a maximum power of 5 kW playing an important role in the CEA research activities. At this facility, a lot of neutron integral experimental programs were undertaken. The purpose of this poster is to show a panorama of the facility from this experimental measurement point of view. A hint at the forthcoming refurbishment will be included. These programs include various experimental measurements (reactivity, distributions of fluxes, reaction rates), performed essentially with fission chambers, in accordance with different methods (noise methods, radial or axial traverses, rod drops) and involving several devices systems (monitors, fission chambers, amplifiers, power supplies, data acquisition systems ...). For this purpose are implemented electronics modules to shape the signals sent from the detectors in various mode (fluctuation, pulse, current). All the electric and electronic devices needed for these measurements and the relating wiring will be fully explained through comprehensive layouts. Data acquired during counting performed at the time of startup phase or rod drops are analyzed by the mean of a Neutronic Measurement Treatment (TMN in French) programmed on the basis of the MATLAB software. This toolbox gives the opportunity of data files management, reactivity valuation from neutronics measurements and transient or divergence simulation at zero power. Particular TMN using at MASURCA will be presented.

Monitoring the Gas Composition of the NIFFTE Time Projection Chamber

NASA Astrophysics Data System (ADS)

Towell, Travis; Travis Towell Collaboration

2017-09-01

The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) at Los Alamos National Laboratory(LANL) is using a Time Projection Chamber (TPC) to measure with high precision the cross section ratio of U238 to P239. When the neutron beam hits a target, it may emit fission fragments. As the fission fragments travels through the chamber, it ionizes the gas it passes through. Based on the time it takes for the ions to drift to the pad planes and the hit location of the ions, the path of fission fragments can be determined. Knowing the composition of the gas mixture is vital to accurately reconstruct the data. A Binary Gas Analyzer (BGA) is used to measure the gas composition. To confirm the accuracy of the BGA, varying amounts of nitrogen and carbon dioxide were flowed through a test gas system. Several tests were performed to validate that the BGA for our gas system is working properly. This poster will describe the test gas system setup, tests of the BGA, and elaborate on the main goals of the NIFFTE experiment.

Hardware Based Technology Assessment in Support of Near-Term Space Fission Missions

NASA Technical Reports Server (NTRS)

Houts, Mike; VanDyke, Melissa; Godfroy, Tom; Martin, James; BraggSitton, Shannon; Carter, Robert; Dickens, Ricky; Salvail, Pat; Williams, Eric; Harper, Roger

2003-01-01

Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and utilized. Successful utilization will most likely occur if frequent, significant hardware-based milestones can be achieved throughout the program. Achieving these milestones will depend on the capability to perform highly realistic non-nuclear testing of nuclear systems. This paper discusses ongoing and potential research that could help achieve these milestones.

Heat Pipe Powered Stirling Conversion for the Demonstration Using Flattop Fission (DUFF) Test

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Briggs, Maxwell H.; Sanzi, James L.; Brace, Michael H.

2013-01-01

Design concepts for small Fission Power Systems (FPS) have shown that heat pipe cooled reactors provide a passive, redundant, and lower mass option to transfer heat from the fuel to the power conversion system, as opposed to pumped loop designs typically associated with larger FPS. Although many systems have been conceptually designed and a few making it to electrically heated testing, none have been coupled to a real nuclear reactor. A demonstration test named DUFF Demonstration Using Flattop Fission, was planned by the Los Alamos National Lab (LANL) to use an existing criticality experiment named Flattop to provide the nuclear heat source. A team from the NASA Glenn Research Center designed, built, and tested a heat pipe and power conversion system to couple to Flattop with the end goal of making electrical power. This paper will focus on the design and testing performed in preparation for the DUFF test.

ENGINEERING TEST REACTOR

DOEpatents

De Boisblanc, D.R.; Thomas, M.E.; Jones, R.M.; Hanson, G.H.

1958-10-21

Heterogeneous reactors of the type which is both cooled and moderated by the same fluid, preferably water, and employs highly enriched fuel are reported. In this design, an inner pressure vessel is located within a main outer pressure vessel. The reactor core and its surrounding reflector are disposed in the inner pressure vessel which in turn is surrounded by a thermal shield, Coolant fluid enters the main pressure vessel, fiows downward into the inner vessel where it passes through the core containing tbe fissionable fuel assemblies and control rods, through the reflector, thence out through the bottom of the inner vessel and up past the thermal shield to the discharge port in the main vessel. The fuel assemblles are arranged in the core in the form of a cross having an opening extending therethrough to serve as a high fast flux test facility.

Materials challenges for nuclear systems

DOE PAGES

Allen, Todd; Busby, Jeremy; Meyer, Mitch; ...

2010-11-26

The safe and economical operation of any nuclear power system relies to a great extent, on the success of the fuel and the materials of construction. During the lifetime of a nuclear power system which currently can be as long as 60 years, the materials are subject to high temperature, a corrosive environment, and damage from high-energy particles released during fission. The fuel which provides the power for the reactor has a much shorter life but is subject to the same types of harsh environments. This article reviews the environments in which fuels and materials from current and proposed nuclearmore » systems operate and then describes how the creation of the Advanced Test Reactor National Scientific User Facility is allowing researchers from across the U.S. to test their ideas for improved fuels and materials.« less

Development of a thin scintillation films fission-fragment detector and a novel neutron source

DOE PAGES

Rusev, Gencho Yordanov; Jandel, Marian; Baramsai, Bayarbadrakh; ...

2015-08-26

Here, investigation of prompt fission and neutron-capture Υ rays from fissile actinide samples at the Detector for Advanced Neutron Capture Experiments (DANCE) requires use of a fission-fragment detector to provide a trigger or a veto signal. A fission-fragment detector based on thin scintillating films and silicon photomultipliers has been built to serve as a trigger/veto detector in neutron-induced fission measurements at DANCE. The fissile material is surrounded by scintillating films providing a 4π detection of the fission fragments. The scintillations were registered with silicon photomultipliers. A measurement of the 235U(n,f) reaction with this detector at DANCE revealed a correct time-of-flightmore » spectrum and provided an estimate for the efficiency of the prototype detector of 11.6(7)%. Design and test measurements with the detector are described. A neutron source with fast timing has been built to help with detector-response measurements. The source is based on the neutron emission from the spontaneous fission of 252Cf and the same type of scintillating films and silicon photomultipliers. Overall time resolution of the source is 0.3 ns. Design of the source and test measurements with it are described. An example application of the source for determining the neutron/gamma pulse-shape discrimination by a stilbene crystal is given.« less

Human Health and Ecological Risk Assessment for the Operation of the Explosives Waste Treatment Facility at Site 300 of the Lawrence Livermore National Laboratory Volume 1: Report of Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gallegos, G; Daniels, J; Wegrecki, A

2006-04-24

This document contains the human health and ecological risk assessment for the Resource Recovery and Conservation Act (RCRA) permit renewal for the Explosives Waste Treatment Facility (EWTF). Volume 1 is the text of the risk assessment, and Volume 2 (provided on a compact disc) is the supporting modeling data. The EWTF is operated by the Lawrence Livermore National Laboratory (LLNL) at Site 300, which is located in the foothills between the cities of Livermore and Tracy, approximately 17 miles east of Livermore and 8 miles southwest of Tracy. Figure 1 is a map of the San Francisco Bay Area, showingmore » the location of Site 300 and other points of reference. One of the principal activities of Site 300 is to test what are known as ''high explosives'' for nuclear weapons. These are the highly energetic materials that provide the force to drive fissionable material to criticality. LLNL scientists develop and test the explosives and the integrated non-nuclear components in support of the United States nuclear stockpile stewardship program as well as in support of conventional weapons and the aircraft, mining, oil exploration, and construction industries. Many Site 300 facilities are used in support of high explosives research. Some facilities are used in the chemical formulation of explosives; others are locations where explosive charges are mechanically pressed; others are locations where the materials are inspected radiographically for such defects as cracks and voids. Finally, some facilities are locations where the machined charges are assembled before they are sent to the on-site test firing facilities, and additional facilities are locations where materials are stored. Wastes generated from high-explosives research are treated by open burning (OB) and open detonation (OD). OB and OD treatments are necessary because they are the safest methods for treating explosives wastes generated at these facilities, and they eliminate the requirement for further handling and transportation that would be required if the wastes were treated off site.« less

Human Health and Ecological Risk Assessment for the Operation of the Explosives Waste Treatment Facility at Site 300 of the Lawrence Livermore National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gallegos, G; Daniels, J; Wegrecki, A

2007-10-01

This document contains the human health and ecological risk assessment for the Resource Recovery and Conservation Act (RCRA) permit renewal for the Explosives Waste Treatment Facility (EWTF). Volume 1 is the text of the risk assessment, and Volume 2 (provided on a compact disc) is the supporting modeling data. The EWTF is operated by the Lawrence Livermore National Laboratory (LLNL) at Site 300, which is located in the foothills between the cities of Livermore and Tracy, approximately 17 miles east of Livermore and 8 miles southwest of Tracy. Figure 1 is a map of the San Francisco Bay Area, showingmore » the location of Site 300 and other points of reference. One of the principal activities of Site 300 is to test what are known as 'high explosives' for nuclear weapons. These are the highly energetic materials that provide the force to drive fissionable material to criticality. LLNL scientists develop and test the explosives and the integrated non-nuclear components in support of the United States nuclear stockpile stewardship program as well as in support of conventional weapons and the aircraft, mining, oil exploration, and construction industries. Many Site 300 facilities are used in support of high explosives research. Some facilities are used in the chemical formulation of explosives; others are locations where explosive charges are mechanically pressed; others are locations where the materials are inspected radiographically for such defects as cracks and voids. Finally, some facilities are locations where the machined charges are assembled before they are sent to the onsite test firing facilities, and additional facilities are locations where materials are stored. Wastes generated from high-explosives research are treated by open burning (OB) and open detonation (OD). OB and OD treatments are necessary because they are the safest methods for treating explosives wastes generated at these facilities, and they eliminate the requirement for further handling and transportation that would be required if the wastes were treated off site.« less

Effect of reactor radiation on the thermal conductivity of TREAT fuel

DOE PAGES

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; ...

2017-02-04

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2more » particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« less

Effect of reactor radiation on the thermal conductivity of TREAT fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2more » particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« less

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W.; Ross Finlay, M.; Moore, Glenn; Medvedev, Pavel; Meyer, Mitch

2017-05-01

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

Design of a heatpipe-cooled Mars-surface fission reactor

NASA Astrophysics Data System (ADS)

Poston, David I.; Kapernick, Richard J.; Guffee, Ray M.; Reid, Robert S.; Lipinski, Ronald J.; Wright, Steven A.; Talandis, Regina A.

2002-01-01

The next generation of robotic missions to Mars will most likely require robust power sources in the range of 3 to 20 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The goal of this study is to design a compact, low-mass fission system that meets Mars-surface power requirements, while maintaining a high level of safety and reliability at a relatively low cost. The Heatpipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive space fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed-which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Heatpipe-Operated Mars Exploration Reactor (HOMER) is a derivative of the HPS designed especially for producing power on the surface of Mars. The HOMER-15 is a 15-kWt reactor that couples with a 3-kWe Stirling engine power system. The reactor contains stainless-steel (SS)-clad uranium nitride (UN) fuel pins that are structurally and thermally bonded to SS/sodium heatpipes. Fission energy is conducted from the fuel pins to the heatpipes, which then carry the heat to the Stirling engine. This paper describes the attributes, specifications, and performance of a 15-kWt HOMER reactor. .

Fission products and nuclear fuel behaviour under severe accident conditions part 1: Main lessons learnt from the first VERDON test

NASA Astrophysics Data System (ADS)

Pontillon, Y.; Geiger, E.; Le Gall, C.; Bernard, S.; Gallais-During, A.; Malgouyres, P. P.; Hanus, E.; Ducros, G.

2017-11-01

This paper describes the first VERDON test performed at the end of September 2011 with special emphasis on the behaviour of fission products (FP) and actinides during the accidental sequence itself. Two other papers discuss in detail the post-test examination results (SEM, EPMA and SIMS) of the VERDON-1 sample. The first VERDON test was devoted to studying UO2 fuel behaviour and fission product releases under reducing conditions at very high temperature (∼2883 K), which was able to confirm the very good performance of the VERDON loop. The fuel sample did not lose its integrity during this test. According to the FP behaviour measured by the online gamma station (fuel sight), the general classification of the FP in relation to their released fraction is very accurate, and the burn-up effect on the release rate is clearly highlighted.

The neutrons for science facility at SPIRAL-2

NASA Astrophysics Data System (ADS)

Ledoux, X.; Aïche, M.; Avrigeanu, M.; Avrigeanu, V.; Balanzat, E.; Ban-d'Etat, B.; Ban, G.; Bauge, E.; Bélier, G.; Bém, P.; Borcea, C.; Caillaud, T.; Chatillon, A.; Czajkowski, S.; Dessagne, P.; Doré, D.; Fischer, U.; Frégeau, M. O.; Grinyer, J.; Guillous, S.; Gunsing, F.; Gustavsson, C.; Henning, G.; Jacquot, B.; Jansson, K.; Jurado, B.; Kerveno, M.; Klix, A.; Landoas, O.; Lecolley, F. R.; Lecouey, J. L.; Majerle, M.; Marie, N.; Materna, T.; Mrázek, J.; Negoita, F.; Novák, J.; Oberstedt, S.; Oberstedt, A.; Panebianco, S.; Perrot, L.; Plompen, A. J. M.; Pomp, S.; Prokofiev, A. V.; Ramillon, J. M.; Farget, F.; Ridikas, D.; Rossé, B.; Sérot, O.; Simakov, S. P.; Šimečková, E.; Štefánik, M.; Sublet, J. C.; Taïeb, J.; Tarrío, D.; Tassan-Got, L.; Thfoin, I.; Varignon, C.

2017-09-01

Numerous domains, in fundamental research as well as in applications, require the study of reactions induced by neutrons with energies from few MeV up to few tens of MeV. Reliable measurements also are necessary to improve the evaluated databases used by nuclear transport codes. This energy range covers a large number of topics like transmutation of nuclear waste, design of future fission and fusion reactors, nuclear medicine or test and development of new detectors. A new facility called Neutrons For Science (NFS) is being built for this purpose on the GANIL site at Caen (France). NFS is composed of a pulsed neutron beam for time-of-flight facility as well as irradiation stations for cross-section measurements. Neutrons will be produced by the interaction of deuteron and proton beams, delivered by the SPIRAL-2 linear accelerator, with thick or thin converters made of beryllium or lithium. Continuous and quasi-mono-energetic spectra will be available at NFS up to 40 MeV. In this fast energy region, the neutron flux is expected to be up to 2 orders of magnitude higher than at other existing time-of-flight facilities. In addition, irradiation stations for neutron-, proton- and deuteron-induced reactions will allow performing cross-section measurements by the activation technique. After a description of the facility and its characteristics, the experiments to be performed in the short and medium term will be presented.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ledoux, X.; Bauge, E.; Belier, G.

The ''Neutrons for Science''(NFS) facility will be a component of SPIRAL-2, the future accelerator dedicated to the production of very intense radioactive ion beams, under construction at GANIL in Caen (France). NFS will be composed of a pulsed neutron beam for in-flight measurements and irradiation stations for cross-section measurements and material studies. Continuous and quasi-monokinetic energy spectra will be available at NFS respectively produced by the interaction of deuteron beam on thick a Be converter and by the {sup 7}Li(p,n) reaction on a thin converter. The flux at NFS will be up to 2 orders of magnitude higher than thosemore » of other existing time-of-flight facilities in the 1 MeV to 40 MeV range. NFS will be a very powerful tool for physics and fundamental research as well as applications like the transmutation of nuclear waste, design of future fission and fusion reactors, nuclear medicine or test and development of new detectors.« less

Photo-ionization of aluminum in a hot cavity for the selective production of exotic species project

NASA Astrophysics Data System (ADS)

Scarpa, D.; Makhathini, L.; Tomaselli, A.; Grassi, D.; Corradetti, S.; Manzolaro, M.; Vasquez, J.; Calderolla, M.; Rossignoli, M.; Monetti, A.; Andrighetto, A.; Prete, G.

2014-02-01

SPES (Selective Production of Exotic Species) is an Isotope Separation On-Line (ISOL) based accelerator facility that will be built in the Legnaro-Istituto Nazionale di Fisica Nucleare (INFN) Laboratory (Italy), intended to provide intense neutron-rich radioactive ion beams obtained by proton-induced fission of a uranium carbide (UCx) target. Besides this main target material, silicon carbide (SiC) will be the first to be used to deliver p-rich beams. This target will also validate the functionality of the SPES facility with aluminum beam as result of impinging SiC target with proton beam. In the past, off line studies on laser photoionization of aluminum have been performed in Pavia Spectroscopy Laboratory and in Laboratori Nazionali di Legnaro; a XeCl excimer laser was installed in order to test the laser ionization in the SPES hot cavity. With the new Wien filter installed a better characterization of the ionization process in terms of efficiency was performed and results are discussed.

Studies for aluminum photoionization in hot cavity for the selective production of exotic species projecta)

NASA Astrophysics Data System (ADS)

Scarpa, D.; Vasquez, J.; Tomaselli, A.; Grassi, D.; Biasetto, L.; Cavazza, A.; Corradetti, S.; Manzolaro, M.; Montano, J.; Andrighetto, A.; Prete, G.

2012-02-01

Selective production of exotic species (SPES) is an ISOL-based accelerator facility that will be built in the Legnaro INFN Laboratory (Italy), intended to provide an intense neutron-rich radioactive ion beams obtained by proton induced fission of an uranium carbide target. Beside this main target, a silicon carbide (SiC) target will the first to be used to deliver some p-rich beams. This target will validate also the functionality of the SPES facility with aluminum beam as result of hitting SiC target with protons. In the past off-line studies on laser photoionization of aluminum have performed in Pavia Spectroscopy Laboratory and in Laboratori Nazionali di Legnaro where, recently, a XeCl excimer laser was installed in order to test the laser ionization in the SPES hot cavity. Results are promising to justify further studies with this technique, aiming a better characterization of the SPES ion extraction capability under laser photoionization.

I-NERI-2007-004-K, DEVELOPMENT AND CHARACTERIZATION OF NEW HIGH-LEVEL WASTE FORMS FOR ACHIEVING WASTE MINIMIZATION FROM PYROPROCESSING

DOE Office of Scientific and Technical Information (OSTI.GOV)

S.M. Frank

Work describe in this report represents the final year activities for the 3-year International Nuclear Energy Research Initiative (I-NERI) project: Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing. Used electrorefiner salt that contained actinide chlorides and was highly loaded with surrogate fission products was processed into three candidate waste forms. The first waste form, a high-loaded ceramic waste form is a variant to the CWF produced during the treatment of Experimental Breeder Reactor-II used fuel at the Idaho National Laboratory (INL). The two other waste forms were developed by researchers at the Korean Atomicmore » Energy Research Institute (KAERI). These materials are based on a silica-alumina-phosphate matrix and a zinc/titanium oxide matrix. The proposed waste forms, and the processes to fabricate them, were designed to immobilize spent electrorefiner chloride salts containing alkali, alkaline earth, lanthanide, and halide fission products that accumulate in the salt during the processing of used nuclear fuel. This aspect of the I-NERI project was to demonstrate 'hot cell' fabrication and characterization of the proposed waste forms. The outline of the report includes the processing of the spent electrorefiner salt and the fabrication of each of the three waste forms. Also described is the characterization of the waste forms, and chemical durability testing of the material. While waste form fabrication and sample preparation for characterization must be accomplished in a radiological hot cell facility due to hazardous radioactivity levels, smaller quantities of each waste form were removed from the hot cell to perform various analyses. Characterization included density measurement, elemental analysis, x-ray diffraction, scanning electron microscopy and the Product Consistency Test, which is a leaching method to measure chemical durability. Favorable results from this demonstration project will provide additional options for fission product immobilization and waste management associated the electrochemical/pyrometallurgical processing of used nuclear fuel.« less

Fission-fragment detector for DANCE based on thin scintillating films

NASA Astrophysics Data System (ADS)

Rusev, G.; Roman, A. R.; Daum, J. K.; Springs, R. K.; Bond, E. M.; Jandel, M.; Baramsai, B.; Bredeweg, T. A.; Couture, A.; Favalli, A.; Ianakiev, K. D.; Iliev, M. L.; Mosby, S.; Ullmann, J. L.; Walker, C. L.

2015-12-01

A fission-fragment detector based on thin scintillating films has been built to serve as a trigger/veto detector in neutron-induced fission measurements at DANCE. The fissile material is surrounded by scintillating films providing 4 π detection of the fission fragments. The scintillation photons were registered with silicon photomultipliers. A measurement of the 235U (n , f) reaction with this detector at DANCE revealed a correct time-of-flight spectrum and provided an estimate for the efficiency of the prototype detector of 11.6(7)%. Design and test measurements with the detector are described.

Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

NASA Technical Reports Server (NTRS)

Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

2010-01-01

Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

Advanced Space Fission Propulsion Systems

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Borowski, Stanley K.

2010-01-01

Fission has been considered for in-space propulsion since the 1940s. Nuclear Thermal Propulsion (NTP) systems underwent extensive development from 1955-1973, completing 20 full power ground tests and achieving specific impulses nearly twice that of the best chemical propulsion systems. Space fission power systems (which may eventually enable Nuclear Electric Propulsion) have been flown in space by both the United States and the Former Soviet Union. Fission is the most developed and understood of the nuclear propulsion options (e.g. fission, fusion, antimatter, etc.), and fission has enjoyed tremendous terrestrial success for nearly 7 decades. Current space nuclear research and technology efforts are focused on devising and developing first generation systems that are safe, reliable and affordable. For propulsion, the focus is on nuclear thermal rockets that build on technologies and systems developed and tested under the Rover/NERVA and related programs from the Apollo era. NTP Affordability is achieved through use of previously developed fuels and materials, modern analytical techniques and test strategies, and development of a small engine for ground and flight technology demonstration. Initial NTP systems will be capable of achieving an Isp of 900 s at a relatively high thrust-to-weight ratio. The development and use of first generation space fission power and propulsion systems will provide new, game changing capabilities for NASA. In addition, development and use of these systems will provide the foundation for developing extremely advanced power and propulsion systems capable of routinely and affordably accessing any point in the solar system. The energy density of fissile fuel (8 x 10(exp 13) Joules/kg) is more than adequate for enabling extensive exploration and utilization of the solar system. For space fission propulsion systems, the key is converting the virtually unlimited energy of fission into thrust at the desired specific impulse and thrust-to-weight ratio. This presentation will discuss potential space fission propulsion options ranging from first generation systems to highly advanced systems. Ongoing research that shows promise for enabling second generation NTP systems with Isp greater than 1000 s will be discussed, as will the potential for liquid, gas, or plasma core systems. Space fission propulsion systems could also be used in conjunction with simple (water-based) propellant depots to enable routine, affordable missions to various destinations (e.g. moon, Mars, asteroids) once in-space infrastructure is sufficiently developed. As fuel and material technologies advance, very high performance Nuclear Electric Propulsion (NEP) systems may also become viable. These systems could enable sophisticated science missions, highly efficient cargo delivery, and human missions to numerous destinations. Commonalities between NTP, fission power systems, and NEP will be discussed.

Fission matrix-based Monte Carlo criticality analysis of fuel storage pools

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farlotti, M.; Ecole Polytechnique, Palaiseau, F 91128; Larsen, E. W.

2013-07-01

Standard Monte Carlo transport procedures experience difficulties in solving criticality problems in fuel storage pools. Because of the strong neutron absorption between fuel assemblies, source convergence can be very slow, leading to incorrect estimates of the eigenvalue and the eigenfunction. This study examines an alternative fission matrix-based Monte Carlo transport method that takes advantage of the geometry of a storage pool to overcome this difficulty. The method uses Monte Carlo transport to build (essentially) a fission matrix, which is then used to calculate the criticality and the critical flux. This method was tested using a test code on a simplemore » problem containing 8 assemblies in a square pool. The standard Monte Carlo method gave the expected eigenfunction in 5 cases out of 10, while the fission matrix method gave the expected eigenfunction in all 10 cases. In addition, the fission matrix method provides an estimate of the error in the eigenvalue and the eigenfunction, and it allows the user to control this error by running an adequate number of cycles. Because of these advantages, the fission matrix method yields a higher confidence in the results than standard Monte Carlo. We also discuss potential improvements of the method, including the potential for variance reduction techniques. (authors)« less

Initial Gamma Spectrometry Examination of the AGR-3/4 Irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

2016-11-01

The initial results from gamma spectrometry examination of the different components from the combined third and fourth US Advanced Gas Reactor Fuel Development TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail particles surrounded by two graphitic diffusion rings (inner and outer) and a graphite sink were irradiated in the Idaho National Laboratory’s Advanced Test Reactor. Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation andmore » evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. Additionally, inner and outer rings were also examined by gamma spectrometry both to evaluate the fission product inventory and the distribution of gamma-emitting fission products within the rings using gamma emission computed tomography. The cesium inventory of the scanned rings compares acceptably well with the expected inventory from fission product transport modeling. The inventory of the graphite fission product sinks is also being evaluated by gamma spectrometry.« less

Progress in Mirror-Based Fusion Neutron Source Development.

PubMed

Anikeev, A V; Bagryansky, P A; Beklemishev, A D; Ivanov, A A; Kolesnikov, E Yu; Korzhavina, M S; Korobeinikova, O A; Lizunov, A A; Maximov, V V; Murakhtin, S V; Pinzhenin, E I; Prikhodko, V V; Soldatkina, E I; Solomakhin, A L; Tsidulko, Yu A; Yakovlev, D V; Yurov, D V

2015-12-04

The Budker Institute of Nuclear Physics in worldwide collaboration has developed a project of a 14 MeV neutron source for fusion material studies and other applications. The projected neutron source of the plasma type is based on the gas dynamic trap (GDT), which is a special magnetic mirror system for plasma confinement. Essential progress in plasma parameters has been achieved in recent experiments at the GDT facility in the Budker Institute, which is a hydrogen (deuterium) prototype of the source. Stable confinement of hot-ion plasmas with the relative pressure exceeding 0.5 was demonstrated. The electron temperature was increased up to 0.9 keV in the regime with additional electron cyclotron resonance heating (ECRH) of a moderate power. These parameters are the record for axisymmetric open mirror traps. These achievements elevate the projects of a GDT-based neutron source on a higher level of competitive ability and make it possible to construct a source with parameters suitable for materials testing today. The paper presents the progress in experimental studies and numerical simulations of the mirror-based fusion neutron source and its possible applications including a fusion material test facility and a fusion-fission hybrid system.

International strategy for fusion materials development

NASA Astrophysics Data System (ADS)

Ehrlich, Karl; Bloom, E. E.; Kondo, T.

2000-12-01

In this paper, the results of an IEA-Workshop on Strategy and Planning of Fusion Materials Research and Development (R&D), held in October 1998 in Risø Denmark are summarised and further developed. Essential performance targets for materials to be used in first wall/breeding blanket components have been defined for the major materials groups under discussion: ferritic-martensitic steels, vanadium alloys and ceramic composites of the SiC/SiC-type. R&D strategies are proposed for their further development and qualification as reactor-relevant materials. The important role of existing irradiation facilities (mainly fission reactors) for materials testing within the next decade is described, and the limits for the transfer of results from such simulation experiments to fusion-relevant conditions are addressed. The importance of a fusion-relevant high-intensity neutron source for the development of structural as well as breeding and special purpose materials is elaborated and the reasons for the selection of an accelerator-driven D-Li-neutron source - the International Fusion Materials Irradiation Facility (IFMIF) - as an appropriate test bed are explained. Finally the necessity to execute the materials programme for fusion in close international collaboration, presently promoted by the International Energy Agency, IEA is emphasised.

Source Term Experiments Project (STEP): Aerosol characterization system

NASA Astrophysics Data System (ADS)

Schlenger, B. J.; Dunn, P. F.

A series of four experiments is being conducted at Argonne National Laboratory's TREAT Reactor. They were designed to provide some of the necessary data regarding magnitude and release rates of fission products from degraded fuel pins, physical and chemical characteristics of released fission products, and aerosol formation and transport phenomena. These are in pile experiments, whereby the test fuel is heated by neutron induced fission and subsequent clad oxidation in steam environments that simulate as closely as practical predicted reactor accident conditions. The test sequences cover a range of pressure and fuel heatup rate, and include the effect of Aq/In/Cd control rod material.

Developments toward Understanding and Improving the Low Energy Measurement Capabilities of a Fission Time Projection Chamber

NASA Astrophysics Data System (ADS)

Bundgaard, Jeremy J.

Nuclear physicists have been recently called upon for new, high precision fission measurements to improve existing fission models, ultimately enabling engineers to design next generation reactors as well as guarding the nation's stockpile. In response, a resurgence in fission research is aimed at developing detectors to design and build new experiments to meet these needs. The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) collaboration has developed the fission Time Projection Chamber (fissionTPC) to measure neutron induced fission with unprecedented precision. The fissionTPC is annually deployed to the Los Alamos Neutron Science Center LANSCE where it operates with a neutron beam passing axially through the drift volume, irradiating heavy actinide targets to induce fission. The fissionTPC was developed at the Lawrence Livermore National Laboratory's (LLNL) TPC lab, where it is tested with spontaneous fission (SF) from radioactive sources, typically 252Cf and 244Cm, to characterize detector response, improve performance, and evolve the design. One of the experiments relevant for both nuclear energy and nonproliferation is to measure the neutron induced fission of 239Pu, which exhibits a high alpha activity, generating a large unwanted background for the fission measurements. The ratio of alpha to fission present in our neutron induced fission measurement of 239Pu is on the same order of magnitude as the 244Cm alpha/SF branching ratio. The high alpha rate required the TPC to be triggering on fission signals during beam time and we set out to build a trigger system, which, using 244Cm to produce a similar alpha to fission ratio as 239Pu in the neutron beam, we successfully demonstrated the viability of this approach. The trigger design has been evolved for use in NIFFTE's current measurements at LANSCE. In addition to several hardware and software contributions in the development and operation of the fissionTPC, a central purpose of this thesis was also to develop analyses to demonstrate the fissionTPC's performance abilities/limitations in measuring the alpha/SF branching ratio of 252Cf and 244Cm. Our method results in benchmarking the fissionTPC's ability to produce a competitive alpha/SF ratio for 252Cf with sub-percent precision.

Neutron-induced fission-cross-section measurements and calculations of selected transplutonic isotopes

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, R.M.; Browne, J.C.

1982-08-27

The neutron-induced fission cross sections of /sup 242m/Am and /sup 245/Cm have been measured over an energy range of 10/sup -4/ eV to approx. 20 MeV in a series of experiments at three facilities during the past several years. The combined results of these measurements, in which only sub-milligram quantities of enriched isotopes were used, yield cross sections with uncertainties of approximately 5% below 10 MeV relative to the /sup 235/U standard cross section used to normalize the data. We summarize the resonance analysis of the /sup 242m/Am(n,f) cross section in the eV region. Hauser-Feshbach statistical calculations of the detailedmore » fission cross sections of /sup 235/U and /sup 245/Cm have been carried out over the energy region from 0.1 to 5 MeV and these results are compared with our experimental data.« less

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

DOE PAGES

Keiser, Jr., Dennis D.; Jue, Jan -Fong; Gan, Jian; ...

2017-02-27

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up tomore » a final temperature of 500°C. The results indicated that two types of grain/cell boundaries were observed in the U- 7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Lastly, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing.« less

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

DOE Office of Scientific and Technical Information (OSTI.GOV)

Keiser, Jr., Dennis D.; Jue, Jan -Fong; Gan, Jian

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up tomore » a final temperature of 500°C. The results indicated that two types of grain/cell boundaries were observed in the U- 7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Lastly, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing.« less

A new UK fission yield evaluation UKFY3.7

NASA Astrophysics Data System (ADS)

Mills, Robert William

2017-09-01

The JEFF neutron induced and spontaneous fission product yield evaluation is currently unchanged from JEFF-3.1.1, also known by its UK designation UKFY3.6A. It is based upon experimental data combined with empirically fitted mass, charge and isomeric state models which are then adjusted within the experimental and model uncertainties to conform to the physical constraints of the fission process. A new evaluation has been prepared for JEFF, called UKFY3.7, that incorporates new experimental data and replaces the current empirical models (multi-Gaussian fits of mass distribution and Wahl Zp model for charge distribution combined with parameter extrapolation), with predictions from GEF. The GEF model has the advantage that one set of parameters allows the prediction of many different fissioning nuclides at different excitation energies unlike previous models where each fissioning nuclide at a specific excitation energy had to be fitted individually to the relevant experimental data. The new UKFY3.7 evaluation, submitted for testing as part of JEFF-3.3, is described alongside initial results of testing. In addition, initial ideas for future developments allowing inclusion of new measurements types and changing from any neutron spectrum type to true neutron energy dependence are discussed. Also, a method is proposed to propagate uncertainties of fission product yields based upon the experimental data that underlies the fission yield evaluation. The covariance terms being determined from the evaluated cumulative and independent yields combined with the experimental uncertainties on the cumulative yield measurements.

Utilization of the Differential Die-Away Self-Interrogation Technique for Characterization and Verification of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trahan, Alexis Chanel

New nondestructive assay techniques are sought to better characterize spent nuclear fuel. One of the NDA instruments selected for possible deployment is differential die-away self-interrogation (DDSI). The proposed DDSI approach for spent fuel assembly assay utilizes primarily the spontaneous fission and (α, n) neutrons in the assemblies as an internal interrogating radiation source. The neutrons released in spontaneous fission or (α,n) reactions are thermalized in the surrounding water and induce fission in fissile isotopes, thereby creating a measurable signal from isotopes of interest that would be otherwise difficult to measure. The DDSI instrument employs neutron coincidence counting with 3He tubesmore » and list-mode-based data acquisition to allow for production of Rossi-alpha distributions (RADs) in post-processing. The list-mode approach to data collection and subsequent construction of RADs has expanded the analytical possibilities, as will be demonstrated throughout this thesis. One of the primary advantages is that the measured signal in the form of a RAD can be analyzed in its entirety including determination of die-away times in different time domains. This capability led to the development of the early die-away method, a novel leakage multiplication determination method which is tested throughout the thesis on different sources in simulation space and fresh fuel experiments. The early die-away method is a robust, accurate, improved method of determining multiplication without the need for knowledge of the (α,n) source term. The DDSI technique and instrument are presented along with the many novel capabilities enabled by and discovered through RAD analysis. Among the new capabilities presented are the early die-away method, total plutonium content determination, and highly sensitive missing pin detection. Simulation of hundreds of different spent and fresh fuel assemblies were used to develop the analysis algorithms and the techniques were tested on a variety of spontaneous fission-driven fresh fuel assemblies at Los Alamos National Laboratory and the BeRP ball at the Nevada National Security Site. The development of the new, improved analysis and characterization methods with the DDSI instrument makes it a viable technique for implementation in a facility to meet material control and safeguards needs.« less

Evaluation of the 235 U resonance parameters to fit the standard recommended values

DOE PAGES

Leal, Luiz; Noguere, Gilles; Paradela, Carlos; ...

2017-09-13

A great deal of effort has been dedicated to the revision of the standard values in connection with the neutron interaction for some actinides. While standard data compilation are available for decades nuclear data evaluations included in existing nuclear data libraries (ENDF, JEFF, JENDL, etc.) do not follow the standard recommended values. Indeed, the majority of evaluations for major actinides do not conform to the standards whatsoever. In particular, for the n + 235U interaction the only value in agreement with the standard is the thermal fission cross section. We performed a resonance re-evaluation of the n + 235U interactionmore » in order to address the issues regarding standard values in the energy range from 10-5 eV to 2250 eV. Recently, 235U fission cross-section measurements have been performed at the CERN Neutron Time-o-Flight facility (TOF), known as n_TOF, in the energy range from 0.7 eV to 10 keV. The data were normalized according to the recommended standard of the fission integral in the energy range 7.8 eV to 11 eV. As a result, the n_TOF averaged fission cross sections above 100 eV are in good agreement with the standard recommended values. The n_TOF data were included in the 235U resonance analysis that was performed with the code SAMMY. In addition to the average standard values related to the fission cross section, standard thermal values for fission, capture, and elastic cross sections were also included in the evaluation. Our paper presents the procedure used for re-evaluating the 235U resonance parameters including the recommended standard values as well as new cross section measurements.« less

Evaluation of the 235 U resonance parameters to fit the standard recommended values

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leal, Luiz; Noguere, Gilles; Paradela, Carlos

A great deal of effort has been dedicated to the revision of the standard values in connection with the neutron interaction for some actinides. While standard data compilation are available for decades nuclear data evaluations included in existing nuclear data libraries (ENDF, JEFF, JENDL, etc.) do not follow the standard recommended values. Indeed, the majority of evaluations for major actinides do not conform to the standards whatsoever. In particular, for the n + 235U interaction the only value in agreement with the standard is the thermal fission cross section. We performed a resonance re-evaluation of the n + 235U interactionmore » in order to address the issues regarding standard values in the energy range from 10-5 eV to 2250 eV. Recently, 235U fission cross-section measurements have been performed at the CERN Neutron Time-o-Flight facility (TOF), known as n_TOF, in the energy range from 0.7 eV to 10 keV. The data were normalized according to the recommended standard of the fission integral in the energy range 7.8 eV to 11 eV. As a result, the n_TOF averaged fission cross sections above 100 eV are in good agreement with the standard recommended values. The n_TOF data were included in the 235U resonance analysis that was performed with the code SAMMY. In addition to the average standard values related to the fission cross section, standard thermal values for fission, capture, and elastic cross sections were also included in the evaluation. Our paper presents the procedure used for re-evaluating the 235U resonance parameters including the recommended standard values as well as new cross section measurements.« less

Evaluation of the 235U resonance parameters to fit the standard recommended values

NASA Astrophysics Data System (ADS)

Leal, Luiz; Noguere, Gilles; Paradela, Carlos; Durán, Ignacio; Tassan-Got, Laurent; Danon, Yaron; Jandel, Marian

2017-09-01

A great deal of effort has been dedicated to the revision of the standard values in connection with the neutron interaction for some actinides. While standard data compilation are available for decades nuclear data evaluations included in existing nuclear data libraries (ENDF, JEFF, JENDL, etc.) do not follow the standard recommended values. Indeed, the majority of evaluations for major actinides do not conform to the standards whatsoever. In particular, for the n + 235U interaction the only value in agreement with the standard is the thermal fission cross section. A resonance re-evaluation of the n + 235U interaction has been performed to address the issues regarding standard values in the energy range from 10-5 eV to 2250 eV. Recently, 235U fission cross-section measurements have been performed at the CERN Neutron Time-of-Flight facility (TOF), known as n_TOF, in the energy range from 0.7 eV to 10 keV. The data were normalized according to the recommended standard of the fission integral in the energy range 7.8 eV to 11 eV. As a result, the n_TOF averaged fission cross sections above 100 eV are in good agreement with the standard recommended values. The n_TOF data were included in the 235U resonance analysis that was performed with the code SAMMY. In addition to the average standard values related to the fission cross section, standard thermal values for fission, capture, and elastic cross sections were also included in the evaluation. This paper presents the procedure used for re-evaluating the 235U resonance parameters including the recommended standard values as well as new cross section measurements.

End-to-End Demonstrator of the Safe Affordable Fission Engine (SAFE) 30: Power Conversion and Ion Engine Operation

NASA Technical Reports Server (NTRS)

Hrbud, Ivana; VanDyke, Melissa; Houts, Mike; Goodfellow, Keith; Schafer, Charles (Technical Monitor)

2001-01-01

The Safe Affordable Fission Engine (SAFE) test series addresses Phase 1 Space Fission Systems issues in particular non-nuclear testing and system integration issues leading to the testing and non-nuclear demonstration of a 400-kW fully integrated flight unit. The first part of the SAFE 30 test series demonstrated operation of the simulated nuclear core and heat pipe system. Experimental data acquired in a number of different test scenarios will validate existing computational models, demonstrated system flexibility (fast start-ups, multiple start-ups/shut downs), simulate predictable failure modes and operating environments. The objective of the second part is to demonstrate an integrated propulsion system consisting of a core, conversion system and a thruster where the system converts thermal heat into jet power. This end-to-end system demonstration sets a precedent for ground testing of nuclear electric propulsion systems. The paper describes the SAFE 30 end-to-end system demonstration and its subsystems.

Irradiation behavior of U 6Mn-Al dispersion fuel elements

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Wiencek, T. C.; Hayes, S. L.; Hofman, G. L.

2000-02-01

Irradiation testing of U 6Mn-Al dispersion fuel miniplates was conducted in the Oak Ridge Research Reactor (ORR). Post-irradiation examination showed that U 6Mn in an unrestrained plate configuration performs similarly to U 6Fe under irradiation, forming extensive and interlinked fission gas bubbles at a fission density of approximately 3×10 27 m-3. Fuel plate failure occurs by fission gas pressure driven `pillowing' on continued irradiation.

Radionuclide Basics: Cesium-137

EPA Pesticide Factsheets

The most common radioactive form of cesium (chemical symbol Cs) is Cesium-137. Cesium-137 is produced by nuclear fission for use in medical devices and gauges and is one of the byproducts of nuclear fission in nuclear reactors and nuclear weapons testing.

Mechanistic prediction of fission-gas behavior during in-cell transient heating tests on LWR fuel using the GRASS-SST and FASTGRASS computer codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rest, J; Gehl, S M

1979-01-01

GRASS-SST and FASTGRASS are mechanistic computer codes for predicting fission-gas behavior in UO/sub 2/-base fuels during steady-state and transient conditions. FASTGRASS was developed in order to satisfy the need for a fast-running alternative to GRASS-SST. Althrough based on GRASS-SST, FASTGRASS is approximately an order of magnitude quicker in execution. The GRASS-SST transient analysis has evolved through comparisons of code predictions with the fission-gas release and physical phenomena that occur during reactor operation and transient direct-electrical-heating (DEH) testing of irradiated light-water reactor fuel. The FASTGRASS calculational procedure is described in this paper, along with models of key physical processes included inmore » both FASTGRASS and GRASS-SST. Predictions of fission-gas release obtained from GRASS-SST and FASTGRASS analyses are compared with experimental observations from a series of DEH tests. The major conclusions is that the computer codes should include an improved model for the evolution of the grain-edge porosity.« less

Neutron Imaging at LANSCE—From Cold to Ultrafast

DOE PAGES

Nelson, Ronald Owen; Vogel, Sven C.; Hunter, James F.; ...

2018-02-23

In recent years, neutron radiography and tomography have been applied at different beam lines at Los Alamos Neutron Science Center (LANSCE), covering a very wide neutron energy range. The field of energy-resolved neutron imaging with epi-thermal neutrons, utilizing neutron absorption resonances for contrast as well as quantitative density measurements, was pioneered at the Target 1 (Lujan center), Flight Path 5 beam line and continues to be refined. Applications include: imaging of metallic and ceramic nuclear fuels, fission gas measurements, tomography of fossils and studies of dopants in scintillators. The technique provides the ability to characterize materials opaque to thermal neutronsmore » and to utilize neutron resonance analysis codes to quantify isotopes to within 0.1 atom %. The latter also allows measuring fuel enrichment levels or the pressure of fission gas remotely. More recently, the cold neutron spectrum at the ASTERIX beam line, also located at Target 1, was used to demonstrate phase contrast imaging with pulsed neutrons. This extends the capabilities for imaging of thin and transparent materials at LANSCE. In contrast, high-energy neutron imaging at LANSCE, using unmoderated fast spallation neutrons from Target 4 [Weapons Neutron Research (WNR) facility] has been developed for applications in imaging of dense, thick objects. Using fast (ns), time-of-flight imaging, enables testing and developing imaging at specific, selected MeV neutron energies. The 4FP-60R beam line has been reconfigured with increased shielding and new, larger collimation dedicated to fast neutron imaging. The exploration of ways in which pulsed neutron beams and the time-of-flight method can provide additional benefits is continuing. We will describe the facilities and instruments, present application examples and recent results of all these efforts at LANSCE.« less

Neutron Imaging at LANSCE—From Cold to Ultrafast

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nelson, Ronald Owen; Vogel, Sven C.; Hunter, James F.

In recent years, neutron radiography and tomography have been applied at different beam lines at Los Alamos Neutron Science Center (LANSCE), covering a very wide neutron energy range. The field of energy-resolved neutron imaging with epi-thermal neutrons, utilizing neutron absorption resonances for contrast as well as quantitative density measurements, was pioneered at the Target 1 (Lujan center), Flight Path 5 beam line and continues to be refined. Applications include: imaging of metallic and ceramic nuclear fuels, fission gas measurements, tomography of fossils and studies of dopants in scintillators. The technique provides the ability to characterize materials opaque to thermal neutronsmore » and to utilize neutron resonance analysis codes to quantify isotopes to within 0.1 atom %. The latter also allows measuring fuel enrichment levels or the pressure of fission gas remotely. More recently, the cold neutron spectrum at the ASTERIX beam line, also located at Target 1, was used to demonstrate phase contrast imaging with pulsed neutrons. This extends the capabilities for imaging of thin and transparent materials at LANSCE. In contrast, high-energy neutron imaging at LANSCE, using unmoderated fast spallation neutrons from Target 4 [Weapons Neutron Research (WNR) facility] has been developed for applications in imaging of dense, thick objects. Using fast (ns), time-of-flight imaging, enables testing and developing imaging at specific, selected MeV neutron energies. The 4FP-60R beam line has been reconfigured with increased shielding and new, larger collimation dedicated to fast neutron imaging. The exploration of ways in which pulsed neutron beams and the time-of-flight method can provide additional benefits is continuing. We will describe the facilities and instruments, present application examples and recent results of all these efforts at LANSCE.« less

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2010 CFR

2010-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2013 CFR

2013-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2014 CFR

2014-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2012 CFR

2012-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2011 CFR

2011-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

Posttest calculations of bundle quench test CORA-13 with ATHLET-CD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bestele, J.; Trambauer, K.; Schubert, J.D.

Gesellschaft fuer Anlagen- und Reaktorsicherheit is developing, in cooperation with the Institut fuer Kernenergetik und Energiesysteme, Stuttgart, the system code Analysis of Thermalhydraulics of Leaks and Transients with Core Degradation (ATHLET-CD). The code consists of detailed models of the thermal hydraulics of the reactor coolant system. This thermo-fluid dynamics module is coupled with modules describing the early phase of the core degradation, like cladding deformation, oxidation and melt relocation, and the release and transport of fission products. The assessment of the code is being done by the analysis of separate effect tests, integral tests, and plant events. The code willmore » be applied to the verification of severe accident management procedures. The out-of-pile test CORA-13 was conducted by Forschungszentrum Karlsruhe in their CORA test facility. The test consisted of two phases, a heatup phase and a quench phase. At the beginning of the quench phase, a sharp peak in the hydrogen generation rate was observed. Both phases of the test have been calculated with the system code ATHLET-CD. Special efforts have been made to simulate the heat losses and the flow distribution in the test facility and the thermal hydraulics during the quench phase. In addition to previous calculations, the material relocation and the quench phase have been modeled. The temperature increase during the heatup phase, the starting time of the temperature escalation, and the maximum temperatures have been calculated correctly. At the beginning of the quench phase, an increased hydrogen generation rate has been calculated as measured in the experiment.« less

Measurement of Fission Neutron Spectrum and Multiplicity using a Gamma Tag Double Time-of-flight Setup

NASA Astrophysics Data System (ADS)

Blain, E.; Daskalakis, A.; Danon, Y.

2014-05-01

Recent efforts have been made to improve the prompt fission neutron spectrum and nu-bar measurements for Uranium and Plutonium isotopes particularly in the keV region. A system has been designed at Rensselaer Polytechnic Institute (RPI) utilizing an array of EJ-301 liquid scintillators as well as lithium glass and plastic scintillators to experimentally determine these values. An array of BaF2 detectors was recently obtained from Oak Ridge National Laboratory to be used in conjunction with the neutron detectors. The system uses a novel gamma tagging method for fission which can offer an improvement over conventional fission chambers due to increased sample mass. A coincidence requirement on the gamma detectors from prompt fission gammas is used as the fission tag for the system as opposed to fission fragments in a conventional fission chamber. The system utilizes pulse digitization using Acqiris 8 bit digitizer boards which allow for gamma/neutron pulse height discrimination on the liquid scintillators during post processing. Additionally, a 252Cf fission chamber was designed and constructed at RPI which allowed for optimization and testing of the system without the need for an external neutron source. The characteristics of the gamma tagging method such as false detection rate and detection efficiency were determined using this fission chamber and verified using MCNP Polimi modeling. Prompt fission neutron spectrum data has been taken using the fission chamber focusing on the minimum detectable neutron energy for each of the various detectors. Plastic scintillators were found to offer a significant improvement over traditional liquid scintillators allowing energy measurements down to 50 keV. Background was also characterized for all detectors and will be discussed.

High temperature UF6 RF plasma experiments applicable to uranium plasma core reactors

NASA Technical Reports Server (NTRS)

Roman, W. C.

1979-01-01

An investigation was conducted using a 1.2 MW RF induction heater facility to aid in developing the technology necessary for designing a self critical fissioning uranium plasma core reactor. Pure, high temperature uranium hexafluoride (UF6) was injected into an argon fluid mechanically confined, steady state, RF heated plasma while employing different exhaust systems and diagnostic techniques to simulate and investigate some potential characteristics of uranium plasma core nuclear reactors. The development of techniques and equipment for fluid mechanical confinement of RF heated uranium plasmas with a high density of uranium vapor within the plasma, while simultaneously minimizing deposition of uranium and uranium compounds on the test chamber peripheral wall, endwall surfaces, and primary exhaust ducts, is discussed. The material tests and handling techniques suitable for use with high temperature, high pressure, gaseous UF6 are described and the development of complementary diagnostic instrumentation and measurement techniques to characterize the uranium plasma, effluent exhaust gases, and residue deposited on the test chamber and exhaust system components is reported.

Stainless Steel NaK Circuit Integration and Fill Submission

NASA Technical Reports Server (NTRS)

Garber, Anne E.

2006-01-01

The Early Flight Fission Test Facilities (EFF-TF) team has been tasked by the Marshall Space Flight Center Nuclear Systems Office to design, fabricate, and test an actively pumped alkali metal flow circuit. The system, which was originally designed to hold a eutectic mixture of sodium potassium (NaK), was redesigned to hold lithium; but due to a shift in focus, it is once again being prepared for use with NaK. Changes made to the actively pumped, high temperature loop include the replacement of the expansion reservoir, addition of remotely operated valves, and modification of the support table. Basic circuit components include: reactor segment, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and a spill reservoir. A 37-pin partial-array core (pin and flow path dimensions are the same as those in a full design) was selected for fabrication and test. This document summarizes the integration and fill of the pumped liquid metal NaK flow circuit.

Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Forsbacka, Matthew

2004-01-01

For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the NASA Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. "virtual" reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of van'ous core deformations. The power delivered to the SAFE-100 prototype was then dusted accordingly via kinetics calculations, The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kilowatt(sub t), to 10 kilowatts(sub t), held approximately constant at 10 kilowatts (sub t), and then allowed to decrease based on the negative thermal reactivity coefficient.

Picosecond resolution on relativistic heavy ions' time-of-flight measurement

NASA Astrophysics Data System (ADS)

Ebran, A.; Taieb, J.; Belier, G.; Chatillon, A.; Laurent, B.; Martin, J.-F.; Pellereau, E.

2013-11-01

We developed a time-of-flight measurement system for relativistic heavy ions with a requested resolution of 40 ps Full Width Half Maximum. Such a resolution is mandatory to assign the correct mass number to every fission fragment, identified using the Bρ-ToF-ΔE method with the recoil spectrometer designed for the SOFIA experiment-which hold very recently at GSI. To achieve such a performance, fast plastic scintillators read-out by dedicated photomultiplier tubes were chosen among other possible options. We have led several test-measurements from 2009 to 2011, in order to investigate: the effect of the addition of a quenching molecule in the scintillator's matrix, the influence of the detector's size and the impact of the photomultiplier tube. The contribution of the dedicated electronics is also characterized. Time-of-flight measurements were performed realized with electron pulses and relativistic heavy ions, respectively provided by the LASER driven electron-accelerator (ELSA) at CEA-DAM Ile-de-France and by the SIS18/FRS facility at GSI. The reported results exhibit a time resolution better than 20 ps Full Width Half Maximum reached with the last prototype at GSI with an Uranium beam. These results confirm that the SOFIA experiment should enable the measurement of the relativistic fission fragments' time-of-flight with the requested resolution.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Panebianco, S.; Dore, D.; Giomataris, I.

Time Projection Chambers are widely used since many years for tracking and identification of charged particles in high energy physics. We present a new R and D project to investigate the feasibility of a Micromegas TPC for low energy heavy ions detection. Two physics cases are relevant for this project. The first is the study of the nuclear fission of actinides by measuring the fission fragments properties (mass, nuclear charge, kinetic energy) that will be performed at different installations and in particular at the NFS facility to be built in the framework of the SPIRAL2 project in GANIL. The secondmore » physics case is the study of heavy ion reactions, like ({alpha},{gamma}), ({alpha},p), ({alpha},n) and all the inverse reactions in the energy range between 1.5 and 3 AMeV using both stable and radioactive beams. These reactions have a key role in p process in nuclear astrophysics to explain the synthesis of heavy proton-rich nuclei. Within the project, a large effort is devoted to Monte-Carlo simulations and a detailed benchmark of different simulation codes on the energy loss and range in gas of heavy ions at low energy has been performed. A new approach for simulating the ion charge state evolution in GEANT4 is also presented. Finally, preliminary results of an experimental test campaign on prototype are discussed.« less

Thermal neutron flux measurement using self-powered neutron detector (SPND) at out-core locations of TRIGA PUSPATI Reactor (RTP)

NASA Astrophysics Data System (ADS)

Ali, Nur Syazwani Mohd; Hamzah, Khaidzir; Mohamad Idris, Faridah; Hairie Rabir, Mohamad

2018-01-01

The thermal neutron flux measurement has been conducted at the out-core location using self-powered neutron detectors (SPNDs). This work represents the first attempt to study SPNDs as neutron flux sensor for developing the fault detection system (FDS) focusing on neutron flux parameters. The study was conducted to test the reliability of the SPND’s signal by measuring the neutron flux through the interaction between neutrons and emitter materials of the SPNDs. Three SPNDs were used to measure the flux at four different radial locations which located at the fission chamber cylinder, 10cm above graphite reflector, between graphite reflector and tank liner and fuel rack. The measurements were conducted at 750 kW reactor power. The outputs from SPNDs were collected through data acquisition system and were corrected to obtain the actual neutron flux due to delayed responses from SPNDs. The measurements showed that thermal neutron flux between fission chamber location near to the tank liner and fuel rack were between 5.18 × 1011 nv to 8.45 × 109 nv. The average thermal neutron flux showed a good agreement with those from previous studies that has been made using simulation at the same core configuration at the nearest irradiation facilities with detector locations.

First steps towards real-time radiography at the NECTAR facility

NASA Astrophysics Data System (ADS)

Bücherl, T.; Wagner, F. M.; v. Gostomski, Ch. Lierse

2009-06-01

The beam tube SR10 at Forschungsneutronenquelle Heinz Maier-Leibnitz (FRM II) provides an intense beam of fission neutrons for medical application (MEDAPP) and for radiography and tomography of technical and other objects (NECTAR). The high neutron flux of up to 9.8E+07 cm -2 s -1 (depending on filters and collimation) with a mean energy of about 1.9 MeV at the sample position at the NECTAR facility prompted an experimental feasibility study to investigate the potential for real-time (RT) radiography.

Nuclear Design of the HOMER-15 Mars Surface Fission Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poston, David I.

2002-07-01

The next generation of robotic missions to Mars will most likely require robust power sources in the range of 3 to 20 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The goal of this study is to design a compact, low-mass fission system that meets Mars surface power requirements, while maintaining a high level of safety and reliability at a relatively low cost. The Heat pipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive spacemore » fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed - which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Heat pipe-Operated Mars Exploration Reactor (HOMER) is a derivative of the HPS designed especially for producing power on the surface of Mars. The HOMER-15 is a 15-kWt reactor that couples with a 3-kWe Stirling engine power system. The reactor contains stainless-steel (SS)-clad uranium nitride (UN) fuel pins that are structurally and thermally bonded to SS/sodium heat pipes. Fission energy is conducted from the fuel pins to the heat pipes, which then carry the heat to the Stirling engine. This paper describes conceptual design and nuclear performance the HOMER-15 reactor. (author)« less

Electron-beam-driven RI separator for SCRIT (ERIS) at RIKEN RI beam factory

NASA Astrophysics Data System (ADS)

Ohnishi, T.; Ichikawa, S.; Koizumi, K.; Kurita, K.; Miyashita, Y.; Ogawara, R.; Tamaki, S.; Togasaki, M.; Wakasugi, M.

2013-12-01

We constructed a radioactive isotope (RI) separator named ERIS (electron-beam-driven RI separator for SCRIT) for the SCRIT (Self-Confinement RI Target) electron scattering facility at RIKEN RI Beam Factory (RIBF). In ERIS, production rate of fission products in the photofission of uranium is estimated to be 2.2 ×1011 fissions/s with 30 g of uranium and a 1-kW electron beam. During the commissioning of ERIS, the mass resolution and overall efficiency, including ionization, extraction, and transmission, were found to be 1660 and 21%, respectively, using natural xenon gas. The preparation of uranium carbide (UC2) RI production targets is described from which a 132Sn beam was successfully separated in our first attempt at RI production.

New statistical scission-point model to predict fission fragment observables

NASA Astrophysics Data System (ADS)

Lemaître, Jean-François; Panebianco, Stefano; Sida, Jean-Luc; Hilaire, Stéphane; Heinrich, Sophie

2015-09-01

The development of high performance computing facilities makes possible a massive production of nuclear data in a full microscopic framework. Taking advantage of the individual potential calculations of more than 7000 nuclei, a new statistical scission-point model, called SPY, has been developed. It gives access to the absolute available energy at the scission point, which allows the use of a parameter-free microcanonical statistical description to calculate the distributions and the mean values of all fission observables. SPY uses the richness of microscopy in a rather simple theoretical framework, without any parameter except the scission-point definition, to draw clear answers based on perfect knowledge of the ingredients involved in the model, with very limited computing cost.

Actinide Sputtering Induced by Fission with Ultra-cold Neutrons

NASA Astrophysics Data System (ADS)

Shi, Tan; Venuti, Michael; Fellers, Deion; Martin, Sean; Morris, Chris; Makela, Mark

2017-09-01

Understanding the effects of actinide sputtering due to nuclear fission is important for a wide range of applications, including nuclear fuel storage, space science, and national defense. A new program at the Los Alamos Neutron Science Center uses ultracold neutrons (UCN) to induce fission in actinides such as uranium and plutonium. By controlling the UCN energy, it is possible to induce fission at the sample surface within a well-defined depth. It is therefore an ideal tool for studying the effects of fission-induced sputtering as a function of interaction depth. Since the mechanism for fission-induced surface damage is not well understood, this work has the potential to deconvolve the various damage mechanisms. During the irradiation with UCN, NaI detectors are used to monitor the fission events and were calibrated by monitoring fission fragments with an organic scintillator. Alpha spectroscopy of the ejected actinide material is performed in an ion chamber to determine the amount of sputtered material. Actinide samples with various sample properties and surface conditions are irradiated and analyzed. In this talk, I will discuss our experimental setup and present the preliminary results from the testing of multiple samples. This work has been supported by Los Alamos National Laboratory and Seaborg Summer Research Fellowship.

Development of prototype induced-fission-based Pu accountancy instrument for safeguards applications.

PubMed

Seo, Hee; Lee, Seung Kyu; An, Su Jung; Park, Se-Hwan; Ku, Jeong-Hoe; Menlove, Howard O; Rael, Carlos D; LaFleur, Adrienne M; Browne, Michael C

2016-09-01

Prototype safeguards instrument for nuclear material accountancy (NMA) of uranium/transuranic (U/TRU) products that could be produced in a future advanced PWR fuel processing facility has been developed and characterized. This is a new, hybrid neutron measurement system based on fast neutron energy multiplication (FNEM) and passive neutron albedo reactivity (PNAR) methods. The FNEM method is sensitive to the induced fission rate by fast neutrons, while the PNAR method is sensitive to the induced fission rate by thermal neutrons in the sample to be measured. The induced fission rate is proportional to the total amount of fissile material, especially plutonium (Pu), in the U/TRU product; hence, the Pu amount can be calibrated as a function of the induced fission rate, which can be measured using either the FNEM or PNAR method. In the present study, the prototype system was built using six (3)He tubes, and its performance was evaluated for various detector parameters including high-voltage (HV) plateau, efficiency profiles, dead time, and stability. The system's capability to measure the difference in the average neutron energy for the FNEM signature also was evaluated, using AmLi, PuBe, (252)Cf, as well as four Pu-oxide sources each with a different impurity (Al, F, Mg, and B) and producing (α,n) neutrons with different average energies. Future work will measure the hybrid signature (i.e., FNEM×PNAR) for a Pu source with an external interrogating neutron source after enlarging the cavity size of the prototype system to accommodate a large-size Pu source (~600g Pu). Copyright © 2016 Elsevier Ltd. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lane, Taylor; Parma, Edward J.

Delayed fission gamma-rays play an important role in determining the time dependent ioniz- ing dose for experiments in the central irradiation cavity of the Annular Core Research Reactor (ACRR). Delayed gamma-rays are produced from both fission product decay and from acti- vation of materials in the core, such as cladding and support structures. Knowing both the delayed gamma-ray emission rate and the time-dependent gamma-ray energy spectrum is nec- essary in order to properly determine the dose contributions from delayed fission gamma-rays. This information is especially important when attempting to deconvolute the time-dependent neutron, prompt gamma-ray, and delayed gamma-ray contribution tomore » the response of a diamond photo-conducting diode (PCD) or fission chamber in time frames of milliseconds to seconds following a reactor pulse. This work focused on investigating delayed gamma-ray character- istics produced from fission products from thermal, fast, and high energy fission of Th-232, U-233, U-235, U-238, and Pu-239. This work uses a modified version of CINDER2008, a transmutation code developed at Los Alamos National Laboratory, to model time and energy dependent photon characteristics due to fission. This modified code adds the capability to track photon-induced transmutations, photo-fission, and the subsequent radiation caused by fission products due to photo-fission. The data is compared against previous work done with SNL- modified CINDER2008 [ 1 ] and experimental data [ 2 , 3 ] and other published literature, includ- ing ENDF/B-VII.1 [ 4 ]. The ability to produce a high-fidelity (7,428 group) energy-dependent photon fluence at various times post-fission can improve the delayed photon characterization for radiation effects tests at research reactors, as well as other applications.« less

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions.

PubMed

Geslot, B; Vermeeren, L; Filliatre, P; Lopez, A Legrand; Barbot, L; Jammes, C; Bréaud, S; Oriol, L; Villard, J-F

2011-03-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 10(20) n∕cm(2). A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

NASA Astrophysics Data System (ADS)

Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

2011-03-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

Overview of experimental support for fission-product transport analyses at Oak Ridge National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wichner, R.P.

The program was designed to determine fission product and aerosol release rates from irradiated fuel under accident conditions, to identify the chemical forms of the released material, and to correlate the results with experimental and specimen conditions with the data from related experiments. These tests of PWR fuel were conducted and fuel specimen and test operating data are presented. The nature and rate of fission product vapor interaction with aerosols were studied. Aerosol deposition rates and transport in the reactor vessel during LWR core-melt accidents were studied. The Nuclear Safety Pilot Plant is dedicated to developing an expanded data basemore » on the behavior of aerosols generated during a severe accident.« less

First-Principle Characterization for Singlet Fission Couplings.

PubMed

Yang, Chou-Hsun; Hsu, Chao-Ping

2015-05-21

The electronic coupling for singlet fission, an important parameter for determining the rate, has been found to be too small unless charge-transfer (CT) components were introduced in the diabatic states, mostly through perturbation or a model Hamiltonian. In the present work, the fragment spin difference (FSD) scheme was generalized to calculate the singlet fission coupling. The largest coupling strength obtained was 14.8 meV for two pentacenes in a crystal structure, or 33.7 meV for a transition-state structure, which yielded a singlet fission lifetime of 239 or 37 fs, generally consistent with experimental results (80 fs). Test results with other polyacene molecules are similar. We found that the charge on one fragment in the S1 diabatic state correlates well with FSD coupling, indicating the importance of the CT component. The FSD approach is a useful first-principle method for singlet fission coupling, without the need to include the CT component explicitly.

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

NASA Astrophysics Data System (ADS)

Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

2017-08-01

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (<20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. Different methods have been employed to fabricate monolithic fuel plates, including hot-rolling with no cold-rolling. L1P09T is a hot-rolled fuel plate irradiated to high fission density in the RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

Acoustic sensors for fission gas characterization: R and D skills devoted to innovative instrumentation in MTR, non-destructive devices in hot lab facilities and specific transducers for measurements of LWR rods in nuclear plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ferrandis, J.Y.; Leveque, G.; Rosenkrantz, E.

2015-07-01

First of all, we will present the main principle of the method. A piezoelectric transducer, driven by a pulse generator, generates the acoustic waves in a cavity that may be the fuel rod or a chamber connected to an instrumented rod. The composition determination consists in measuring the time of flight of the acoustic signal emitted. The pressure can be estimated by a calibration process, above the measurement of the amplitude of the signal. Two projects will then be detailed. The first project consists in the development of advanced instrumentation for in-pile experiments in Material Testing Reactor. It constitutes amore » main goal for the improvement of the nuclear fuel behavior knowledge. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). As a first step of the development program, we performed in-pile tests on the most sensitive component, i.e., the piezoelectric transducer. For this purpose, the active part of this sensor has been qualified on gamma and neutron radiations and at high temperature. Various industrial piezo-ceramics were exposed to a high activity Cobalt source for few days. The cumulated dose was ranged from 50 kGy up to 2 MGy. Next, these devices were placed inside a Material Test Reactor to investigate their reliability towards neutron fluence. The final fluence after 150 days of irradiation was up to 1.6.10{sup 21}n/cm{sup 2} (for thermal neutron). Irreversible variations have been measured. Next, a specific sensor has been implemented on an instrumented fuel rod and tested in the frame of a REMORA 3 Irradiation test. It was the first experiment under high mixed, temperature neutron and gamma flux. A first irradiation phase took place in March 2010 in the OSIRIS reactor and in November 2010 for the second step of the irradiation. The instrumented fuel rod incorporating the ultrasonic gas composition sensor was finally irradiated during 2 weeks in nominal conditions. Neutronics calculation will be performed in order to calculate the thermal and fast neutron fluence and the gamma dose absorbed by acoustic sensor. A first evaluation gives a thermal fluence about 4,5.10{sup 19} n/cm{sup 2}, a fast neutrons fluence about 4,5.1018 n/cm{sup 2} and a total gamma dose up to 0,25 MGy The maximal temperature during the irradiation test was about 150 C. Although the ultrasonic sensor appears to be damaged, the optimization of the electrical attack parameters and the development of a new signal processing maintain the measurement feasibility up the end of the irradiation campaign. It was the first time that the composition of fission gas has been monitored all along an irradiation experiment in a MTR, giving access to the gas release kinetics. New researches involve thick film transducers produced by screen-printing process in order to propose piezoelectric structures for harsh temperature and irradiation measurements. The second project consists in the development of a non-destructive device that can be directly applied on a LWR fuel rod. The problem to be solved relates to the measurement of the fission gas pressure and composition in a fuel rod using a non-destructive method. Fuel rod internal pressure is one of the safety criteria applied in nuclear power analyses. This criterion must be verified in order to avoid any fuel-cladding gap reopening risk and therefore any local clad ballooning. Apart from the safety implications, this parameter is also a fuel behaviour indicator and reflects the overall fuel performance in operation, but also during shipping and long-term storage. Rod internal pressure is one criterion amongst others, like cladding corrosion, against which the acceptable fuel burn-up limit is set. A sensor has been achieved in 2007. A full-scale hot cell test of the internal gas pressure and composition measurement by an acoustic sensor was conducted successfully between 2008 and 2010 on 5 high burn-up MOX fuel rods and 2 very high burn-up UO{sub 2} fuel rods in LECA Facility at Cadarache Centre. An improvement of this sensor has been proposed, allowing us to divide by two the uncertainty on the pressure measurement. In the case of hot-cell measurements, viscous liquid can be used to couple the sensor with the rod. For gas content with a pressure exceeding 15 bars and a 10% Xe/Kr ratio, such coupling may reduce relative acoustic method accuracy by ±7% for pressure measurement result and ±0.25 % for the assessment of gas composition. These results make it possible to demonstrate the feasibility of the technique on LWR fuel rods. The transducer and the associated methodology are now operational for non-destructive measurements in hot lab facilities and allow characterising the fission gas without puncturing the fuel rods. Up to now, any other non-destructive method can be proposed. A next step will be the development of an industrial application in a fuel storage pool in order to perform a large number of measurements on a fuel assembly in nuclear plants.« less

Effects of Irradiation on the Microstructure of U-7Mo Dispersion Fuel with Al-2Si Matrix

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dennis D. Keiser, Jr.; Jan-Fong Jue; Adam B. Robinson

2012-06-01

The Reduced Enrichment for Research and Test Reactor program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt% Si added to the matrix, fuel plates were tested to medium burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fissionmore » rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, high fission rate) was performed in the RERTR-9A, RERTR-9B and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth of the fuel/matrix interaction layer (FMI), which was present in the samples to some degree after fabrication, during irradiation; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation more Si diffuses from the matrix to the FMI layer/matrix interface, and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.« less

Possible consequences of operation with KIVN fuel elements in K Zircaloy process tubes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carlson, P.A.

1963-08-06

From considerations of the results of experimental simulations of non-axial placement of fuel elements in process tubes and in-reactor experience, it is concluded that the ultimate outcome of a charging error which results in operation with one or more unsupported fuel elements in a K Zircaloy-2 process tube would be multiple fuel failure and failure of the process tube. The outcome of the accident is determined by the speed with which the fuel failure is detected and the reactor is shut down. The release of fission products would be expected to be no greater than that which has occurred followingmore » severe fuel failure incidents. The highest probability for fission product release occurs during the discharge of failed fuel elements, when a small fraction of the exposed uranium of the fuel element may be oxidized when exposed to air before the element falls into the water-filled discharge chute. The confinement and fog spray facilities were installed to reduce the amount of fission products which might escape from the reactor building after such an event.« less

Progress on the chemical separation of fission fragments from 236Np produced by proton irradiation of natural uranium target

NASA Astrophysics Data System (ADS)

Larijani, C.; Jerome, S. M.; Lorusso, G.; Ivanov, P.; Russell, B.; Pearce, A. K.; Regan, P. H.

2017-11-01

The aim of the current work is to develop and validate a radiochemical separation scheme capable of separating both 236gNp and 236Pu from a uranium target of natural isotopic composition ( 1 g uranium) and 200 MBq of fission decay products. A target containing 1.2 g of UO2 was irradiated with a beam of 25 MeV protons with a typical beam current of 30 μA for 19 h in December 2013 at the University of Birmingham Cyclotron facility. Using literature values for the production cross-section for fusion of protons with uranium targets, we estimate that an upper limit of approximately 250 Bq of activity from the 236Np ground state was produced in this experiment. Using a radiochemical separation scheme, Np and Pu fractions were separated from the produced fission decay products, with analyses of the target-based final reaction products made using Inductively Couple Plasma Mass Spectrometry (ICP-MS) and high-resolution α particle and γ-ray spectrometry.

Potential impact of releases from a new Molybdenum-99 production facility on regional measurements of airborne xenon isotopes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bowyer, Ted W.; Eslinger, Paul W.; Cameron, Ian M.

2014-03-01

The monitoring of the radioactive xenon isotopes 131mXe, 133Xe, 133mXe, and 135Xe is important for the detection of nuclear explosions. While backgrounds of the xenon isotopes are short-lived, they are constantly replenished from activities dominated by the fission-based production of 99Mo used for medical procedures. One of the most critical locations on earth for the monitoring of nuclear explosions is the Korean peninsula, where the Democratic Republic of North Korea (DPRK) has announced that it had conducted three nuclear tests between 2009 and 2013. This paper explores the backgrounds that would be caused by the medium to large scale productionmore » of 99Mo in the region of the Korean peninsula.« less

Powder Processing of High Temperature Cermets and Carbides at Marshall Space Flight Center

NASA Technical Reports Server (NTRS)

Salvail, Pat; Panda, Binayak; Hickman, Robert R.

2007-01-01

The Materials and Processing Laboratory at NASA Marshall Space Flight Center is developing Powder Metallurgy (PM) processing techniques for high temperature cermet and carbide material consolidation. These new group of materials would be utilized in the nuclear core for Nuclear Thermal Rockets (NTR). Cermet materials offer several advantages for NTR such as retention of fission products and fuels, better thermal shock resistance, hydrogen compatibility, high thermal conductivity, and high strength. Carbide materials offer the highest operating temperatures but are sensitive to thermal stresses and are difficult to process. To support the effort, a new facility has been setup to process refractory metal, ceramic, carbides and depleted uranium-based powders. The facility inciudes inert atmosphere glove boxes for the handling of reactive powders, a high temperature furnace, and powder processing equipment used for blending, milling, and sieving. The effort is focused on basic research to identify the most promising compositions and processing techniques. Several PM processing methods including Cold and Hot Isostatic Pressing are being evaluated to fabricate samples for characterization and hot hydrogen testing.

Saclay Compact Accelerator-driven Neutron Sources (SCANS)

NASA Astrophysics Data System (ADS)

Marchix, A.; Letourneau, A.; Tran, HN; Chauvin, N.; Menelle, A.; Ott, F.; Schwindling, J.

2018-06-01

For next decade, the European neutron scattering community will face of important changes, as many facilities will close, strictly fission-based sources. This statement mainly concerns France with the planned closure of Orphee and ILL. At CEA-Saclay, the project SONATE has been launched in order to provide a high intensity neutron source in Saclay site, this project is based on Compact Accelerator-driven Neutron Sources technology coupled to high-intensity beams. The goal of SONATE is to develop a 50 kW target, aiming to produce at least a neutron yield of 1013 s-1 in pulse mode with a peak current of 100 mA. We have investigated in this document the best combinations of beam/target which would lead to this substantial neutron yields. Further investigations and tests have to be carry out, especially due to sparse data on thick target and such low-energy beams considered in this document. An intermediate step to the SONATE project is under test and development, called IPHI-NEUTRON, which would lead to provide a small-size neutron facility mainly devoted to neutron imagery for industry. This step is based on the existing 3 MeV proton beam, named IPHI. Best target candidates are Lithium and Beryllium, leading respectively to a neutron yield of about 2.1013 s-1 and 4.1012 s-1.

Non-destructive Assay Measurements Using the RPI Lead Slowing Down Spectrometer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Becker, Bjorn; Weltz, Adam; Kulisek, Jonathan A.

2013-10-01

The use of a Lead Slowing-Down Spectrometer (LSDS) is consid- ered as a possible option for non-destructive assay of fissile material of used nuclear fuel. The primary objective is to quantify the 239Pu and 235U fissile content via a direct measurement, distinguishing them through their characteristic fission spectra in the LSDS. In this pa- per, we present several assay measurements performed at the Rensse- laer Polytechnic Institute (RPI) to demonstrate the feasibility of such a method and to provide benchmark experiments for Monte Carlo cal- culations of the assay system. A fresh UOX fuel rod from the RPI Criticality Researchmore » Facility, a 239PuBe source and several highly en- riched 235U discs were assayed in the LSDS. The characteristic fission spectra were measured with 238U and 232Th threshold fission cham- bers, which are only sensitive to fission neutron with energy above the threshold. Despite the constant neutron and gamma background from the PuBe source and the intense interrogation neutron flux, the LSDS system was able to measure the characteristic 235U and 239Pu responses. All measurements were compared to Monte Carlo simula- tions. It was shown that the available simulation tools and models are well suited to simulate the assay, and that it is possible to calculate the absolute count rate in all investigated cases.« less

Development of position-sensitive time-of-flight spectrometer for fission fragment research

DOE PAGES

Arnold, C. W.; Tovesson, F.; Meierbachtol, K.; ...

2014-07-09

A position-sensitive, high-resolution time-of-flight detector for fission fragments has been developed. The SPectrometer for Ion DEtermination in fission Research (SPIDER) is a 2E–2v spectrometer designed to measure the mass of light fission fragments to a single mass unit. The time pick-off detector pairs to be used in SPIDER have been tested with α-particles from 229Th and its decay chain and α-particles and spontaneous fission fragments from 252Cf. Each detector module is comprised of thin electron conversion foil, electrostatic mirror, microchannel plates, and delay-line anodes. Particle trajectories on the order of 700 mm are determined accurately to within 0.7 mm. Flightmore » times were measured with 250 ps resolution FWHM. Computed particle velocities are accurate to within 0.06 mm/ns corresponding to a precision of 0.5%. As a result, an ionization chamber capable of 400 keV energy resolution coupled with the velocity measurements described here will pave the way for modestly efficient measurements of light fission fragments with unit mass resolution.« less

Highly Competitive Reindeer Males Control Female Behavior during the Rut

PubMed Central

Body, Guillaume; Weladji, Robert B.; Holand, Øystein; Nieminen, Mauri

2014-01-01

During the rut, female ungulates move among harems or territories, either to sample mates or to avoid harassment. Females may be herded by a male, may stay with a preferred male, or aggregate near a dominant male to avoid harassment from other males. In fission-fusion group dynamics, female movement is best described by the group’s fission probability, instead of inter-harem movement. In this study, we tested whether male herding ability, female mate choice or harassment avoidance influence fission probability. We recorded group dynamics in a herd of reindeer Rangifer tarandus equipped with GPS collars with activity sensors. We found no evidence that the harassment level in the group affected fission probability, or that females sought high rank (i.e. highly competitive and hence successful) males. However, the behavior of high ranked males decreased fission probability. Male herding activity was synchronous with the decrease of fission probability observed during the rut. We concluded that male herding behavior stabilized groups, thereby increasing average group size and consequently the opportunity for sexual selection. PMID:24759701

Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. Chen; CM Regan; D. Noe

2006-01-09

Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas releasemore » and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.« less

The Neutrons for Science Facility at SPIRAL-2.

PubMed

Ledoux, X; Aïche, M; Avrigeanu, M; Avrigeanu, V; Balanzat, E; Ban-d'Etat, B; Ban, G; Bauge, E; Bélier, G; Bém, P; Borcea, C; Caillaud, T; Chatillon, A; Czajkowski, S; Dessagne, P; Doré, D; Fischer, U; Frégeau, M O; Grinyer, J; Guillous, S; Gunsing, F; Gustavsson, C; Henning, G; Jacquot, B; Jansson, K; Jurado, B; Kerveno, M; Klix, A; Landoas, O; Lecolley, F R; Lecouey, J L; Majerle, M; Marie, N; Materna, T; Mrázek, J; Novák, J; Oberstedt, S; Oberstedt, A; Panebianco, S; Perrot, L; Plompen, A J M; Pomp, S; Prokofiev, A V; Ramillon, J M; Farget, F; Ridikas, D; Rossé, B; Serot, O; Simakov, S P; Šimecková, E; Stanoiu, M; Štefánik, M; Sublet, J C; Taïeb, J; Tarrío, D; Tassan-Got, L; Thfoin, I; Varignon, C

2017-11-21

The neutrons for science (NFS) facility is a component of SPIRAL-2, the new superconducting linear accelerator built at GANIL in Caen (France). The proton and deuteron beams delivered by the accelerator will allow producing intense neutron fields in the 100 keV-40 MeV energy range. Continuous and quasi-mono-kinetic energy spectra, respectively, will be available at NFS, produced by the interaction of a deuteron beam on a thick Be converter and by the 7Li(p,n) reaction on thin converter. The pulsed neutron beam, with a flux up to two orders of magnitude higher than those of other existing time-of-flight facilities, will open new opportunities of experiments in fundamental research as well as in nuclear data measurements. In addition to the neutron beam, irradiation stations for neutron-, proton- and deuteron-induced reactions will be available for cross-sections measurements and for the irradiation of electronic devices or biological cells. NFS, whose first experiment is foreseen in 2018, will be a very powerful tool for physics, fundamental research as well as applications like the transmutation of nuclear waste, design of future fission and fusion reactors, nuclear medicine or test and development of new detectors. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Bruyères-le-Châtel Neutron Evaluations of Actinides with the TALYS Code: The Fission Channel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Romain, P., E-mail: pascal.romain@cea.fr; Morillon, B.; Duarte, H.

For several years, various neutron evaluations of plutonium and uranium isotopes have been performed at Bruyères-le-Châtel (BRC), from 1 keV up to 30 MeV. Since only nuclear reaction models have been used to produce these evaluations, our approach was named the “Full Model” approach. Total, shape elastic and direct inelastic cross sections were obtained from the coupled channels model using a dispersive optical potential developed for actinides, with a large enough coupling scheme including the lowest octupolar band. All other cross sections were calculated using the Hauser-Feshbach theory (TALYS code) with a pre-equilibrium component above 8–10 MeV. In this paper,more » we focus our attention on the fission channel. More precisely, we will present the BRC contribution to fission modeling and the philosophy adopted in our “Full Model” approach. Performing evaluations with the “Full Model” approach implies the optimization of a large number of model parameters. With increasing neutron incident energy, many residual nuclei produced by nucleon emission also lead to fission. All available experimental data assigned to various fission mechanisms of the same nucleus were used to determine fission barrier parameters. For uranium isotopes, triple-humped fission barriers were required in order to reproduce accurately variations of the experimental fission cross sections. Our BRC fission modeling has shown that the effects of the class II or class III states located in the wells of the fission barrier sometimes provide an anti-resonant transmission rather than a resonant one. Consistent evaluations were produced for a large series of U and Pu isotopes. Resulting files were tested against integral data.« less

Feasibility study on the use of uranium in photoneutron target and BSA optimization for Linac based BNCT

NASA Astrophysics Data System (ADS)

Rahmani, Faezeh; Shahriari, Majid; Minoochehr, Abdolhamid; Nedaie, Hasan

2011-06-01

A hybrid photoneutron target including natural uranium has been studied for a 20 MeV linear electron accelerator (Linac) based Boron Neutron Capture Therapy (BNCT) facility. In this study the possibility of using uranium to increase the neutron intensity has been investigated by focusing on the time dependence behavior of the build-up and decay of the delayed gamma rays from fission fragments and activation products through photo-fission reactions in the BSA (Beam Shaping Assembly) configuration design. Delayed components of neutrons and photons were calculated. The obtained BSA parameters are in agreement with the IAEA recommendation and compared to the hybrid photoneutron target without U. The epithermal flux in the suggested design is 2.67E9 (n/cm 2s/mA).

Counting neutrons from the spontaneous fission of {sup 238}U using scintillation detectors and mixed field analysers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parker, Helen M. O'D.; Joyce, Malcolm J.; Jones, Ashley

2015-07-01

It is well documented that {sup 238}U decays by spontaneous fission, and that it is the main component of most nuclear fuels. As nuclear fuels are largely classed as Special Nuclear Material (SNM), they have to be fully accounted for by owners and processing facilities. One possible method for verifying declared amounts of SNM is to count the spontaneous neutrons produced from {sup 238}U. Using four EJ-309 liquid scintillation detectors and a mixed field analyser, spontaneous neutrons from 16.4 g of depleted uranium (0.3% enrichment) have been assayed. The assay method shows promising results and this proof of principle willmore » be researched further in order for it to be applied in an industrial setting. (authors)« less

Advanced Borobond™ Shields for Nuclear Materials Containment and Borobond™ Immobilization of Volatile Fission Products - Final CRADA Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagh, Arun S.

2016-05-19

Borobond is a company-proprietary material developed by the CRADA partner in collaboration with Argonne, and is based on Argonne's Ceramicrete technology. It is being used by DOE for nuclear materials safe storage, and Boron Products, LLC is the manufacturer and supplier of Borobond. The major objective of this project was to produce a more versatile composition of this material and find new applications. Major target applications were use for nuclear radiation shields, such as in dry storage casks; use in immobilization of most difficult waste streams, such as Hanford K-Basin waste; use for soluble and volatile fission products, such asmore » Cs, Tc, Sr, and I; and use for corrosion and fire protection applications in nuclear facilities.« less

Decommissioning ALARA programs Cintichem decommissioning experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adler, J.J.; LaGuardia, T.S.

1995-03-01

The Cintichem facility, originally the Union Carbide Nuclear Company (UCNC) Research Center, consisted primarily of a 5MW pool type reactor linked via a four-foot-wide by twelve-foot-deep water-filled canal to a bank of five adjacent hot cells. Shortly after going into operations in the early 1960s, the facility`s operations expanded to provide various reactor-based products and services to a multitude of research, production, medical, and education groups. From 1968 through 1972, the facility developed a process of separating isotopes from mixed fission products generated by irradiating enriched Uranium target capsules. By the late 1970s, 20 to 30 capsules were being processedmore » weekly, with about 200,000 curies being produced per week. Several isotopes such as Mo{sup 99}, I{sup 131}, and Xe{sup 133} were being extracted for medical use.« less

World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1979-06-01

Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment. (JWR)

Fission product release from fuel under LWR accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osborne, M.F.; Lorenz, R.A.; Norwood, K.S.

Three tests have provided additional data on fission product release under LWR accident conditions in a temperature range (1400 to 2000/sup 0/C). In the release rate data are compared with curves from a recent NRC-sponsored review of available fission product release data. Although the iodine release in test HI-3 was inexplicably low, the other data points for Kr, I, and Cs fall reasonably close to the corresponding curve, thereby tending to verify the NRC review. The limited data for antimony and silver release fall below the curves. Results of spark source mass spectrometric analyses were in agreement with the gammamore » spectrometric results. Nonradioactive fission products such as Rb and Br appeared to behave like their chemical analogs Cs and I. Results suggest that Te, Ag, Sn, and Sb are released from the fuel in elemental form. Analysis of the cesium and iodine profiles in the thermal gradient tube indicates that iodine was deposited as CsT along with some other less volatile cesium compound. The cesium profiles and chemical reactivity indicate the presence of more than one cesium species.« less

Compact multiwire proportional counters for the detection of fission fragments

NASA Astrophysics Data System (ADS)

Jhingan, Akhil; Sugathan, P.; Golda, K. S.; Singh, R. P.; Varughese, T.; Singh, Hardev; Behera, B. R.; Mandal, S. K.

2009-12-01

Two large area multistep position sensitive (two dimensional) multiwire proportional counters have been developed for experiments involving study of fission dynamics using general purpose scattering chamber facility at IUAC. Both detectors have an active area of 20×10 cm2 and provide position signals in horizontal (X) and vertical (Y) planes, timing signal for time of flight measurements and energy signal giving the differential energy loss in the active volume. The design features are optimized for the detection of low energy heavy ions at very low gas pressures. Special care was taken in setting up the readout electronics, constant fraction discriminators for position signals in particular, to get optimum position and timing resolutions along with high count rate handling capability of low energy heavy ions. A custom made charge sensitive preamplifier, having lower gain and shorter decay time, has been developed for extracting the differential energy loss signal. The position and time resolutions of the detectors were determined to be 1.1 mm full width at half maximum (FWHM) and 1.7 ns FWHM, respectively. The detector could handle heavy ion count rates exceeding 20 kHz without any breakdown. Time of flight signal in combination with differential energy loss signal gives a clean separation of fission fragments from projectile and target like particles. The timing and position signals of the detectors are used for fission coincidence measurements and subsequent extraction of their mass, angular, and total kinetic energy distributions. This article describes systematic study of these fission counters in terms of efficiency, time resolution, count rate handling capability, position resolution, and the readout electronics. The detector has been operated with both five electrode geometry and four electrode geometry, and a comparison has been made in their performances.

LOS ALAMOS NEUTRON SCIENCE CENTER CONTRIBUTIONS TO THE DEVELOPMENT OF FUTURE POWER REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

GAVRON, VICTOR I.; HILL, TONY S.; PITCHER, ERIC J.

The Los Alamos Neutron Science Center (LANSCE) is a large spallation neutron complex centered around an 800 MeV high-currently proton accelerator. Existing facilities include a highly-moderated neutron facility (Lujan Center) where neutrons between thermal and keV energies are produced, and the Weapons Neutron Research Center (WNR), where a bare spallation target produces neutrons between 0.1 and several hundred MeV.The LANSCE facility offers a unique capability to provide high precision nuclear data over a large energy region, including that for fast reactor systems. In an ongoing experimental program the fission and capture cross sections are being measured for a number ofmore » minor actinides relevant for Generation-IV reactors and transmutation technology. Fission experiments makes use of both the highly moderated spallation neutron spectrum at the Lujan Center, and the unmoderated high energy spectrum at WNR. By combininb measurements at these two facilities the differential fission cross section is measured relative to the {sup 235}U(n,f) standard from subthermal energies up to about 200 MeV. An elaborate data acquisition system is designed to deal with all the different types of background present when spanning 10 energy decades. The first isotope to be measured was {sup 237}Np, and the results were used to improve the current ENDF/B-VII evaluation. Partial results have also been obtained for {sup 240}Pu and {sup 242}Pu, and the final results are expected shortly. Capture cross sections are measured at LANSCE using the Detector for Advanced Neutron Capture Experiments (DANCE). This unique instrument is highly efficient in detecting radiative capture events, and can thus handle radioactive samples of half-lives as low as 100 years. A number of capture cross sections important to fast reaction applications have been measured with DANCE. The first measurement was on {sup 237}Np(n,{gamma}), and the results have been submitted for publication. Other capture measurements in progress include {sup 240}Pu and {sup 242}Pu. The United States recently announced the Global Nuclear Energy Partnership (GNEP), with the goal of closing the commercial nuclear fuel cycle while minimizing proliferation risk. GNEP achieves these goals using fast-spectrum nuclear reactors powered by new transmutation fuels that contain significant quantities of minor actinides. The proposed Materials Test Station (MTS) will provide the GNEP with a cost-effective means of obtaining domestic fast-spectrum irradiations of advanced transmutation fuel forms and structural materials, which is an important step in the fuels qualification process. The MTS will be located at the LANSCE, and will be driven by a 1.08-MW proton beam. Th epeak neutron flux in the irradiation region is 1.67 x 10{sup 15} n/cm{sup 2}/s, and the energy spectrum is similar to that of a fast reactor, with the addition of a high-energy tail. The facility is expected to operate at least 4,400 hours per year. Fuel burnup rates will exceed 4% per year, and the radiation damage rate in iron will be 18 dpa (displacements per atom) per year. The construction cost is estimated to be $73M (including 25% contingency), with annual operating costs in the range of $6M to $10M. Appropriately funded, the MTS could begin operation in 2010.« less

Fusion–fission experiments in Aphidius: evolutionary split without isolation in response to environmental bimodality

PubMed Central

Emelianov, I; Hernandes-Lopez, A; Torrence, M; Watts, N

2011-01-01

Studying host-based divergence naturally maintained by a balance between selection and gene flow can provide valuable insights into genetic underpinnings of host adaptation and ecological speciation in parasites. Selection-gene flow balance is often postulated in sympatric host races, but direct experimental evidence is scarce. In this study, we present such evidence obtained in host races of Aphidius ervi, an important hymenopteran agent of biological control of aphids in agriculture, using a novel fusion–fission method of gene flow perturbation. In our study, between-race genetic divergence was obliterated by means of advanced hybridisation, followed by a multi-generation exposure of the resulting genetically uniform hybrid swarm to a two-host environment. This fusion–fission procedure was implemented under two contrasting regimes of between-host gene flow in two replicated experiments involving different racial pairs. Host-based genetic fission in response to environmental bimodality occurred in both experiments in as little as six generations of divergent adaptation despite continuous gene flow. We demonstrate that fission recovery of host-based divergence evolved faster and hybridisation-induced linkage disequilibrium decayed slower under restricted (6.7%) compared with unrestricted gene flow, directly pointing at a balance between gene flow and divergent selection. We also show, in four separate tests, that random drift had no or little role in the observed genetic split. Rates and patterns of fission divergence differed between racial pairs. Comparative linkage analysis of these differences is currently under way to test for the role of genomic architecture of adaptation in ecology-driven divergent evolution. PMID:20924399

Evaluation of cooling concepts and specimen geometries for high heat flux tests on neutron irradiated divertor elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linke, J.; Bolt. H.; Breitbach, G.

1994-12-31

To assess the lifetime and the long term heat removal capabilities of plasma facing components in future thermonuclear fusion reactors such as ITER, neutron irradiation and subsequent high heat flux tests will be most essential. The effect of neutron damage will be simulated in material test reactors (such as the HFR-Petten) in a fission neutron environment. To investigate the heat loads during normal and off-normal operation scenarios a 60 kW electron beam test stand (Juelich Divertor Test Facility in Hot Cells, JUDITH) has been installed in a hot cell which can be operated by remote handling techniques. In this facilitymore » inertially cooled test coupons can be handled as well as small actively cooled divertor mock-ups. A special clamping mechanism for small test coupons (25 mm x 25 mm x 35 mm) with an integrated coolant channel within a copper or TZM heat sink has been developed and tested in an electron beam test bed. This method is an attractive alternative to costly large scale tests on complete divertor modules. The temperature and stress fields in individual CFC or beryllium tiles brazed to metallic heat sink (e.g. copper or TZM) can be investigated before and after neutron irradiation with moderate efforts.« less

The Alto Tandem and Isol Facility at IPN Orsay

NASA Astrophysics Data System (ADS)

Franchoo, Serge

Alto is an infrastructure for experimental nuclear physics in France that comprises both an on-line isotope-separation facility based on the photofission of uranium and a stable-ion beam facility based on a 14.5-MV tandem accelerator. The isotope-separation on-line section of Alto is dedicated to the production of neutron-rich radioactive ion beams (RIB) from the interaction of the γ-flux induced by a 50-MeV 10-µA electron beam in a uranium-carbide target. It is dimensioned for 1011 fissions per second. The RIB facility is exploited in alternating mode with the tandem-based section of Alto, capable of accelerating both light ions for nuclear astrophysics and heavy ions for γ-spectroscopy. The facility thereby offers the opportunity to deliver beams to a large range of physics programmes from nuclear to interdisciplinary physics. In this article, we present the Alto facility as well as some of the highlights and prospects of the experimental programme.

CONTROL CONSOLE FOR MTR FISSION PRODUCT MONITOR, USED TO DETECT ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

CONTROL CONSOLE FOR MTR FISSION PRODUCT MONITOR, USED TO DETECT BREAKS IN CLADDING OF FUEL ELEMENTS. COUNT-RATE METER IN TOP PANEL INDICATES AMOUNT OF RADIOACTIVITY. LOWER PANELS SUPPLY POWER AND AMPLIFICATION OF SIGNALS GENERATED BY SCINTILLATION COUNTER/PHOTOMULTIPLIER TUBE COMBINATION IN RESPONSE TO RADIOACTIVITY IN A SAMPLE OF THE COOLING WATER. INL NEGATIVE NO. 56-771. Jack L. Anderson, Photographer, 3/15/1956. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Plant model of KIPT neutron source facility simulator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cao, Yan; Wei, Thomas Y.; Grelle, Austin L.

2016-02-01

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine are collaborating on constructing a neutron source facility at KIPT, Kharkov, Ukraine. The facility has 100-kW electron beam driving a subcritical assembly (SCA). The electron beam interacts with a natural uranium target or a tungsten target to generate neutrons, and deposits its power in the target zone. The total fission power generated in SCA is about 300 kW. Two primary cooling loops are designed to remove 100-kW and 300-kW from the target zone and the SCA, respectively. A secondary cooling system ismore » coupled with the primary cooling system to dispose of the generated heat outside the facility buildings to the atmosphere. In addition, the electron accelerator has a low efficiency for generating the electron beam, which uses another secondary cooling loop to remove the generated heat from the accelerator primary cooling loop. One of the main functions the KIPT neutron source facility is to train young nuclear specialists; therefore, ANL has developed the KIPT Neutron Source Facility Simulator for this function. In this simulator, a Plant Control System and a Plant Protection System were developed to perform proper control and to provide automatic protection against unsafe and improper operation of the facility during the steady-state and the transient states using a facility plant model. This report focuses on describing the physics of the plant model and provides several test cases to demonstrate its capabilities. The plant facility model uses the PYTHON script language. It is consistent with the computer language of the plant control system. It is easy to integrate with the simulator without an additional interface, and it is able to simulate the transients of the cooling systems with system control variables changing on real-time.« less

Exhaust gas treatment in testing nuclear rocket engines

NASA Astrophysics Data System (ADS)

Zweig, Herbert R.; Fischler, Stanley; Wagner, William R.

1993-01-01

With the exception of the last test series of the Rover program, Nuclear Furnace 1, test-reactor and rocket engine hydrogen gas exhaust generated during the Rover/NERVA program was released directly to the atmosphere, without removal of the associated fission products and other radioactive debris. Current rules for nuclear facilities (DOE Order 5480.6) are far more protective of the general environment; even with the remoteness of the Nevada Test Site, introduction of potentially hazardous quantities of radioactive waste into the atmosphere must be scrupulously avoided. The Rocketdyne treatment concept features a diffuser to provide altitude simulation and pressure recovery, a series of heat exchangers to gradually cool the exhaust gas stream to 100 K, and an activated charcoal bed for adsorption of inert gases. A hydrogen-gas fed ejector provides auxiliary pumping for startup and shutdown of the engine. Supplemental filtration to remove particulates and condensed phases may be added at appropriate locations in the system. The clean hydrogen may be exhausted to the atmosphere and flared, or the gas may be condensed and stored for reuse in testing. The latter approach totally isolates the working gas from the environment.

Safe Affordable Fission Engine-(SAFE-) 100a Heat Exchanger Thermal and Structural Analysis

NASA Technical Reports Server (NTRS)

Steeve, B. E.

2005-01-01

A potential fission power system for in-space missions is a heat pipe-cooled reactor coupled to a Brayton cycle. In this system, a heat exchanger (HX) transfers the heat of the reactor core to the Brayton gas. The Safe Affordable Fission Engine- (SAFE-) 100a is a test program designed to thermally and hydraulically simulate a 95 Btu/s prototypic heat pipe-cooled reactor using electrical resistance heaters on the ground. This Technical Memorandum documents the thermal and structural assessment of the HX used in the SAFE-100a program.

Tests of the lunar hypothesis

NASA Technical Reports Server (NTRS)

Taylor, S. R.

1984-01-01

The concept that the Moon was fissioned from the Earth after core separation is the most readily testable hypothesis of lunar origin, since direct comparisons of lunar and terrestrial compositions can be made. Differences found in such comparisons introduce so many ad hoc adjustments to the fission hypothesis that it becomes untestable. Further constraints may be obtained from attempting to date the volatile-refractory element fractionation. The combination of chemical and isotopic problems suggests that the fission hypothesis is no longer viable, and separate terrestrial and lunar accretion from a population of fractionated precursor planetesimals provides a more reasonable explanation.

Design, Fabrication and Integration of a NaK-Cooled Circuit

NASA Technical Reports Server (NTRS)

Garber, Anne; Godfroy, Thomas

2006-01-01

The Early Flight Fission Test Facilities (EFF-TF) team has been tasked by the NASA Marshall Space Flight Center Nuclear Systems Office to design, fabricate, and test an actively pumped alkali metal flow circuit. The system, which was originally designed for use with a eutectic mixture of sodium potassium (NaK), was redesigned to for use with lithium. Due to a shi$ in focus, it is once again being prepared for use with NaK. Changes made to the actively pumped, high temperature circuit include the replacement of the expansion reservoir, addition of remotely operated valves, and modification of the support table. Basic circuit components include: reactor segment, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and a spill reservoir. A 37-pin partial-array core (pin and flow path dimensions are the same as those in a fill design) was selected for fabrication and test. This paper summarizes the integration and preparations for the fill of the pumped liquid metal NaK flow circuit.

HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sell, D. A.; Baily, C. E.; Malewitz, T. J.

2016-09-01

A system has been developed to perform measurements on irradiated, sodium bonded-metallic fuel elements to determine the amount of fission gas retained in the fuel material after release of the gas to the element plenum. During irradiation of metallic fuel elements, most of the fission gas developed is released from the fuel and captured in the gas plenums of the fuel elements. A significant amount of fission gas, however, remains captured in closed porosities which develop in the fuel during irradiation. Additionally, some gas is trapped in open porosity but sealed off from the plenum by frozen bond sodium aftermore » the element has cooled in the hot cell. The Retained fission Gas (RFG) system has been designed, tested and implemented to capture and measure the quantity of retained fission gas in characterized cut pieces of sodium bonded metallic fuel. Fuel pieces are loaded into the apparatus along with a prescribed amount of iron powder, which is used to create a relatively low melting, eutectic composition as the iron diffuses into the fuel. The apparatus is sealed, evacuated, and then heated to temperatures in excess of the eutectic melting point. Retained fission gas release is monitored by pressure transducers during the heating phase, thus monitoring for release of fission gas as first the bond sodium melts and then the fuel. A separate hot cell system is used to sample the gas in the apparatus and also characterize the volume of the apparatus thus permitting the calculation of the total fission gas release from the fuel element samples along with analysis of the gas composition.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barty, Christopher P.J.

Lasers and laser-based sources are now routinely used to control and manipulate nuclear processes, e.g. fusion, fission and resonant nuclear excitation. Two such “nuclear photonics” activities with the potential for profound societal impact will be reviewed in this presentation: the pursuit of laser-driven inertial confinement fusion at the National Ignition Facility and the development of laser-based, mono-energetic gamma-rays for isotope-specific detection, assay and imaging of materials.

Development of silver nanoparticle-doped adsorbents for the separation and recovery of radioactive iodine from alkaline solutions.

PubMed

Kim, Taewoon; Lee, Seung-Kon; Lee, Suseung; Lee, Jun Sig; Kim, Sang Wook

2017-11-01

Removing radioactive iodine from solutions containing fission products is essential for nuclear facility decontamination, radioactive waste treatment, and medical isotope production. For example, the production of high-purity fission 99 Mo by irradiation of 235 U with neutrons involves the removal of iodine from an alkaline solution of the irradiated target (which contains numerous fission products and a large quantity of aluminate ions) using silver-based materials or anion-exchange resins. To be practically applicable, the utilized iodine adsorbent should exhibit a decontamination factor of at least 200. Herein, the separation of radioactive iodine from alkaline solutions was achieved using alumina doped with silver nanoparticles (Ag NPs). Ag NPs have a larger surface area than Ag powder/wires and can thus adsorb iodine more effectively and economically, whereas alumina is a suitable inert support that does not adsorb 99 Mo and is stable under basic conditions. The developed adsorbents with less impurities achieved iodine removal and recovery efficiencies of 99.7 and 62%, respectively, thus being useful for the production of 131 I, a useful medical isotope. Copyright © 2017. Published by Elsevier Ltd.

Comparison of deterministic and stochastic approaches for isotopic concentration and decay heat uncertainty quantification on elementary fission pulse

NASA Astrophysics Data System (ADS)

Lahaye, S.; Huynh, T. D.; Tsilanizara, A.

2016-03-01

Uncertainty quantification of interest outputs in nuclear fuel cycle is an important issue for nuclear safety, from nuclear facilities to long term deposits. Most of those outputs are functions of the isotopic vector density which is estimated by fuel cycle codes, such as DARWIN/PEPIN2, MENDEL, ORIGEN or FISPACT. CEA code systems DARWIN/PEPIN2 and MENDEL propagate by two different methods the uncertainty from nuclear data inputs to isotopic concentrations and decay heat. This paper shows comparisons between those two codes on a Uranium-235 thermal fission pulse. Effects of nuclear data evaluation's choice (ENDF/B-VII.1, JEFF-3.1.1 and JENDL-2011) is inspected in this paper. All results show good agreement between both codes and methods, ensuring the reliability of both approaches for a given evaluation.

Data summary report for fission product release test VI-6

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osborne, M.F.; Lorenz, R.A.; Travis, J.R.

Test VI-6 was the sixth test in the VI series conducted in the vertical furnace. The fuel specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium. The fuel had experienced a burnup of {approximately}42 MWd/kg, with inert gas release during irradiation of {approximately}2%. The fuel specimen was heated in an induction furnace at 2300 K for 60 min, initially in hydrogen, then in a steam atmosphere. The released fission products were collected in three sequentially operated collection trains designed to facilitate sampling and analysis. The fission product inventories in the fuel were measured directlymore » by gamma-ray spectrometry, where possible, and were calculated by ORIGEN2. Integral releases were 75% for {sup 85}Kr, 67% for {sup 129}I, 64% for {sup 125}Sb, 80% for both {sup 134}Cs and {sup 137}Cs, 14% for {sup 154}Eu, 63% for Te, 32% for Ba, 13% for Mo, and 5.8% for Sr. Of the totals released from the fuel, 43% of the Cs, 32% of the Sb, and 98% of the Eu were deposited in the outlet end of the furnace. During the heatup in hydrogen, the Zircaloy cladding melted, ran down, and reacted with some of the UO{sub 2} and fission products, especially Te and Sb. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.57 g, almost equally divided between thermal gradient tubes and filters. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL Diffusion Model.« less

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gan, J.; Miller, B. D.; Keiser, D. D.

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (< 20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. This paper discusses the TEM results of the U-10Mo/Zr/Al6061 monolithic fuel plate (Plate ID: L1P09T, ~ 59% U-235 enrichment) irradiated in Advancedmore » Test Reactor at Idaho National Laboratory as part of RERTR-9B irradiation campaign with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 C, respectively. A total of 5 TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (> 1 µm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ~ 30 at% and ~ 7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.« less

Advantages and challenges in automated apatite fission track counting

NASA Astrophysics Data System (ADS)

Enkelmann, E.; Ehlers, T. A.

2012-04-01

Fission track thermochronometer data are often a core element of modern tectonic and denudation studies. Soon after the development of the fission track methods interest emerged for the developed an automated counting procedure to replace the time consuming labor of counting fission tracks under the microscope. Automated track counting became feasible in recent years with increasing improvements in computer software and hardware. One such example used in this study is the commercial automated fission track counting procedure from Autoscan Systems Pty that has been highlighted through several venues. We conducted experiments that are designed to reliably and consistently test the ability of this fully automated counting system to recognize fission tracks in apatite and a muscovite external detector. Fission tracks were analyzed in samples with a step-wise increase in sample complexity. The first set of experiments used a large (mm-size) slice of Durango apatite cut parallel to the prism plane. Second, samples with 80-200 μm large apatite grains of Fish Canyon Tuff were analyzed. This second sample set is characterized by complexities often found in apatites in different rock types. In addition to the automated counting procedure, the same samples were also analyzed using conventional counting procedures. We found for all samples that the fully automated fission track counting procedure using the Autoscan System yields a larger scatter in the fission track densities measured compared to conventional (manual) track counting. This scatter typically resulted from the false identification of tracks due surface and mineralogical defects, regardless of the image filtering procedure used. Large differences between track densities analyzed with the automated counting persisted between different grains analyzed in one sample as well as between different samples. As a result of these differences a manual correction of the fully automated fission track counts is necessary for each individual surface area and grain counted. This manual correction procedure significantly increases (up to four times) the time required to analyze a sample with the automated counting procedure compared to the conventional approach.

Future research program on prompt γ-ray emission in nuclear fission

NASA Astrophysics Data System (ADS)

Oberstedt, S.; Billnert, R.; Hambsch, F.-J.; Lebois, M.; Oberstedt, A.; Wilson, J. N.

2015-12-01

In recent years the measurement of prompt fission γ-ray spectra (PFGS) has gained renewed interest, after about forty years since the first comprehensive studies of the reactions 235U(n th , f), 239Pu(n th ,f) and 252Cf(sf). The renaissance was initiated by requests for new values especially for γ-ray multiplicity and average total energy release per fission in neutron-induced fission of 235U and 239Pu. Both isotopes are considered the most important ones with respect to the modeling of innovative cores required for the Generation-IV reactors, the majority working with fast neutrons. During the last 5 years we have conducted a systematic study of spectral data for thermal-neutron-induced fission on 235U and 241Pu as well as for the spontaneous fission of 252Cf with unprecedented accuracy. From the new data we conclude that those reactions do not considerably contribute to the observed heat excess and suspect other reactions playing a significant role. Possible contributions may originate from fast-neutron-induced reactions on 238U, which is largely present in the fuel, or from γ-induced fission from neutron capture in the construction material. A first experiment campaign on prompt γ-ray emission from fast-neutron-induced fission on 235,238U was successfully performed in order to test our assumptions. In the following we attempt to summarize, what has been done in the field to date, and to motivate future measurement campaigns exploiting dedicated neutron and photon beams as well as upcoming highly efficient detector assemblies.

Radiochemical Processing Laboratory (RPL) at PNNL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peurrung, Tony; Clark, Sue; Bryan, Sam

2017-03-23

Nuclear research is one of the core components of PNNL's mission. The centerpiece of PNNL's nuclear research is the Radiochemical Processing Laboratory (RPL), a Category 2 nuclear facility with state-of-the-art instrumentation, scientific expertise, and specialized capabilities that enable research with significant quantities of fissionable materials and other radionuclides—from tritium to plutonium. High impact radiological research has been conducted in the RPL since the 1950's, when nuclear weapons and energy production at Hanford were at the forefront of national defense. Since then, significant investments have been made in the RPL to maintain it as a premier nuclear science research facility supportingmore » multiple programs. Most recently, PNNL is developing a world-class analytical electron microscopy facility dedicated to the characterization of radiological materials.« less

A method for development of efficient 3D models for neutronic calculations of ASTRA critical facility using experimental information

DOE Office of Scientific and Technical Information (OSTI.GOV)

Balanin, A. L.; Boyarinov, V. F.; Glushkov, E. S.

The application of experimental information on measured axial distributions of fission reaction rates for development of 3D numerical models of the ASTRA critical facility taking into account azimuthal asymmetry of the assembly simulating a HTGR with annular core is substantiated. Owing to the presence of the bottom reflector and the absence of the top reflector, the application of 2D models based on experimentally determined buckling is impossible for calculation of critical assemblies of the ASTRA facility; therefore, an alternative approach based on the application of the extrapolated assembly height is proposed. This approach is exemplified by the numerical analysis ofmore » experiments on measurement of efficiency of control rods mockups and protection system (CPS).« less

Fabrication of (U, Zr) C-fueled/tungsten-clad specimens for irradiation in the Plum Brook Reactor Facility

NASA Technical Reports Server (NTRS)

1972-01-01

Fuel samples, 90UC - 10 ZrC, and chemically vapor deposited tungsten fuel cups were fabricated for the study of the long term dimensional stability and compatibility of the carbide-tungsten fuel-cladding systems under irradiation. These fuel samples and fuel cups were assembled into the fuel pins of two capsules, designated as V-2E and V-2F, for irradiation in NASA Plum Brook Reactor Facility at a fission power density of 172 watts/c.c. and a miximum cladding temperature of 1823 K. Fabrication methods and characteristics of the fuel samples and fuel cups prepared are described.

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

Data are summarized on fission-product activity during the period from October 1958 to October 1959. Measurements of airborne activity were made both above and below the tropopause layer. The effects were studied of heavy stratospheric injections from nuclear tests during 1958, after a year free from nuclear tests. lt was estimated that about 40 megatons of total fission energy was released during l958. About 20 megatons were released during the tests in the Arctic during September and October, 1958. The measured abundance of various nuclides indicates that the bulk of the heavy fall-out over Scandinavia was injected during October l958.more » Data are tabulated. (C.H.)« less

Data summary report for fission product release test VI-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osborne, M.F.; Collins, J.L.; Lorenz, R.A.

The first in a series of high-temperature fission product release test in a new vertical test apparatus was conducted in flowing steam. The test specimen was a 15.2-cm-long section of a fuel rod from the Oconee 1 PWR; it had been irradiated to a burnup of /approximately/42 MWd/kg. Using an induction furnace, it was heated under simulated LWR accident conditions -- 20 min at 2000 K and 20 min at 2300 K -- in a hot cell-mounted test apparatus. Posttest inspection showed severe oxidation but only minimal fragmentation of the fuel specimen; cladding melting was apparent only near the topmore » end. Based on fission product measured in the fuel and/or calculated by ORIGEN, analyses of test components showed total releases from the fuel of 47% for /sup 85/Kr, 33% for /sup 125/Sb, 37% for /sup 129/I, 84% for /sup 110m/Ag, and 63% for /sup 137/Cs. Large fractions (36% and 30%, respectively) of the released /sup 110m/Ag and /sup 125/Sb were retained in the furnace above the fuel. Pretest and posttest analysis of the fuel specimen indicated a /sup 134/Cs release of 65%, which is very good agreement with the /sup 137/Cs value. 21 refs., 24 figs., 16 tabs.« less

Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael R. Kruzic

2008-06-01

Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility (Figure 1) was used in the early to mid-1960s for testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles. The TCA facility, known as Corrective Action Unit 115, was decontaminated and decommissioned (D&D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the Federal Facility Agreement and Consentmore » Order. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously, provided technical decisions are made by an experienced decision maker within the site conceptual site model. Facility closure involved a seven-step decommissioning strategy. First, preliminary investigation activities were performed, including review of process knowledge documentation, targeted facility radiological and hazardous material surveys, concrete core drilling and analysis, shield wall radiological characterization, and discrete sampling, which proved to be very useful and cost-effective in subsequent decommissioning planning and execution and worker safety. Second, site setup and mobilization of equipment and personnel were completed. Third, early removal of hazardous materials, including asbestos, lead, cadmium, and oil, was performed ensuring worker safety during more invasive demolition activities. Process piping was to be verified void of contents. Electrical systems were de-energized and other systems were rendered free of residual energy. Fourth, areas of high radiological contamination were decontaminated using multiple methods. Contamination levels varied across the facility. Fixed beta/gamma contamination levels ranged up to 2 million disintegrations per minute (dpm)/100 centimeters squared (cm2) beta/gamma. Removable beta/gamma contamination levels seldom exceeded 1,000 dpm/100 cm2, but, in railroad trenches on the reactor pad containing soil on the concrete pad in front of the shield wall, the beta dose rates ranged up to 120 milli-roentgens per hour from radioactivity entrained in the soil. General area dose rates were less than 100 micro-roentgens per hour. Prior to demolition of the reactor shield wall, removable and fixed contaminated surfaces were decontaminated to the best extent possible, using traditional decontamination methods. Fifth, large sections of the remaining structures were demolished by mechanical and open-air controlled explosive demolition (CED). Mechanical demolition methods included the use of conventional demolition equipment for removal of three main buildings, an exhaust stack, and a mobile shed. The 5-foot (ft), 5-inch (in.) thick, neutron-activated reinforced concrete shield was demolished by CED, which had never been performed at the NTS.« less

Swelling and gas release in oxide fuels during fast temperature transients

NASA Astrophysics Data System (ADS)

Dollins, C. C.; Jursich, M.

1982-05-01

A previously reported intergranular swelling and gas release model for oxide fuels has been modified to predict fission gas behavior during fast temperature transients. Under steady state or slowly varying conditions it has been assumed in the previous model that the pressure caused by the fission gas within the gas bubbles is in equilibrium with the surface tension of the bubbles. During a fast transient, however, net vacancy migration to the bubbles may be insufficient to maintain this equilibrium. In order to ascertain the net vacancy flow, it is necessary to model the point defect behavior in the fuel. Knowing the net flow of vacancies to the bubble and the bubble size, the bubble diffusivity can be determined and the long range migration of the gas out of the fuel can be calculated. The model has also been modified to allow release of all the gas on the grain boundaries during a fast temperature transient. The gas release predicted by the revised model shows good agreement to fast transient gas release data from an EBR-II TREAT H-3 (Transient Reactor Test Facility) test. Agreement has also been obtained between predictions using the model and gas release data obtained by Argonne National Laboratory from out-of-reactor transient heating experiments on irradiated UO 2. It was found necessary to increase the gas bubble diffusivity used in the model by a factor of thirty during the transient to provide agreement between calculations and measurements. Other workers have also found that such an increase is necessary for agreement and attribute the increased diffusivity to yielding at the bubble surface due to the increased pressure.

Evaporation channel as a tool to study fission dynamics

NASA Astrophysics Data System (ADS)

Di Nitto, A.; Vardaci, E.; La Rana, G.; Nadtochy, P. N.; Prete, G.

2018-03-01

The dynamics of the fission process is expected to affect the evaporation residue cross section because of the fission hindrance due to the nuclear viscosity. Systems of intermediate fissility constitute a suitable environment for testing such hypothesis since they are characterized by evaporation residue cross sections comparable or larger than the fission ones. Observables related to emitted charged particles, due to their relatively high emission probability, can be used to put stringent constraints on models describing the excited nucleus decay and to recognize the effects of fission dynamics. In this work model simulations are compared with the experimental data collected via the 32S +100 Mo reaction at Elab = 200 MeV. Consequently we pointed out, exploring an extended set of evaporation channel observables, the limits of the statistical model and the large improvement obtained with a dynamical model. Moreover we stress the importance of using an apparatus covering a large fraction of 4π to extract observables. Finally, we discuss the opportunity to measure more sensitive observables by a new detection device in operation at LNL.

Derivation of effective fission gas diffusivities in UO2 from lower length scale simulations and implementation of fission gas diffusion models in BISON

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andersson, Anders David Ragnar; Pastore, Giovanni; Liu, Xiang-Yang

2014-11-07

This report summarizes the development of new fission gas diffusion models from lower length scale simulations and assessment of these models in terms of annealing experiments and fission gas release simulations using the BISON fuel performance code. Based on the mechanisms established from density functional theory (DFT) and empirical potential calculations, continuum models for diffusion of xenon (Xe) in UO 2 were derived for both intrinsic conditions and under irradiation. The importance of the large X eU3O cluster (a Xe atom in a uranium + oxygen vacancy trap site with two bound uranium vacancies) is emphasized, which is a consequencemore » of its high mobility and stability. These models were implemented in the MARMOT phase field code, which is used to calculate effective Xe diffusivities for various irradiation conditions. The effective diffusivities were used in BISON to calculate fission gas release for a number of test cases. The results are assessed against experimental data and future directions for research are outlined based on the conclusions.« less

Fission Spectrum

DOE R&D Accomplishments Database

Bloch, F.; Staub, H.

1943-08-18

Measurements of the spectrum of the fission neutrons of 25 are described, in which the energy of the neutrons is determined from the ionization produced by individual hydrogen recoils. The slow neutrons producing fission are obtained by slowing down the fast neutrons from the Be-D reaction of the Stanford cyclotron. In order to distinguish between fission neutrons and the remaining fast cyclotron neutrons both the cyclotron current and the pusle amplifier are modulated. A hollow neutron container, in which slow neutrons have a lifetime of about 2 milliseconds, avoids the use of large distances. This method results in much higher intensities than the usual modulation arrangement. The results show a continuous distribution of neutrons with a rather wide maximum at about 0.8 MV falling off to half of its maximum value at 2.0 MV. The total number of netrons is determined by comparison with the number of fission fragments. The result seems to indicate that only about 30% of the neutrons have energies below .8 MV. Various tests are described which were performed in order to rule out modification of the spectrum by inelastic scattering. Decl. May 4, 1951

Covariance generation and uncertainty propagation for thermal and fast neutron induced fission yields

NASA Astrophysics Data System (ADS)

Terranova, Nicholas; Serot, Olivier; Archier, Pascal; De Saint Jean, Cyrille; Sumini, Marco

2017-09-01

Fission product yields (FY) are fundamental nuclear data for several applications, including decay heat, shielding, dosimetry, burn-up calculations. To be safe and sustainable, modern and future nuclear systems require accurate knowledge on reactor parameters, with reduced margins of uncertainty. Present nuclear data libraries for FY do not provide consistent and complete uncertainty information which are limited, in many cases, to only variances. In the present work we propose a methodology to evaluate covariance matrices for thermal and fast neutron induced fission yields. The semi-empirical models adopted to evaluate the JEFF-3.1.1 FY library have been used in the Generalized Least Square Method available in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation) to generate covariance matrices for several fissioning systems such as the thermal fission of U235, Pu239 and Pu241 and the fast fission of U238, Pu239 and Pu240. The impact of such covariances on nuclear applications has been estimated using deterministic and Monte Carlo uncertainty propagation techniques. We studied the effects on decay heat and reactivity loss uncertainty estimation for simplified test case geometries, such as PWR and SFR pin-cells. The impact on existing nuclear reactors, such as the Jules Horowitz Reactor under construction at CEA-Cadarache, has also been considered.

KEY RESULTS FROM IRRADIATION AND POST-IRRADIATION EXAMINATION OF AGR-1 UCO TRISO FUEL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul A.; Hunn, John D.; Petti, David A.

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3×105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time average, volume average irradiation temperatures of the individual compacts ranged from 955 to 1136°C. This paper focuses on key resultsmore » from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior within the US program. The fuel exhibited a very low incidence of TRISO coating failure during irradiation and post-irradiation safety testing at temperatures up to 1800°C. Advanced PIE methods have allowed particles with SiC coating failure to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of lessons learned from AGR-1 to fuel fabrication and post-irradiation examination for subsequent fuel irradiation experiments as part of the US fuel program is also discussed.« less

Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul A.; Hunn, John D.; Petti, David A.

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3 × 105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time-average, volume-average irradiation temperatures of the individual compacts ranged from 955 to 1136 °C. This paper focuses on keymore » results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior. The fuel exhibited zero TRISO coating failures (failure of all three dense coating layers) during irradiation and post-irradiation safety testing at temperatures up to 1700 °C. Advanced PIE methods have allowed particles with SiC coating failure that were discovered to be present in a very-low population to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of lessons learned from AGR-1 to fuel fabrication and post-irradiation examination for subsequent fuel irradiation experiments as part of the US fuel program are also discussed.« less

Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel

DOE PAGES

Demkowicz, Paul A.; Hunn, John D.; Petti, David A.; ...

2017-09-10

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3 × 105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time-average, volume-average irradiation temperatures of the individual compacts ranged from 955 to 1136 °C. This paper focuses on keymore » results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior. The fuel exhibited zero TRISO coating failures (failure of all three dense coating layers) during irradiation and post-irradiation safety testing at temperatures up to 1700 °C. Advanced PIE methods have allowed particles with SiC coating failure that were discovered to be present in a very-low population to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of lessons learned from AGR-1 to fuel fabrication and post-irradiation examination for subsequent fuel irradiation experiments as part of the US fuel program are also discussed.« less

Report on the FY 1986 Activities of the Defense Science Study Group. Volume 1.

DTIC Science & Technology

1987-05-01

Reactors Improved Techniques for Wavefront Sensing and Correction in Adaptive Optics Hypervelocity Launchers Underground Facilities 0 Automated...oceanography and sound propagation in partially coherent media such as the turbulent ocean. There are large fixed arrays such as the Sound Surveillance System...Aircraft Continuous Patrol Aircraft Miscellaneous Studies Review of the Plutonium Special Isotope Separation Program of the DOE 4r Fusion Fission Hybrid

Experimental demonstration of radiation effects on the performance of a stirling-alternator convertor and candidate materials evaluation

NASA Astrophysics Data System (ADS)

Mireles, Omar R.

Free-piston Stirling power convertors are under consideration by NASA for service in the Advanced Stirling Radioisotope Generator (ASRG) and Fission Surface Power (FSP) systems to enable aggressive exploration missions by providing a reliable and constant power supply. The ASRG must withstand environmental radiation conditions, while the FSP system must tolerate a mixed neutron and gamma-ray environment resulting from self-irradiation. Stirling-alternators utilize rare earth magnets and a variety of organic materials whose radiation limits dominate service life estimates and shielding requirements. The project objective was to demonstrate the performance of the alternator, identify materials that exhibit excessive radiation sensitivity, identify radiation tolerant substitutes, establish empirical dose limits, and demonstrate the feasibility of cost effective nuclear and radiation tests by selection of the appropriate personnel and test facilities as a function of hardware maturity. The Stirling Alternator Radiation Test Article (SARTA) was constructed from linear alternator components of a Stirling convertor and underwent significant pre-exposure characterization. The SARTA was operated at the Sandia National Laboratories Gamma Irradiation Facility to a dose of over 40 Mrad. Operating performance was within nominal variation, although modestly decreasing trends occurred in later runs as well as the detection of an electrical fault after the final exposure. Post-irradiation disassembly and internal inspection revealed minimal degradation of the majority of the organic components. Radiation testing of organic material coupons was conducted since the majority of the literature was inconsistent. These inconsistencies can be attributed to testing at environmental conditions vastly different than those Stirling-alternator organics will experience during operation. Samples were irradiated at the Texas A&M TRIGA reactor to above expected FSP neutron fluence. A thorough materials evaluation followed and results indicate that the majority of material properties experienced minimal statistically significant change.

Spes: An intense source of Neutron-Rich Radioactive Beams at Legnaro

NASA Astrophysics Data System (ADS)

Andrighetto, A.; Manzolaro, M.; Corradetti, S.; Scarpa, D.; Monetti, A.; Rossignoli, M.; Ballan, M.; Borgna, F.; D'Agostini, F.; Gramegna, F.; Prete, G.; Meneghetti, G.; Ferrari, M.; Zenoni, A.

2018-02-01

The Isotope Separation On-Line (ISOL) method for the production of Radioactive Ion Beams (RIB) is attracting significant interest in the worldwide nuclear physics community. Within this context the SPES (Selective Production of Exotic Species) RIB facility is now under construction at INFN LNL (Istituto Nazionale di Fisica Nucleare Laboratori Nazionali di Legnaro). This technique is established as one of the main techniques for high intensity and high quality beams production. The SPES facility will produce n-rich isotopes by means of a 40 MeV proton beam, emitted by a cyclotron, impinging on a uranium carbide multi-foil fission target. The aim of this work is to describe the most important results obtained by the study of the on-line behavior of the SPES production target assembly. This target system will produce RIBs at a rate of about 1013 fissions per second, it will be able to dissipate a total power of up to 10 kW, and it is planned to work continuously for 2 week-runs of irradiation. ISOL beams of 24 different elements will be produced, therefore a target and ion source development is ongoing to ensure a great variety of produced isotopes and to improve the beam intensity and purity.

A method for reducing the largest relative errors in Monte Carlo iterated-fission-source calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunter, J. L.; Sutton, T. M.

2013-07-01

In Monte Carlo iterated-fission-source calculations relative uncertainties on local tallies tend to be larger in lower-power regions and smaller in higher-power regions. Reducing the largest uncertainties to an acceptable level simply by running a larger number of neutron histories is often prohibitively expensive. The uniform fission site method has been developed to yield a more spatially-uniform distribution of relative uncertainties. This is accomplished by biasing the density of fission neutron source sites while not biasing the solution. The method is integrated into the source iteration process, and does not require any auxiliary forward or adjoint calculations. For a given amountmore » of computational effort, the use of the method results in a reduction of the largest uncertainties relative to the standard algorithm. Two variants of the method have been implemented and tested. Both have been shown to be effective. (authors)« less

Fission properties of Po isotopes in different macroscopic-microscopic models

NASA Astrophysics Data System (ADS)

Bartel, J.; Pomorski, K.; Nerlo-Pomorska, B.; Schmitt, Ch

2015-11-01

Fission-barrier heights of nuclei in the Po isotopic chain are investigated in several macroscopic-microscopic models. Using the Yukawa-folded single-particle potential, the Lublin-Strasbourg drop (LSD) model, the Strutinsky shell-correction method to yield the shell corrections and the BCS theory for the pairing contributions, fission-barrier heights are calculated and found in quite good agreement with the experimental data. This turns out, however, to be only the case when the underlying macroscopic, liquid-drop (LD) type, theory is well chosen. Together with the LSD approach, different LD parametrizations proposed by Moretto et al are tested. Four deformation parameters describing respectively elongation, neck-formation, reflectional-asymmetric, and non-axiality of the nuclear shape thus defining the so called modified Funny Hills shape parametrization are used in the calculation. The present study clearly demonstrates that nuclear fission-barrier heights constitute a challenging and selective tool to discern between such different macroscopic approaches.

Germanium-gated γ-γ fast timing of excited states in fission fragments using the EXILL&FATIMA spectrometer

NASA Astrophysics Data System (ADS)

Régis, J.-M.; Simpson, G. S.; Blanc, A.; de France, G.; Jentschel, M.; Köster, U.; Mutti, P.; Paziy, V.; Saed-Samii, N.; Soldner, T.; Ur, C. A.; Urban, W.; Bruce, A. M.; Drouet, F.; Fraile, L. M.; Ilieva, S.; Jolie, J.; Korten, W.; Kröll, T.; Lalkovski, S.; Mach, H.; Mărginean, N.; Pascovici, G.; Podolyak, Zs.; Regan, P. H.; Roberts, O. J.; Smith, J. F.; Townsley, C.; Vancraeyenest, A.; Warr, N.

2014-11-01

A high-granularity mixed spectrometer consisting of high-resolution Ge and very fast LaBr3(Ce)-scintillator detectors has been installed around a fission target at the cold-neutron guide PF1B of the high-flux reactor of the Institut Laue-Langevin. Lifetimes of excited states in the range of 10 ps to 10 ns can be measured in around 100 exotic neutron-rich fission fragments using Ge-gated LaBr3(Ce)-LaBr3(Ce) or Ge-Ge-LaBr3(Ce)-LaBr3(Ce) coincidences. We report on various characteristics of the EXILL&FATIMA spectrometer for the energy range of 40 keV up to 6.8 MeV and present results of ps-lifetime test measurements in a fission fragment. The results are discussed with respect to possible systematic errors induced by background contributions.

The DART dispersion analysis research tool: A mechanistic model for predicting fission-product-induced swelling of aluminum dispersion fuels. User`s guide for mainframe, workstation, and personal computer applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rest, J.

1995-08-01

This report describes the primary physical models that form the basis of the DART mechanistic computer model for calculating fission-product-induced swelling of aluminum dispersion fuels; the calculated results are compared with test data. In addition, DART calculates irradiation-induced changes in the thermal conductivity of the dispersion fuel, as well as fuel restructuring due to aluminum fuel reaction, amorphization, and recrystallization. Input instructions for execution on mainframe, workstation, and personal computers are provided, as is a description of DART output. The theory of fission gas behavior and its effect on fuel swelling is discussed. The behavior of these fission products inmore » both crystalline and amorphous fuel and in the presence of irradiation-induced recrystallization and crystalline-to-amorphous-phase change phenomena is presented, as are models for these irradiation-induced processes.« less

Characterization studies of prototype ISOL targets for the RIA

NASA Astrophysics Data System (ADS)

Greene, John P.; Burtseva, Tatiana; Neubauer, Janelle; Nolen, Jerry A.; Villari, Antonio C. C.; Gomes, Itacil C.

2005-12-01

Targets employing refractory compounds are being developed for the rare isotope accelerator (RIA) facility to produce ion species far from stability. With the 100 kW beams proposed for the production targets, dissipation of heat becomes a challenging issue. In our two-step target design, neutrons are generated in a refractory primary target, inducing fission in the surrounding uranium carbide. The interplay of density, grain size, thermal conductivity and diffusion properties of the UC2 needs to be well understood before fabrication. Thin samples of uranium carbide were prepared for thermal conductivity measurements using an electron beam to heat the sample and an optical pyrometer to observe the thermal radiation. Release efficiencies and independent thermal analysis on these samples are being undertaken at Oak Ridge National Laboratory (ORNL). An alternate target concept for RIA, the tilted slab approach promises to be simple with fast ion release and capable of withstanding high beam intensities while providing considerable yields via spallation. A proposed small business innovative research (SBIR) project will design a prototype tilted target, exploring the materials needed for fabrication and testing at an irradiation facility to address issues of heat transfer and stresses within the target.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Deyglun, C.; Simony, B.; Perot, B.

The quantification of radioactive material is essential in the fields of safeguards, criticality control of nuclear processes, dismantling of nuclear facilities and components, or radioactive waste characterization. The Nuclear Measurement Laboratory (LMN) of CEA is involved in the development of time-correlated neutron detection techniques using plastic scintillators. Usually, 3He proportional counters are used for passive neutron coincidence counting owing to their high thermal neutron capture efficiency and gamma insensitivity. However, the global {sup 3}He shortage in the past few years has made these detectors extremely expensive. In addition, contrary to {sup 3}He counters for which a few tens of microsecondsmore » are needed to thermalize fast neutrons, in view to maximize the {sup 3}He(n,p){sup 3}H capture cross section, plastic scintillators are based on elastic scattering and therefore the light signal is formed within a few nanoseconds, correlated pulses being detected within a few dozen- or hundred nanoseconds. This time span reflects fission particles time of flight, which allows reducing accordingly the duration of the coincidence gate and thus the rate of random coincidences, which may totally blind fission coincidences when using {sup 3}He counters in case of a high (α,n) reaction rate. However, plastic scintillators are very sensitive to gamma rays, requiring the use of a thick metallic shield to reduce the corresponding background. Cross talk between detectors is also a major issue, which consists on the detection of one particle by several detectors due to elastic or inelastic scattering, leading to true but undesired coincidences. Data analysis algorithms are tested to minimize cross-talk in simultaneously activated detectors. The distinction between useful fission coincidences and the correlated background due to cross-talk, (α,n) and induced (n,2n) or (n,n'γ) reactions, is achieved by measuring 3-fold coincidences. The performances of a passive neutron coincidence counting system for radioactive waste drums using plastic scintillators have been studied using the Monte Carlo radiation transport code MCNPX-PoliMi v2.0 coupled to data processing algorithms developed with ROOT data analysis software. In addition to the correlated background, accidental coincidences are taken into account in the simulation by randomly merging pulses from different calculations with fission and (α,n) sources. (authors)« less

Deceleration of Fusion–Fission Cycles Improves Mitochondrial Quality Control during Aging

PubMed Central

Meyer-Hermann, Michael; Osiewacz, Heinz D.

2012-01-01

Mitochondrial dynamics and mitophagy play a key role in ensuring mitochondrial quality control. Impairment thereof was proposed to be causative to neurodegenerative diseases, diabetes, and cancer. Accumulation of mitochondrial dysfunction was further linked to aging. Here we applied a probabilistic modeling approach integrating our current knowledge on mitochondrial biology allowing us to simulate mitochondrial function and quality control during aging in silico. We demonstrate that cycles of fusion and fission and mitophagy indeed are essential for ensuring a high average quality of mitochondria, even under conditions in which random molecular damage is present. Prompted by earlier observations that mitochondrial fission itself can cause a partial drop in mitochondrial membrane potential, we tested the consequences of mitochondrial dynamics being harmful on its own. Next to directly impairing mitochondrial function, pre-existing molecular damage may be propagated and enhanced across the mitochondrial population by content mixing. In this situation, such an infection-like phenomenon impairs mitochondrial quality control progressively. However, when imposing an age-dependent deceleration of cycles of fusion and fission, we observe a delay in the loss of average quality of mitochondria. This provides a rational why fusion and fission rates are reduced during aging and why loss of a mitochondrial fission factor can extend life span in fungi. We propose the ‘mitochondrial infectious damage adaptation’ (MIDA) model according to which a deceleration of fusion–fission cycles reflects a systemic adaptation increasing life span. PMID:22761564

FASTGRASS: A mechanistic model for the prediction of Xe, I, Cs, Te, Ba, and Sr release from nuclear fuel under normal and severe-accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rest, J.; Zawadzki, S.A.

The primary physical/chemical models that form the basis of the FASTGRASS mechanistic computer model for calculating fission-product release from nuclear fuel are described. Calculated results are compared with test data and the major mechanisms affecting the transport of fission products during steady-state and accident conditions are identified.

Feynman variance for neutrons emitted from photo-fission initiated fission chains - a systematic simulation for selected speacal nuclear materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soltz, R. A.; Danagoulian, A.; Sheets, S.

Theoretical calculations indicate that the value of the Feynman variance, Y2F for the emitted distribution of neutrons from ssionable exhibits a strong monotonic de- pendence on a the multiplication, M, of a quantity of special nuclear material. In 2012 we performed a series of measurements at the Passport Inc. facility using a 9- MeV bremsstrahlung CW beam of photons incident on small quantities of uranium with liquid scintillator detectors. For the set of objects studies we observed deviations in the expected monotonic dependence, and these deviations were later con rmed by MCNP simulations. In this report, we modify the theorymore » to account for the contri- bution from the initial photo- ssion and benchmark the new theory with a series of MCNP simulations on DU, LEU, and HEU objects spanning a wide range of masses and multiplication values.« less

A Novel Approach to β-delayed Neutron Spectroscopy Using the Beta-decay Paul Trap

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scielzo, N.D., E-mail: scielzo1@llnl.gov; Yee, R.M.; Department of Nuclear Engineering, University of California, Berkeley, CA 94720

A new approach to β-delayed neutron spectroscopy has been demonstrated that circumvents the many limitations associated with neutron detection by instead inferring the decay branching ratios and energy spectra of the emitted neutrons by studying the nuclear recoil. Using the Beta-decay Paul Trap, fission-product ions were trapped and confined to within a 1-mm{sup 3} volume under vacuum using only electric fields. Results from recent measurements of {sup 137}I{sup +} and plans for development of a dedicated ion trap for future experiments using the intense fission fragment beams from the Californium Rare Isotope Breeder Upgrade (CARIBU) facility at Argonne National Laboratorymore » are summarized. The improved nuclear data that can be collected is needed in many fields of basic and applied science such as nuclear energy, nuclear astrophysics, and stockpile stewardship.« less

Comment on "radioactive fallout in the United States due to the Fukushima nuclear plant accident" by P. Thakur, S. Ballard and R. Nelson, J. Environ. Monit., 2012, 14, 1317-1324.

PubMed

Rose, Paula S

2014-07-01

The May 2012 paper "Radioactive fallout in the United States due to the Fukushima nuclear plant accident" (P. Thakur, S. Ballard and R. Nelson, J. Environ. Monit., 2012, 14, 1317-1324), does not address medical patient excreta as a source of (131)I (t1/2 = 8.04 d) to the environment. While (131)I is generated during fission reactions and may be released to the environment from nuclear power plants, nuclear weapons tests, nuclear fuel reprocessing and weapons production facilities, it is also produced for medical use. Iodine-131 administered to patients, excreted and discharged to sewer systems is readily measureable in sewage and the environment; the patient-to-sewage pathway is the only source of (131)I in many locations.

Targeted Modification of Neutron Energy Spectra for National Security Applications

NASA Astrophysics Data System (ADS)

Bevins, James Edward

At its core, research represents an attempt to break from the "this is the way we have always done it" paradigm. This idea is evidenced from the start in this research effort by the problem formulation to develop a new way to generate synthetic debris that mimics the samples that would be collected for forensics purposes following a nuclear weapon attack on the U.S. or its allies. The philosophy is also demonstrated by the design methodology used to solve the synthetic debris problem, using methods not commonly applied to nuclear engineering problems. Through this research, the bounds of what is deemed possible in neutron spectral shaping are moved ever so slightly. A capability for the production of synthetic debris and fission products was developed for the National Ignition Facility (NIF). Synthetic debris has historically been made in a limited fashion using sample doping techniques since the cessation of nuclear weapons testing, but a more robust alternative approach using neutron spectral shaping was proposed and developed by the University of California-Berkeley and Lawrence Livermore National Laboratory (LLNL). Using NIF as a starting source spectrum, the energy tuning assembly (ETA) developed in this work can irradiate samples with a combined thermonuclear and prompt fission neutron spectrum (TN+PFNS). When used with fissile foils, this irradiation will produce a synthetic fission product distribution that is realistic across all mass chains. To design the ETA, traditional parametric point design approaches were discarded in favor of formal optimization techniques. Finding a lack of suitable algorithms in the literature, a metaheuristic-based optimization algorithm, Gnowee, was developed for rapid convergence to nearly globally optimum solutions for complex, constrained engineering problems with mixed-integer and combinatorial design vectors and high-cost, noisy, discontinuous, black box objective function evaluations. Comparisons between Gnowee and several well-established metaheuristic algorithms are made for a set of continuous, mixed-integer, and combinatorial benchmarks. These results demonstrated Gnoweee to have superior flexibility and convergence characteristics over a wide range of design spaces. The Gnowee algorithm was implemented in Coeus, a new piece of software, to perform optimization of design problems requiring radiation transport for the evaluation of their objective functions. Currently, Coeus solves ETA optimization problems using hybrid radiation transport (ADVANTG and MCNP) to assess design permutations developed by Gnowee. Future enhancements of Coeus will look to expand the geometries and objective functions considered to those beyond ETA design. Coeus was used to generate an ETA design for the TN+PFNS application on NIF. The design achieved a reasonable match with the objective TN+PFNS and associated fission product distributions within the size and weight constraints imposed by the NIF facility. The ETA design was built by American Elements, and initial validation tests were conducted at the Lawrence Berkeley National Laboratory's 88-Inch Cyclotron. These experiments used foil activation and pulse height spectroscopy to measure the ETA-modified spectrum. Additionally, pulse height spectroscopy measurements were taken as the ETA was built-up component-by-component to measure the impact of nuclear data on the ability to model the ETA performance. Some initial analysis of these results is included here. Finally, an integral validation experiment on NIF was proposed using the Coeus generated ETA design. A scoping study conducted by LLNL determined the proposed experiment and ETA design are within NIF facility limitations and current radio-chemistry capabilities. The study found that the proposed ETA experiment was "low risk," has "no show stoppers," and has a "reasonable cost." All that is needed is a sponsor to close the last funding gap and bring the experiment to fruition. This research broke with the current sample doping approach and applied neutron spectral shaping to design an ETA that can create realistic synthetic fission and activation products and improve technical nuclear forensics outcomes. However, the ETA presented in this research represents more than a stand alone point design with a limited scope and application. It is proof of a concept and the product of a unique capability that has a wide range of potential applications. This research demonstrates that the concept of neutron spectral shaping can be used to engineer complex neutron spectra within the confines of physics. There are many possible applications that could benefit from the ability to generate custom energy neutron spectra that fall outside of current sources and methods. The ETA is the product of a general-purpose optimization algorithm, Gnowee, and design framework, Coeus, which enables the use of Gnowee for complex nuclear design problems. Through Gnowee and Coeus, new ETA neutronics designs can be generated in days, not months or years, with a drastic reduction in the research effort required to do so. (Abstract shortened by ProQuest.).

Analytical dose evaluation of neutron and secondary gamma-ray skyshine from nuclear facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hayashi, K.; Nakamura, T.

1985-11-01

The skyshine dose distributions of neutron and secondary gamma rays were calculated systematically using the Monte Carlo method for distances up to 2 km from the source. The energy of source neutrons ranged from thermal to 400 MeV; their emission angle from 0 to 90 deg from the ver tical was treated with a distribution of the direction cosine containing five equal intervals. Calculated dose distributions D(r) were fitted to the formula; D(r) = Q exp (-r/lambda)/r. The value of Q and lambda are slowly varied functions of energy. This formula was applied to the benchmark problems of neutron skyshinemore » from fission, fusion, and accelerator facilities, and good agreement was achieved. This formula will be quite useful for shielding designs of various nuclear facilities.« less

Testing in Support of Space Fission System Development and Qualification

NASA Technical Reports Server (NTRS)

Houts, Mike; Bragg-Sitton, Shannon; Garber, Anne; Godfrey, Tom; Martin, Jim; Pearson, Boise; Webster, Kenny

2007-01-01

Extensive data would be required for the qualification of a fission surface power (FSP) system. The strategy for qualifying a FSP system could have a significant programmatic impact. This paper explores potential options that could be used for qualifying FSP systems, including cost-effective means for obtaining required data. three methods for obtaining qualification data are analysis, non-nuclear testing, and nuclear testing. It has been over 40 years since the US qualified a space reactor for launch. During that time, advances have been made related to all three methods. Perhaps the greatest advancement has occurred in the area of computational tools for design and analysis. Tools that have been developed, coupled with modem computers, would have a significant impact on a FSP qualification. This would be especially true for systems with materials and fuels operating well within temperature, irradiation damage, and burnup limits. The ability to perform highly realistic non-nuclear testing has also advanced throughout the past four decades. Instrumented thermal simulators were developed during the 1970s and 1980s to assist in the development, operation, and assessment of terrestrial fission systems. Instrumented thermal simulators optimized for assisting in the development, operation, and assessment of modem FSP systems have been under development (and utilized) since 1998. These thermal simulators enable heat from fission to be closely mimicked (axial power profile, radial power profile, temperature, heat flux, etc.} and extensive data to be taken from the core region. Both steady-state and transient operation can be tested. For transient testing, reactivity feedback is calculated (or measured in cold/warm criticals) based on reactor temperature and/or dimensional changes. Pin power during a transient is then calculated based on the reactivity feedback that would occur given measured values of temperature and/or dimensional change. In this way nonnuclear testing can be used to provide very realistic information related to nuclear operation. Non-nuclear testing can be used at all levels, including component, subsystem, and integrated system testing. Realistic non-nuclear testing is most useful for systems operating within known temperature, irradiation damage, and burnup capabilities.

Fission-neutrons source with fast neutron-emission timing

NASA Astrophysics Data System (ADS)

Rusev, G.; Baramsai, B.; Bond, E. M.; Jandel, M.

2016-05-01

A neutron source with fast timing has been built to help with detector-response measurements. The source is based on the neutron emission from the spontaneous fission of 252Cf. The time is provided by registering the fission fragments in a layer of a thin scintillation film with a signal rise time of 1 ns. The scintillation light output is measured by two silicon photomultipliers with rise time of 0.5 ns. Overall time resolution of the source is 0.3 ns. Design of the source and test measurements using it are described. An example application of the source for determining the neutron/gamma pulse-shape discrimination by a stilbene crystal is given.

Exploiting Fission Chain Reaction Dynamics to Image Fissile Materials

NASA Astrophysics Data System (ADS)

Chapman, Peter Henry

Radiation imaging is one potential method to verify nuclear weapons dismantlement. The neutron coded aperture imager (NCAI), jointly developed by Oak Ridge National Laboratory (ORNL) and Sandia National Laboratories (SNL), is capable of imaging sources of fast (e.g., fission spectrum) neutrons using an array of organic scintillators. This work presents a method developed to discriminate between non-multiplying (i.e., non-fissile) neutron sources and multiplying (i.e., fissile) neutron sources using the NCAI. This method exploits the dynamics of fission chain-reactions; it applies time-correlated pulse-height (TCPH) analysis to identify neutrons in fission chain reactions. TCPH analyzes the neutron energy deposited in the organic scintillator vs. the apparent neutron time-of-flight. Energy deposition is estimated from light output, and time-of-flight is estimated from the time between the neutron interaction and the immediately preceding gamma interaction. Neutrons that deposit more energy than can be accounted for by their apparent time-of-flight are identified as fission chain-reaction neutrons, and the image is reconstructed using only these neutron detection events. This analysis was applied to measurements of weapons-grade plutonium (WGPu) metal and 252Cf performed at the Nevada National Security Site (NNSS) Device Assembly Facility (DAF) in July 2015. The results demonstrate it is possible to eliminate the non-fissile 252Cf source from the image while preserving the fissileWGPu source. TCPH analysis was also applied to additional scenes in which theWGPu and 252Cf sources were measured individually. The results of these separate measurements further demonstrate the ability to remove the non-fissile 252Cf source and retain the fissileWGPu source. Simulations performed using MCNPX-PoliMi indicate that in a one hour measurement, solid spheres ofWGPu are retained at a 1sigma level for neutron multiplications M -˜ 3.0 and above, while hollowWGPu spheres are retained for M -˜ 2.7 and above.

Sub-barrier quasifission in heavy element formation reactions with deformed actinide target nuclei

NASA Astrophysics Data System (ADS)

Hinde, D. J.; Jeung, D. Y.; Prasad, E.; Wakhle, A.; Dasgupta, M.; Evers, M.; Luong, D. H.; du Rietz, R.; Simenel, C.; Simpson, E. C.; Williams, E.

2018-02-01

Background: The formation of superheavy elements (SHEs) by fusion of two massive nuclei is severely inhibited by the competing quasifission process. Low excitation energies favor SHE survival against fusion-fission competition. In "cold" fusion with spherical target nuclei near 208Pb, SHE yields are largest at beam energies significantly below the average capture barrier. In "hot" fusion with statically deformed actinide nuclei, this is not the case. Here the elongated deformation-aligned configurations in sub-barrier capture reactions inhibits fusion (formation of a compact compound nucleus), instead favoring rapid reseparation through quasifission. Purpose: To determine the probabilities of fast and slow quasifission in reactions with prolate statically deformed actinide nuclei, through measurement and quantitative analysis of the dependence of quasifission characteristics at beam energies spanning the average capture barrier energy. Methods: The Australian National University Heavy Ion Accelerator Facility and CUBE fission spectrometer have been used to measure fission and quasifission mass and angle distributions for reactions with projectiles from C to S, bombarding Th and U target nuclei. Results: Mass-asymmetric quasifission occurring on a fast time scale, associated with collisions with the tips of the prolate actinide nuclei, shows a rapid increase in probability with increasing projectile charge, the transition being centered around projectile atomic number ZP=14 . For mass-symmetric fission events, deviations of angular anisotropies from expectations for fusion fission, indicating a component of slower quasifission, suggest a similar transition, but centered around ZP˜8 . Conclusions: Collisions with the tips of statically deformed prolate actinide nuclei show evidence for two distinct quasifission processes of different time scales. Their probabilities both increase rapidly with the projectile charge. The probability of fusion can be severely suppressed by these two quasifission processes, since the sub-barrier heavy element yield is likely to be determined by the product of the probabilities of surviving each quasifission process.

Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

2002-08-01

Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

DOE Office of Scientific and Technical Information (OSTI.GOV)

S. Bragg-Sitton; J. Bess; J. Werner

2011-09-01

Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al.,more » 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).« less

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

NASA Astrophysics Data System (ADS)

Seibert, Rachel L.; Terrani, Kurt A.; Velázquez, Daniel; Hunn, John D.; Baldwin, Charles A.; Montgomery, Fred C.; Terry, Jeff

2018-03-01

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Program and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form PdxSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. They may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.

Paramecium: a promising non-animal bioassay to study the effect of 808 nm infrared diode laser photobiomodulation.

PubMed

Amaroli, Andrea; Parker, Steven; Dorigo, Gianluca; Benedicenti, Alberico; Benedicenti, Stefano

2015-01-01

Photobiostimulation and photobiomodulation (PBM) are terms applied to the manipulation of cellular behavior using low intensity light sources, which works on the principle of inducing a biological response through energy transfer. The aim of this investigation was to identify a laboratory assay to test the effect of an infrared diode laser light (808 nm) on cell fission rate. Sixty cells of Paramecium primaurelia were divided in two groups of 30. The first group (test group) was irradiated, at a temperature of 24°C, for 50 sec by a 808 nm diode laser with a flat top handpiece [1 cm of spot diameter, 1 W in continuous wave (CW), 50 sec irradiation time, 64 J/cm(2) of fluence]. The second group (control group) received no laser irradiation. All cells were transferred onto a depression slide, fed, and incubated in a moist chamber at a temperature of 24°C. The cells were exposed and monitored for 10 consecutive fission rates. Changes in temperature and pH were also evaluated. The exposed cells had a fission rate rhythm faster than the control cells, showing a binary fission significantly (p<0.05) shorter than unexposed cells. No significant effects of laser irradiation on pH and temperature of Paramecium's lettuce infusion medium were observed. The 808 nm infrared diode laser light, at the irradiation parameters used in our work, results in a precocious fission rate in P. primaurelia cells, probably through an increase in metabolic activity, secondary to an energy transfer.

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

DOE PAGES

Seibert, Rachel L.; Terrani, Kurt A.; Velázquez, Daniel; ...

2018-03-01

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Programmore » and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form Pd xSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. In conclusion, they may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.« less

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

DOE Office of Scientific and Technical Information (OSTI.GOV)

Seibert, Rachel L.; Terrani, Kurt A.; Velázquez, Daniel

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Programmore » and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form Pd xSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. In conclusion, they may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.« less

Temperature-Dependent Growth and Fission Rate Plasticity Drive Seasonal and Geographic Changes in Body Size in a Clonal Sea Anemone.

PubMed

Ryan, Will H

2018-02-01

The temperature-size rule is a commonly observed pattern where adult body size is negatively correlated with developmental temperature. In part, this may occur as a consequence of allometric scaling, where changes in the ratio of surface area to mass limit oxygen diffusion as body size increases. As oxygen demand increases with temperature, a smaller body should be favored as temperature increases. For clonal animals, small changes in growth and/or fission rate can rapidly alter the average body size of clonal descendants. Here I test the hypothesis that the clonal sea anemone Diadumene lineata is able to track an optimal body size through seasonal temperature changes using fission rate plasticity. Individuals from three regions (Florida, Georgia, and Massachusetts) across the species' latitudinal range were grown in a year-long reciprocal common garden experiment mimicking seasonal temperature changes at three sites. Average body size was found to be smaller and fission rates higher in warmer conditions, consistent with the temperature-size rule pattern. However, seasonal size and fission patterns reflect a complex interaction between region-specific thermal reaction norms and the local temperature regime. These details provide insight into both the range of conditions required for oxygen limitation to contribute to a negative correlation between body size and temperature and the role that fission rate plasticity can play in tracking a rapidly changing optimal phenotype.

Scotty, I Need More Power - The Fission System Gateway to Abundant Power for Exploration

NASA Technical Reports Server (NTRS)

Palac, Donald T.

2011-01-01

In planning and in crisis, electrical power has been a key consideration when humans venture into space. Since the 1950's, nuclear fission (splitting of atoms) power has been a logical alternative in both fact and fiction, due to its ability to provide abundant power with high energy density, reliability, and immunity to severe environments. Bringing space fission power to a state of readiness for exploration has depended on clearing the hurdle of technology readiness demonstration. Due to the happy coincidence of heritage from prior space fission development efforts such as the Prometheus program, foresight from NASA's Exploration Mission Systems Directorate in the mid-2000's, and relative budget stability through the late 2000's, National Aeronautics and Space Administration (NASA) and Department of Energy (DOE), with their industry partners, are poised to push through to this objective. Hardware for a 12 kWe non-nuclear Fission Power System Technology Demonstration Unit is being fabricated now on a schedule that will enable a low-cost demonstration of technology readiness in the mid-2010s, with testing beginning as early as 2012. With space fission power system technology demonstrated, exploration mission planners will have the flexibility to respond to a broad variety of missions and will be able to provide abundant power so that future explorers will, in planning or crisis, have the power they need when they most need it.

TEM characterization of irradiated U-7Mo/Mg dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gan, J.; Keiser, D. D.; Miller, B. D.

This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 10 21 f/cm 3, 7.4 × 10 14 f/cm 3/s and 123 °C, and 5.5 × 10more » 21 f/cm3, 11.0 × 10 14 f/cm 3/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al 3Mg 2 and Al 12Mg 17 along with precipitates of MgO, Mg 2Si and FeAl 5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.« less

TEM characterization of irradiated U-7Mo/Mg dispersion fuel

DOE PAGES

Gan, J.; Keiser, D. D.; Miller, B. D.; ...

2017-07-15

This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 10 21 f/cm 3, 7.4 × 10 14 f/cm 3/s and 123 °C, and 5.5 × 10more » 21 f/cm3, 11.0 × 10 14 f/cm 3/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al 3Mg 2 and Al 12Mg 17 along with precipitates of MgO, Mg 2Si and FeAl 5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.« less

Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ludewigt, Bernhard; Mozin, Vladimir; Campbell, Luke

2015-06-01

High-­energy, beta-delayed gamma-­ray spectroscopy is a potential, non-­destructive assay techniques for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle and for directly assaying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification. Experimental measurements were performed to evaluate fission fragment yields, to test methods for determining isotopic fractions, and to benchmark the modeling code package. Experimental measurement campaigns were carried outmore » at the IAC using a photo-­neutron source and at OSU using a thermal neutron beam from the TRIGA reactor to characterize the emission of high-­energy delayed gamma rays from 235U, 239Pu, and 241Pu targets following neutron induced fission. Data were collected for pure and combined targets for several irradiation/spectroscopy cycle times ranging from 10/10 seconds to 15/30 minutes.The delayed gamma-ray signature of 241Pu, a significant fissile constituent in spent fuel, was measured and compared to 239Pu. The 241Pu/ 239Pu ratios varied between 0.5 and 1.2 for ten prominent lines in the 2700-­3600 keV energy range. Such significant differences in relative peak intensities make it possible to determine relative fractions of these isotopes in a mixed sample. A method for determining fission product yields by fitting the energy and time dependence of the delayed gamma-­ray emission was developed and demonstrated on a limited 235U data set. De-­convolution methods for determining fissile fractions were developed and tested on the experimental data. The use of high count-­rate LaBr 3 detectors was investigated as a potential alternative to HPGe detectors. Modeling capabilities were added to an existing framework and codes were adapted as needed for analyzing experiments and assessing application-­specific assay concepts. A de-­convolution analysis of the delayed gamma-­ray response spectra modeled for spent fuel assemblies was performed using the same method that was applied to the experimental spectra.« less

The role of fission on neutron star mergers and its impact on the r-process peaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eichler, M., E-mail: marius.eichler@unibas.ch; Thielemann, F.-K.; Arcones, A.

2016-06-21

The comparison between observational abundance features and those obtained from nucleosynthesis predictions of stellar evolution and/or explosion simulations can scrutinize two aspects: (a) the conditions in the astrophysical production site and (b) the quality of the nuclear physics input utilized. Here we test the abundance features of r-process nucleosynthesis calculations using four different fission fragment distribution models. Furthermore, we explore the origin of a shift in the third r-process peak position in comparison with the solar r-process abundances which has been noticed in a number of merger nucleosynthesis predictions. We show that this shift occurs during the r-process freeze-out whenmore » neutron captures and β-decays compete and an (n,γ)-(γ,n) equilibrium is not maintained anymore. During this phase neutrons originate mainly from fission of material above A = 240. We also investigate the role of β-decay half-lives from recent theoretical advances, which lead either to a smaller amount of fissioning nuclei during freeze-out or a faster (and thus earlier) release of fission neutrons, which can (partially) prevent this shift and has an impact on the second and rare-earth peak as well.« less

Neutron induced fission of 237Np - status, challenges and opportunities

NASA Astrophysics Data System (ADS)

Ruskov, Ivan; Goverdovski, Andrei; Furman, Walter; Kopatch, Yury; Shcherbakov, Oleg; Hambsch, Franz-Josef; Oberstedt, Stephan; Oberstedt, Andreas

2018-03-01

Nowadays, there is an increased interest in a complete study of the neutron-induced fission of 237Np. This is due to the need of accurate and reliable nuclear data for nuclear science and technology. 237Np is generated (and accumulated) in the nuclear reactor core during reactor operation. As one of the most abundant long-lived isotopes in spent fuel ("waste"), the incineration of 237Np becomes an important issue. One scenario for burning of 237Np and other radio-toxic minor actinides suggests they are to be mixed into the fuel of future fast-neutron reactors, employing the so-called transmutation and partitioning technology. For testing present fission models, which are at the basis of new generation nuclear reactor developments, highly accurate and detailed neutron-induced nuclear reaction data is needed. However, the EXFOR nuclear database for 237Np on neutron-induced capture cross-section, σγ, and fission cross-section, σf, as well as on the characteristics of capture and fission resonance parameters (Γγ, Γf, σoΓf, fragments mass-energy yield distributions, multiplicities of neutrons vn and γ-rays vγ), has not been updated for decades.

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harp, Jason Michael; Stempien, John Dennis; Demkowicz, Paul Andrew

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO 2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. Thesemore » data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO 2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO 2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These datamore » were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

Online Oxide Contamination Measurement and Purification Demonstration

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Godfroy, T. J.; Webster, K. L.; Garber, A. E.; Polzin, K. A.; Childers, D. J.

2011-01-01

Liquid metal sodium-potassium (NaK) has advantageous thermodynamic properties indicating its use as a fission reactor coolant for a surface (lunar, martian) power system. A major area of concern for fission reactor cooling systems is system corrosion due to oxygen contaminants at the high operating temperatures experienced. A small-scale, approximately 4-L capacity, simulated fission reactor cooling system employing NaK as a coolant was fabricated and tested with the goal of demonstrating a noninvasive oxygen detection and purification system. In order to generate prototypical conditions in the simulated cooling system, several system components were designed, fabricated, and tested. These major components were a fully-sealed, magnetically-coupled mechanical NaK pump, a graphite element heated reservoir, a plugging indicator system, and a cold trap. All system components were successfully demonstrated at a maximum system flow rate of approximately 150 cc/s at temperatures up to 550 C. Coolant purification was accomplished using a cold trap before and after plugging operations which showed a relative reduction in oxygen content.

Irradiation behavior of the interaction product of U-Mo fuel particle dispersion in an Al matrix

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Hofman, G. L.

2012-06-01

Irradiation performance of U-Mo fuel particles dispersed in Al matrix is stable in terms of fuel swelling and is suitable for the conversion of research and test reactors from highly enriched uranium (HEU) to low enriched uranium (LEU). However, tests of the fuel at high temperatures and high burnups revealed obstacles caused by the interaction layers forming between the fuel particle and matrix. In some cases, fission gas filled pores grow and interconnect in the interdiffusion layer resulting in fuel plate failure. Postirradiation observations are made to examine the behavior of the interdiffusion layers. The interdiffusion layers show a fluid-like behavior characteristic of amorphous materials. In the amorphous interdiffusion layers, fission gas diffusivity is high and the material viscosity is low so that the fission gas pores readily form and grow. Based on the observations, a pore formation mechanism is proposed and potential remedies to suppress the pore growth are also introduced.

Development of a full-length external-fuel thermionic converter for in-pile testing.

NASA Technical Reports Server (NTRS)

Schock, A.; Raab, B.

1971-01-01

Description of an external-fuel thermionic converter which utilizes a thoriated-tungsten fuel-emitter body. Performance in out-of-pile tests was comparable to that of an arc-cast tungsten emitter body, with 400-eW output power (about 5 W/sq cm) at 10.8% efficiency. Maximum fuel clad temperature averaged from 1650 to 1700 C during the 300-hour test. This converter has been processed for in-pile testing. The various processing steps, including the installation of six emitter thermocouples, encapsulation in the secondary container, and joining to the fission-gas collection system, are described in detail. In addition to the converter assembly, a doubly contained fission gas collection assembly with radiation-hardened differential pressure transducers was fabricated. The experiment support plate required for the in-pile test, containing electrically insulated instrumentation feedthroughs and coolant line feedthroughs to the vacuum test chamber, was also fabricated.

NEET Micro-Pocket Fission Detector. Final Project report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Unruh, T.; Rempe, Joy; McGregor, Douglas

2014-09-01

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Alternative Energies and Atomic Energy Commission, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), is funded by the Nuclear Energy Enabling Technologies (NEET) program to develop and test Micro-Pocket Fission Detectors (MPFDs), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package. When deployed, these sensors will significantly advance flux detection capabilities for irradiation tests in US Material Test Reactors (MTRs). Ultimately, evaluations may lead to a more compact, more accurate, andmore » longer lifetime flux sensor for critical mock-ups, and high performance reactors, allowing several Department of Energy Office of Nuclear Energy (DOE-NE) programs to obtain higher accuracy/higher resolution data from irradiation tests of candidate new fuels and materials. Specifically, deployment of MPFDs will address several challenges faced in irradiations performed at MTRs: Current fission chamber technologies do not offer the ability to measure fast flux, thermal flux and temperature within a single compact probe; MPFDs offer this option. MPFD construction is very different than current fission chamber construction; the use of high temperature materials allow MPFDs to be specifically tailored to survive harsh conditions encountered in-core of high performance MTRs. The higher accuracy, high fidelity data available from the compact MPFD will significantly enhance efforts to validate new high-fidelity reactor physics codes and new multi-scale, multi-physics codes. MPFDs can be built with variable sensitivities to survive the lifetime of an experiment or fuel assembly in some MTRs, allowing for more efficient and cost effective power monitoring. The small size of the MPFDs allows multiple sensors to be deployed, offering the potential to accurately measure the flux and temperature profiles in the reactor. This report summarizes the status at the end of year two of this three year project. As documented in this report, all planned accomplishments for developing this unique new, compact, multipurpose sensor have been completed.« less

Experimental Studies of NaK in a Simulated Space Environment

NASA Technical Reports Server (NTRS)

Gibons, Marc; Sanzi, James; Ljubanovic, Damir

2011-01-01

Space fission power systems are being developed at the National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) with a short term goal of building a full scale, non-nuclear, Technology Demonstration Unit (TDU) test at NASA's Glenn Research Center. Due to the geometric constraints, mass restrictions, and fairly high temperatures associated with space reactors, liquid metals are typically used as the primary coolant. A eutectic mixture of sodium (22 percent) and potassium (78 percent), or NaK, has been chosen as the coolant for the TDU with a total system capacity of approximately 55 L. NaK, like all alkali metals, is very reactive, and warrants certain safety considerations. To adequately examine the risk associated with the personnel, facility, and test hardware during a potential NaK leak in the large scale TDU test, a small scale experiment was performed in which NaK was released in a thermal vacuum chamber under controlled conditions. The study focused on detecting NaK leaks in the vacuum environment as well as the molecular flow of the NaK vapor. This paper reflects the work completed during the NaK experiment and provides results and discussion relative to the findings.

Stainless Steel NaK-Cooled Circuit (SNaKC) Fabrication and Assembly

NASA Technical Reports Server (NTRS)

Godfroy, Thomas J.

2007-01-01

An actively pumped Stainless Steel NaK Circuit (SNaKC) has been designed and fabricated by the Early Flight Fission Test Facility (EFF-TF) team at NASA's Marshall Space Flight Center. This circuit uses the eutectic mixture of sodium and potassium (NaK) as the working fluid building upon the experience and accomplishments of the SNAP reactor program from the late 1960's The SNaKC enables valuable experience and liquid metal test capability to be gained toward the goal of designing and building an affordable surface power reactor. The basic circuit components include a simulated reactor core a NaK to gas heat exchanger, an electromagnetic (EM) liquid metal pump, a liquid metal flow meter, an expansion reservoir and a drain/fill reservoir To maintain an oxygen free environment in the presence of NaK, an argon system is utilized. A helium and nitrogen system are utilized for core, pump, and heat exchanger operation. An additional rest section is available to enable special component testing m an elevated temperature actively pumped liquid metal environment. This paper summarizes the physical build of the SNaKC the gas and pressurization systems, vacuum systems, as well as instrumentation and control methods.

Studies of Lanthanide Transport in Metallic Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Jinsuo; Taylor, Christopher

Metallic nuclear fuels were tested in fast reactor programs and performed well. However, metallic fuels have shown the phenomenon of FCCI that are due to deleterious reactions between lanthanide fission products and cladding material. As the burnup is increased, lanthanide fission products that contact with the cladding could react with cladding constituents such as iron and chrome. These reactions produce higher-melting intermetallic compounds and low-melting alloys, and weaken the mechanical integrity.

The U.S. Geological Survey's TRIGA® reactor

USGS Publications Warehouse

DeBey, Timothy M.; Roy, Brycen R.; Brady, Sally R.

2012-01-01

The U.S. Geological Survey (USGS) operates a low-enriched uranium-fueled, pool-type reactor located at the Federal Center in Denver, Colorado. The mission of the Geological Survey TRIGA® Reactor (GSTR) is to support USGS science by providing information on geologic, plant, and animal specimens to advance methods and techniques unique to nuclear reactors. The reactor facility is supported by programs across the USGS and is organizationally under the Associate Director for Energy and Minerals, and Environmental Health. The GSTR is the only facility in the United States capable of performing automated delayed neutron analyses for detecting fissile and fissionable isotopes. Samples from around the world are submitted to the USGS for analysis using the reactor facility. Qualitative and quantitative elemental analyses, spatial elemental analyses, and geochronology are performed. Few research reactor facilities in the United States are equipped to handle the large number of samples processed at the GSTR. Historically, more than 450,000 sample irradiations have been performed at the USGS facility. Providing impartial scientific information to resource managers, planners, and other interested parties throughout the world is an integral part of the research effort of the USGS.

Examination of T-111 clad uranium nitride fuel pins irradiated up to 13,000 hours at a clad temperature of 990 C

NASA Technical Reports Server (NTRS)

Slaby, J. G.; Siegel, B. L.

1973-01-01

The examination of 27 fuel pins irradiated for up to 13,000 hours at 990 C is described. The fuel pin clad was a tantalum alloy with uranium nitride as the nuclear fuel. Two nominal fuel pin diameters were tested with a maximum burnup of 2.34 atom percent. Twenty-two fuel pins were tested for fission gas leaks; thirteen pins leaked. Clad ductility tests indicated clad embrittlement. The embrittlement is attributed to hydrogen from an n,p reaction in the fuel. Fuel swelling was burnup dependent, and the amount of fission gas release was low, generally less than 0.5 percent. No incompatibilities between fuel, liner, and clad were in evidence.

Interpretation and modelling of fission product Ba and Mo releases from fuel

NASA Astrophysics Data System (ADS)

Brillant, G.

2010-02-01

The release mechanisms of two fission products (namely barium and molybdenum) in severe accident conditions are studied using the VERCORS experimental observations. Barium is observed to be mostly released under reducing conditions while molybdenum release is most observed under oxidizing conditions. As well, the volatility of some precipitates in fuel is evaluated by thermodynamic equilibrium calculations. The polymeric species (MoO 3) n are calculated to largely contribute to molybdenum partial pressure and barium volatility is greatly enhanced if the gas atmosphere is reducing. Analytical models of fission product release from fuel are proposed for barium and molybdenum. Finally, these models have been integrated in the ASTEC/ELSA code and validation calculations have been performed on several experimental tests.

Thermal-stress fracture and fractography in UO/sub 2/

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kennedy, C.R.; Bandyopadhyay, G.

1976-01-01

Pressed and sintered UO/sub 2/ pellets were thermally shocked by quenching into a water bath at room temperature. The cracking behavior and strength degradation, as measured by the diametral compression technique, in these quench tests are discussed. Fractography of the thermally shocked specimens by scanning-electron microscopy indicated predominantly intergranular fracture in UO/sub 2/ in severe thermal-shock tests. The implication of this observation is that intergranular cracking may occur during the initial heat up in a reactor. Because fission gas bubbles tend to migrate toward the grain boundary, preferential microcracking along the boundary may strongly affect subsequent fission gas release behavior.

Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

DOEpatents

Lloyd, M.H.

1981-01-09

Method for direct coprocessing of nuclear fuels derived from a product stream of fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

DOEpatents

Lloyd, Milton H.

1983-01-01

Method for direct coprocessing of nuclear fuels derived from a product stream of a fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

Radioxenon spiked air

DOE PAGES

Watrous, Matthew G.; Delmore, James E.; Hague, Robert K.; ...

2015-08-27

Four of the radioactive xenon isotopes ( 131mXe, 133mXe, 133Xe and 135Xe) with half-lives ranging from 9 h to 12 days are produced from nuclear fission and can be detected from days to weeks following their production and release. Being inert gases, they are readily transported through the atmosphere. Sources for release of radioactive xenon isotopes include operating nuclear reactors via leaks in fuel rods, medical isotope production facilities, and nuclear weapons' detonations. They are not normally released from fuel reprocessing due to the short half-lives. The Comprehensive Nuclear-Test-Ban Treaty has led to creation of the International Monitoring System. Themore » International Monitoring System, when fully implemented, will consist of one component with 40 stations monitoring radioactive xenon around the globe. Monitoring these radioactive xenon isotopes is important to the Comprehensive Nuclear-Test-Ban Treaty in determining whether a seismically detected event is or is not a nuclear detonation. A variety of radioactive xenon quality control check standards, quantitatively spiked into various gas matrices, could be used to demonstrate that these stations are operating on the same basis in order to bolster defensibility of data across the International Monitoring System. This study focuses on Idaho National Laboratory's capability to produce three of the xenon isotopes in pure form and the use of the four xenon isotopes in various combinations to produce radioactive xenon spiked air samples that could be subsequently distributed to participating facilities.« less

Radioxenon spiked air

DOE Office of Scientific and Technical Information (OSTI.GOV)

Watrous, Matthew G.; Delmore, James E.; Hague, Robert K.

Four of the radioactive xenon isotopes ( 131mXe, 133mXe, 133Xe and 135Xe) with half-lives ranging from 9 h to 12 days are produced from nuclear fission and can be detected from days to weeks following their production and release. Being inert gases, they are readily transported through the atmosphere. Sources for release of radioactive xenon isotopes include operating nuclear reactors via leaks in fuel rods, medical isotope production facilities, and nuclear weapons' detonations. They are not normally released from fuel reprocessing due to the short half-lives. The Comprehensive Nuclear-Test-Ban Treaty has led to creation of the International Monitoring System. Themore » International Monitoring System, when fully implemented, will consist of one component with 40 stations monitoring radioactive xenon around the globe. Monitoring these radioactive xenon isotopes is important to the Comprehensive Nuclear-Test-Ban Treaty in determining whether a seismically detected event is or is not a nuclear detonation. A variety of radioactive xenon quality control check standards, quantitatively spiked into various gas matrices, could be used to demonstrate that these stations are operating on the same basis in order to bolster defensibility of data across the International Monitoring System. This study focuses on Idaho National Laboratory's capability to produce three of the xenon isotopes in pure form and the use of the four xenon isotopes in various combinations to produce radioactive xenon spiked air samples that could be subsequently distributed to participating facilities.« less

Heavy ion linear accelerator for radiation damage studies of materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kutsaev, Sergey V.; Mustapha, Brahim; Ostroumov, Peter N.

A new eXtreme MATerial (XMAT) research facility is being proposed at Argonne National Laboratory to enable rapid in situ mesoscale bulk analysis of ion radiation damage in advanced materials and nuclear fuels. This facility combines a new heavy-ion accelerator with the existing high-energy X-ray analysis capability of the Argonne Advanced Photon Source. The heavy-ion accelerator and target complex will enable experimenters to emulate the environment of a nuclear reactor making possible the study of fission fragment damage in materials. Material scientists will be able to use the measured material parameters to validate computer simulation codes and extrapolate the response ofmore » the material in a nuclear reactor environment. Utilizing a new heavy-ion accelerator will provide the appropriate energies and intensities to study these effects with beam intensities which allow experiments to run over hours or days instead of years. The XMAT facility will use a CW heavy-ion accelerator capable of providing beams of any stable isotope with adjustable energy up to 1.2 MeV/u for U-238(50+) and 1.7 MeV for protons. This energy is crucial to the design since it well mimics fission fragments that provide the major portion of the damage in nuclear fuels. The energy also allows damage to be created far from the surface of the material allowing bulk radiation damage effects to be investigated. The XMAT ion linac includes an electron cyclotron resonance ion source, a normal-conducting radio-frequency quadrupole and four normal-conducting multi-gap quarter-wave resonators operating at 60.625 MHz. This paper presents the 3D multi-physics design and analysis of the accelerating structures and beam dynamics studies of the linac.« less

Heavy ion linear accelerator for radiation damage studies of materials

NASA Astrophysics Data System (ADS)

Kutsaev, Sergey V.; Mustapha, Brahim; Ostroumov, Peter N.; Nolen, Jerry; Barcikowski, Albert; Pellin, Michael; Yacout, Abdellatif

2017-03-01

A new eXtreme MATerial (XMAT) research facility is being proposed at Argonne National Laboratory to enable rapid in situ mesoscale bulk analysis of ion radiation damage in advanced materials and nuclear fuels. This facility combines a new heavy-ion accelerator with the existing high-energy X-ray analysis capability of the Argonne Advanced Photon Source. The heavy-ion accelerator and target complex will enable experimenters to emulate the environment of a nuclear reactor making possible the study of fission fragment damage in materials. Material scientists will be able to use the measured material parameters to validate computer simulation codes and extrapolate the response of the material in a nuclear reactor environment. Utilizing a new heavy-ion accelerator will provide the appropriate energies and intensities to study these effects with beam intensities which allow experiments to run over hours or days instead of years. The XMAT facility will use a CW heavy-ion accelerator capable of providing beams of any stable isotope with adjustable energy up to 1.2 MeV/u for 238U50+ and 1.7 MeV for protons. This energy is crucial to the design since it well mimics fission fragments that provide the major portion of the damage in nuclear fuels. The energy also allows damage to be created far from the surface of the material allowing bulk radiation damage effects to be investigated. The XMAT ion linac includes an electron cyclotron resonance ion source, a normal-conducting radio-frequency quadrupole and four normal-conducting multi-gap quarter-wave resonators operating at 60.625 MHz. This paper presents the 3D multi-physics design and analysis of the accelerating structures and beam dynamics studies of the linac.

Nuclear criticality safety bounding analysis for the in-tank-precipitation (ITP) process, impacted by fissile isotopic weight fractions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, C.E.

The In-Tank Precipitation process (ITP) receives High Level Waste (HLW) supernatant liquid containing radionuclides in waste processing tank 48H. Sodium tetraphenylborate, NaTPB, and monosodium titanate (MST), NaTi{sub 2}O{sub 5}H, are added for removal of radioactive Cs and Sr, respectively. In addition to removal of radio-strontium, MST will also remove plutonium and uranium. The majority of the feed solutions to ITP will come from the dissolution of supernate that had been concentrated by evaporation to a crystallized salt form, commonly referred to as saltcake. The concern for criticality safety arises from the adsorption of U and Pt onto MST. If sufficientmore » mass and optimum conditions are achieved then criticality is credible. The concentration of u and Pt from solution into the smaller volume of precipitate represents a concern for criticality. This report supplements WSRC-TR-93-171, Nuclear Criticality Safety Bounding Analysis For The In-Tank-Precipitation (ITP) Process. Criticality safety in ITP can be analyzed by two bounding conditions: (1) the minimum safe ratio of MST to fissionable material and (2) the maximum fissionable material adsorption capacity of the MST. Calculations have provided the first bounding condition and experimental analysis has established the second. This report combines these conditions with canyon facility data to evaluate the potential for criticality in the ITP process due to the adsorption of the fissionable material from solution. In addition, this report analyzes the potential impact of increased U loading onto MST. Results of this analysis demonstrate a greater safety margin for ITP operations than the previous analysis. This report further demonstrates that the potential for criticality in the ITP process due to adsorption of fissionable material by MST is not credible.« less

Novel neutron sources at the Radiological Research Accelerator Facility

PubMed Central

Xu, Yanping; Garty, Guy; Marino, Stephen A.; Massey, Thomas N.; Randers-Pehrson, Gerhard; Johnson, Gary W.; Brenner, David J.

2012-01-01

Since the 1960s, the Radiological Research Accelerator Facility (RARAF) has been providing researchers in biology, chemistry and physics with advanced irradiation techniques, using charged particles, photons and neutrons. We are currently developing a unique facility at RARAF, to simulate neutron spectra from an improvised nuclear device (IND), based on calculations of the neutron spectrum at 1.5 km from the epicenter of the Hiroshima atom bomb. This is significantly different from a standard fission spectrum, because the spectrum changes as the neutrons are transported through air, and is dominated by neutron energies between 0.05 and 8 MeV. This facility will be based on a mixed proton/deuteron beam impinging on a thick beryllium target. A second, novel facility under development is our new neutron microbeam. The neutron microbeam will, for the first time, provide a kinematically collimated neutron beam, 10–20 micron in diameter. This facility is based on a Proton Microbeam, impinging on a thin lithium target near the threshold of the 7Li(p,n)7Be reaction. This novel neutron microbeam will enable studies of neutron damage to small targets, such as single cells, individual organs within small animals or microelectronic components. PMID:22545061

Novel neutron sources at the Radiological Research Accelerator Facility.

PubMed

Xu, Yanping; Garty, Guy; Marino, Stephen A; Massey, Thomas N; Randers-Pehrson, Gerhard; Johnson, Gary W; Brenner, David J

2012-03-16

Since the 1960s, the Radiological Research Accelerator Facility (RARAF) has been providing researchers in biology, chemistry and physics with advanced irradiation techniques, using charged particles, photons and neutrons.We are currently developing a unique facility at RARAF, to simulate neutron spectra from an improvised nuclear device (IND), based on calculations of the neutron spectrum at 1.5 km from the epicenter of the Hiroshima atom bomb. This is significantly different from a standard fission spectrum, because the spectrum changes as the neutrons are transported through air, and is dominated by neutron energies between 0.05 and 8 MeV. This facility will be based on a mixed proton/deuteron beam impinging on a thick beryllium target.A second, novel facility under development is our new neutron microbeam. The neutron microbeam will, for the first time, provide a kinematically collimated neutron beam, 10-20 micron in diameter. This facility is based on a Proton Microbeam, impinging on a thin lithium target near the threshold of the (7)Li(p,n)(7)Be reaction. This novel neutron microbeam will enable studies of neutron damage to small targets, such as single cells, individual organs within small animals or microelectronic components.

Novel neutron sources at the Radiological Research Accelerator Facility

DOE PAGES

Xu, Yanping; Garty, G.; Marino, S. A.; ...

2012-03-16

Since the 1960s, the Radiological Research Accelerator Facility (RARAF) has been providing researchers in biology, chemistry and physics with advanced irradiation techniques, using charged particles, photons and neutrons. We are currently developing a unique facility at RARAF, to simulate neutron spectra from an improvised nuclear device (IND), based on calculations of the neutron spectrum at 1.5 km from the epicenter of the Hiroshima atom bomb. This is significantly different from a standard fission spectrum, because the spectrum changes as the neutrons are transported through air, and is dominated by neutron energies between 0.05 and 8 MeV. This facility will bemore » based on a mixed proton/deuteron beam impinging on a thick beryllium target. A second, novel facility under development is our new neutron microbeam. The neutron microbeam will, for the first time, provide a kinematically collimated neutron beam, 10-20 micron in diameter. This facility is based on a proton microbeam, impinging on a thin lithium target near the threshold of the Li-7(p,n)Be-7 reaction. Lastly, this novel neutron microbeam will enable studies of neutron damage to small targets, such as single cells, individual organs within small animals or microelectronic components.« less

Novel neutron sources at the Radiological Research Accelerator Facility

NASA Astrophysics Data System (ADS)

Xu, Y.; Garty, G.; Marino, S. A.; Massey, T. N.; Randers-Pehrson, G.; Johnson, G. W.; Brenner, D. J.

2012-03-01

Since the 1960s, the Radiological Research Accelerator Facility (RARAF) has been providing researchers in biology, chemistry and physics with advanced irradiation techniques, using charged particles, photons and neutrons. We are currently developing a unique facility at RARAF, to simulate neutron spectra from an improvised nuclear device (IND), based on calculations of the neutron spectrum at 1.5 km from the epicenter of the Hiroshima atom bomb. This is significantly different from a standard fission spectrum, because the spectrum changes as the neutrons are transported through air, and is dominated by neutron energies between 0.05 and 8 MeV. This facility will be based on a mixed proton/deuteron beam impinging on a thick beryllium target. A second, novel facility under development is our new neutron microbeam. The neutron microbeam will, for the first time, provide a kinematically collimated neutron beam, 10-20 micron in diameter. This facility is based on a proton microbeam, impinging on a thin lithium target near the threshold of the 7Li(p,n)7Be reaction. This novel neutron microbeam will enable studies of neutron damage to small targets, such as single cells, individual organs within small animals or microelectronic components.

THE ROLE OF FISSION IN NEUTRON STAR MERGERS AND ITS IMPACT ON THE r-PROCESS PEAKS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eichler, M.; Panov, I.; Rauscher, T.

2015-07-20

Comparing observational abundance features with nucleosynthesis predictions of stellar evolution or explosion simulations, we can scrutinize two aspects: (a) the conditions in the astrophysical production site and (b) the quality of the nuclear physics input utilized. We test the abundance features of r-process nucleosynthesis calculations for the dynamical ejecta of neutron star merger simulations based on three different nuclear mass models: The Finite Range Droplet Model, the (quenched version of the) Extended Thomas Fermi Model with Strutinsky Integral, and the Hartree–Fock–Bogoliubov mass model. We make use of corresponding fission barrier heights and compare the impact of four different fission fragmentmore » distribution models on the final r-process abundance distribution. In particular, we explore the abundance distribution in the second r-process peak and the rare-earth sub-peak as a function of mass models and fission fragment distributions, as well as the origin of a shift in the third r-process peak position. The latter has been noticed in a number of merger nucleosynthesis predictions. We show that the shift occurs during the r-process freeze-out when neutron captures and β-decays compete and an (n,γ)–(γ,n) equilibrium is no longer maintained. During this phase neutrons originate mainly from fission of material above A = 240. We also investigate the role of β-decay half-lives from recent theoretical advances, which lead either to a smaller amount of fissioning nuclei during freeze-out or a faster (and thus earlier) release of fission neutrons, which can (partially) prevent this shift and has an impact on the second and rare-earth peak as well.« less

Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

DOE PAGES

Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.; ...

2016-05-18

The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium,more » and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.« less

Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.

The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium,more » and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.« less

Measurement of activation of helium gas by 238U beam irradiation at about 11 A MeV

NASA Astrophysics Data System (ADS)

Akashio, A.; Tanaka, K.; Imao, H.; Uwamino, Y.

2017-09-01

A new helium-gas stripper system has been applied at the 11 A MeV uranium beam of the Radioactive Isotope Beam Factory of the RIKEN accelerator facility. Although the gas stripper is important for the heavy-ion accelerator facility, the residual radiation that is generated is a serious problem for maintenance work. The residual dose was evaluated by using three-layered activation samples of aluminium and bismuth. The γ-rays from produced radionuclides with in-flight fission of the 238U beam and from the material of the chamber activated by neutrons were observed by using a Ge detector and compared with the values calculated by using the Monte-Carlo simulation code PHITS.

GEND planning report

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

The Three Mile Island (TMI) Unit 2 accident on March 28, 1979 was and is of great concern to the nuclear industry; electric power generating companies and their customers, regulatory and other government agencies, the entire nuclear community, and to the country as a whole. While the accident resulted in only limited external plant radiation exposure, the plant itself suffered extensive damage with high radiation contamination within the reactor and auxiliary system facilities. The GEND Planning Report for cleanup activities at TMI-2 covers the areas of: instrumentation and electrical equipment survivability; fission product transport; decontamination/radiation dose reduction technology; data bankmore » organization and sample archive facility; characterization of primary system pressure boundary and mechanical components; core damage assessment; and fuel handling, removal, examination and disposal.« less

Neutron cross-sections for next generation reactors: new data from n_TOF.

PubMed

Colonna, N; Abbondanno, U; Aerts, G; Alvarez, H; Alvarez-Velarde, F; Andriamonje, S; Andrzejewski, J; Assimakopoulos, P; Audouin, L; Badurek, G; Baumann, P; Becvar, F; Berthoumieux, E; Calviani, M; Calviño, F; Cano-Ott, D; Capote, R; de Albornoz, A Carrillo; Cennini, P; Chepel, V; Chiaveri, E; Cortes, G; Couture, A; Cox, J; Dahlfors, M; David, S; Dillman, I; Dolfini, R; Domingo-Pardo, C; Dridi, W; Duran, I; Eleftheriadis, C; Ferrant, L; Ferrari, A; Ferreira-Marques, R; Frais-Koelbl, H; Fujii, K; Furman, W; Goncalves, I; González-Romero, E; Goverdovski, A; Gramegna, F; Griesmayer, E; Guerrero, C; Gunsing, F; Haas, B; Haight, R; Heil, M; Herrera-Martinez, A; Igashira, M; Isaev, S; Jericha, E; Käppeler, F; Kadi, Y; Karadimos, D; Karamanis, D; Kerveno, M; Ketlerov, V; Koehler, P; Konovalov, V; Kossionides, E; Krticka, M; Lampoudis, C; Leeb, H; Lindote, A; Lopes, I; Lozano, M; Lukic, S; Marganiec, J; Marques, L; Marrone, S; Martínez, T; Massimi, C; Mastinu, P; Mengoni, A; Milazzo, P M; Moreau, C; Mosconi, M; Neves, F; Oberhummer, H; O'Brien, S; Oshima, M; Pancin, J; Papachristodoulou, C; Papadopoulos, C; Paradela, C; Patronis, N; Pavlik, A; Pavlopoulos, P; Perrot, L; Pigni, M T; Plag, R; Plompen, A; Plukis, A; Poch, A; Pretel, C; Quesada, J; Rauscher, T; Reifarth, R; Rosetti, M; Rubbia, C; Rudolf, G; Rullhusen, P; Salgado, J; Sarchiapone, L; Savvidis, I; Stephan, C; Tagliente, G; Tain, J L; Tassan-Got, L; Tavora, L; Terlizzi, R; Vannini, G; Vaz, P; Ventura, A; Villamarin, D; Vicente, M C; Vlachoudis, V; Vlastou, R; Voss, F; Walter, S; Wendler, H; Wiescher, M; Wisshak, K

2010-01-01

In 2002, an innovative neutron time-of-flight facility started operation at CERN: n_TOF. The main characteristics that make the new facility unique are the high instantaneous neutron flux, high resolution and wide energy range. Combined with state-of-the-art detectors and data acquisition system, these features have allowed to collect high accuracy neutron cross-section data on a variety of isotopes, many of which radioactive, of interest for Nuclear Astrophysics and for applications to advanced reactor technologies. A review of the most important results on capture and fission reactions obtained so far at n_TOF is presented, together with plans for new measurements related to nuclear industry. Copyright 2010 Elsevier Ltd. All rights reserved.

Testing of Liquid Metal Components for Nuclear Surface Power Systems

NASA Technical Reports Server (NTRS)

Polzin, Kurt A.; Godfroy, Thomas J.; Pearson, J. Boise

2010-01-01

The Early Flight Fission Test Facility (EFF-TF) was established by the Marshall Space Flight Center (MSFC) to provide a capability for performing hardware-directed activities to support multiple in-space nuclear reactor concepts by using a non-nuclear test methodology. This includes fabrication and testing at both the module/component level and near prototypic reactor configurations. The EFF-TF is currently supporting an effort to develop an affordable fission surface power (AFSP) system that could be deployed on the Lunar surface. The AFSP system is presently based on a pumped liquid metal-cooled (Sodium-Potassium eutectic, NaK-78) reactor design. This design was derived from the only fission system that the United States has deployed for space operation, the Systems for Nuclear Auxiliary Power (SNAP) 10A reactor, which was launched in 1965. Two prototypical components recently tested at MSFC were a pair of Stirling power conversion units that would be used in a reactor system to convert heat to electricity, and an annular linear induction pump (ALIP) that uses travelling electromagnetic fields to pump the liquid metal coolant through the reactor loop. First ever tests were conducted at MSFC to determine baseline performance of a pair of 1 kW Stirling convertors using NaK as the hot side working fluid. A special test rig was designed and constructed and testing was conducted inside a vacuum chamber at MSFC. This test rig delivered pumped NaK for the hot end temperature to the Stirlings and water as the working fluid on the cold end temperature. These test were conducted through a hot end temperature range between 400 to 550C in increments of 50 C and a cold end temperature range from 30 to 70 C in 20 C increments. Piston amplitudes were varied from 6 to 1 1mm in .5 mm increments. A maximum of 2240 Watts electric was produced at the design point of 550 hot end, 40 C cold end with a piston amplitude of 10.5mm. This power level was reached at a gross thermal efficiency of 28%. A baseline performance map was established for the pair of 1kW Stirling convertors. The performance data will then be used for design modification to the Stirling convertors. The ALIP tested at MSFC has no moving parts and no direct electrical connections to the liquid metal containing components. Pressure is developed by the interaction of the magnetic field produced by the stator and the current which flows as a result of the voltage induced in the liquid metal contained in the pump duct. Flow is controlled by variation of the voltage supplied to the pump windings. Under steady-state conditions, pump performance is measured for flow rates from 0.5-4.3 kg/s. The pressure rise developed by the pump to support these flow rates is roughly 5-65 kPa. The RMS input voltage (phase-to-phase voltage) ranges from 5-120 V, while the frequency can be varied arbitrarily up to 60 Hz. Performance is quantified at different loop temperature levels from 50 C up to 650 C, which is the peak operating temperature of the proposed AFSP reactor. The transient response of the pump is also evaluated to determine its behavior during startup and shut-down procedures.

Detailed measurements of local thickness changes for U-7Mo dispersion fuel plates with Al-3.5Si matrix after irradiation at different powers in the RERTR-9B experiment

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Williams, Walter; Robinson, Adam; Wachs, Dan; Moore, Glenn; Crawford, Doug

2017-10-01

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. Swelling is an important irradiation behavior that needs to be well understood. Data from high resolution thickness measurements performed on U-7Mo dispersion fuel plates with Al-Si alloy matrices that were irradiated at high power is sparse. This paper reports the results of detailed thickness measurements performed on two dispersion fuel plates that were irradiated at relatively high power to high fission densities in the Advanced Test Reactor in the same RERTR-9B experiment. Both plates were irradiated to similar fission densities, but one was irradiated at a higher power than the other. The goal of this work is to identify any differences in the swelling behavior when fuel plates are irradiated at different powers to the same fission densities. Based on the results of detailed thickness measurments, more swelling occurs when a U-7Mo dispersion fuel with Al-3.5Si matrix is irradiated to a high fission density at high power compared to one irradiated at a lower power to high fission density.

Mechanistic approach for nitride fuel evolution and fission product release under irradiation

NASA Astrophysics Data System (ADS)

Dolgodvorov, A. P.; Ozrin, V. D.

2017-01-01

A model for describing uranium-plutonium mixed nitride fuel pellet burning was developed. Except fission products generating, the model includes impurities of oxygen and carbon. Nitrogen behaviour in nitride fuel was analysed and the nitrogen chemical potential in solid solution with uranium-plutonium nitride was constructed. The chemical program module was tested with the help of thermodynamic equilibrium phase distribution calculation. Results were compared with analogous data in literature, quite good agreement was achieved, especially for uranium sesquinitride, metallic species and some oxides. Calculation of a process of nitride fuel burning was also conducted. Used mechanistic approaches for fission product evolution give the opportunity to find fission gas release fractions and also volumes of intergranular secondary phases. Calculations present that the most massive secondary phases are the oxide and metallic phases. Oxide phase contain approximately 1 % wt of substance over all time of burning with slightly increasing of content. Metallic phase has considerable rising of mass and by the last stage of burning it contains about 0.6 % wt of substance. Intermetallic phase has less increasing rate than metallic phase and include from 0.1 to 0.2 % wt over all time of burning. The highest element fractions of released gaseous fission products correspond to caesium and iodide.

Preliminary Evaluation of Convective Heat Transfer in a Water Shield for a Surface Power Reactor

NASA Technical Reports Server (NTRS)

Pearson J. Boise; Reid, Robert S.

2007-01-01

As part of the Vision for Space Exploration, the end of the next decade will bring man back to the surface of the moon. A crucial issue for the establishment of human presence on the moon will be the availability of compact power sources. This presence could require greater than 10's of kWt's in follow on years. Nuclear reactors are well suited to meet the needs for power generation on the lunar or Martian surface. Radiation shielding is a key component of any surface power reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), and boron carbide. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to fix the location of any vapor that could form radiation streaming paths. The water shield concept relies on the predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. NASA Marshall Space Flight Center has developed the experience and facilities necessary to do this evaluation in its Early Flight Fission - Test Facility (EFF-TF).

The Quest for Fusion at the National Ignition Facility

NASA Astrophysics Data System (ADS)

Hartouni, Edward

2017-01-01

Arthur Eddington speculated in 1920 on the internal constitution of stars and described the possibility of nuclear fusion based on the then new results from special relativity and measurements of light nuclei masses. By 1929 Atkinson and Houtermans worked out the calculations for nuclear fusion in stars and initiating nuclear astrophysics. All of these sciences were pressed into service during the World War II, and the applications developed, particularly under the auspices of the Manhattan Project provided both weapons with which to wage and win that conflict, but also the possibilities to harness these applications of the nuclear processes of fission and fusion for peaceful purposes. 32 years after Eddington's speculation the United States demonstrated the application of fusion in a famous nuclear weapons test. In the following years many ideas for producing ``controlled'' fusion through inertial confinement were pursued. The invention of the laser opened up new avenues which have culminated in the National Ignition Facility, NIF. I will attempt to cover the ground between Eddington, through the Manhattan Project and provide a current status of this quest at NIF. LLNL-ABS-704367-DRAFT. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

An End-To-End Test of A Simulated Nuclear Electric Propulsion System

NASA Technical Reports Server (NTRS)

VanDyke, Melissa; Hrbud, Ivana; Goddfellow, Keith; Rodgers, Stephen L. (Technical Monitor)

2002-01-01

The Safe Affordable Fission Engine (SAFE) test series addresses Phase I Space Fission Systems issues in it particular non-nuclear testing and system integration issues leading to the testing and non-nuclear demonstration of a 400-kW fully integrated flight unit. The first part of the SAFE 30 test series demonstrated operation of the simulated nuclear core and heat pipe system. Experimental data acquired in a number of different test scenarios will validate existing computational models, demonstrated system flexibility (fast start-ups, multiple start-ups/shut downs), simulate predictable failure modes and operating environments. The objective of the second part is to demonstrate an integrated propulsion system consisting of a core, conversion system and a thruster where the system converts thermal heat into jet power. This end-to-end system demonstration sets a precedent for ground testing of nuclear electric propulsion systems. The paper describes the SAFE 30 end-to-end system demonstration and its subsystems.

Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

NASA Astrophysics Data System (ADS)

Wright, Steven A.; Houts, Michael

2001-02-01

Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul Andrew

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A seriesmore » of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).« less

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul; Harp, Jason; Winston, Phil

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance, and fission product transport (INL 2015). A series ofmore » fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously.« less

The performance of hafnium and gadolinium self powered neutron detectors in the TREAT reactor

NASA Astrophysics Data System (ADS)

Imel, G. R.; Hart, P. R.

1996-05-01

The use of gadolinium and hafnium self powered neutron detectors in a transient reactor is described in this paper. The detectors were calibrated to the fission rate of U-235 using calibrated fission chambers; the calibration factors were tested in two reactors in steady state and found to be consistent. Calibration of the detectors in transient reactor conditions was done by using uranium wires that were analyzed by radiochemistry techniques to determine total fissions during the transient. This was correlated to the time-integrated current of the detectors during the transient. A temperature correction factor was derived to account for self-shielding effects in the hafnium and gadolinium detectors. The dynamic response of the detectors under transient conditions was studied, and found to be excellent.

Fukushima-derived fission nuclides monitored around Taiwan: Free tropospheric versus boundary layer transport

NASA Astrophysics Data System (ADS)

Huh, Chih-An; Hsu, Shih-Chieh; Lin, Chuan-Yao

2012-02-01

The 2011 Fukushima nuclear accident in Japan was the worst nuclear disaster following the 1986 Chernobyl accident. Fission products (nuclides) released from the Fukushima plant site since March 12, 2011 had been detected around the northern hemisphere in about two weeks and also in the southern hemisphere about one month later. We report here detailed time series of radioiodine and radiocesium isotopes monitored in a regional network around Taiwan, including one high-mountain and three ground-level sites. Our results show several pulses of emission from a sequence of accidents in the Fukushima facility, with the more volatile 131I released preferentially over 134Cs and 137Cs at the beginning. In the middle of the time series, there was a pronounced peak of radiocesium observed in northern Taiwan, with activity concentrations of 134Cs and 137Cs far exceeding that of 131I during that episode. From the first arrival time of these fission nuclides and their spatial and temporal variations at our sampling sites and elsewhere, we suggest that Fukushima-derived radioactive nuclides were transported to Taiwan and its vicinity via two pathways at different altitudes. One was transported in the free troposphere by the prevailing westerly winds around the globe; the other was transported in the planetary boundary layer by the northeast monsoon wind directly toward Taiwan.

Study of the fission spectrum of less than 1 MeV neutrons using a Lithium-glass detector

NASA Astrophysics Data System (ADS)

Bastola, Suraj; Rees, Lawrence; Bart, Czirr

2011-10-01

The fission spectrum of neutrons with kinetic energies less than 1 MeV is of considerable practical importance for the design of nuclear reactors. However, it is not as precisely known as that for higher energy neutrons. One of the major problems scientists have previously encountered is room return neutrons. These are neutrons that reflect from the walls, ceiling or floor of the lab. Another problem is finding a way to measure accurately the neutron time of flight. This is the time neutrons take to travel from a fission event to the detector. Time of flight is used to measure the neutron energy. To avoid the room return, I am going to perform an experiment about 45 feet above the ground in the BYU Indoor Practice Facility, so that neutrons from the source will not scatter from nearby surfaces and return to the detector. To find the time of flight to a greater accuracy, I have been using a Time to Amplitude Converter (TAC). A TAC has a capacitor that charges linearly as the voltage builds up. With a 12-bit digitizer system, we can measure the time to 0.1 nanoseconds, whereas the same digitizer can only measure time in steps of 4 nanoseconds. So, we will get a more accurate measurement of time of flight with the TAC.

ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology

NASA Astrophysics Data System (ADS)

Chadwick, M. B.; Obložinský, P.; Herman, M.; Greene, N. M.; McKnight, R. D.; Smith, D. L.; Young, P. G.; MacFarlane, R. E.; Hale, G. M.; Frankle, S. C.; Kahler, A. C.; Kawano, T.; Little, R. C.; Madland, D. G.; Moller, P.; Mosteller, R. D.; Page, P. R.; Talou, P.; Trellue, H.; White, M. C.; Wilson, W. B.; Arcilla, R.; Dunford, C. L.; Mughabghab, S. F.; Pritychenko, B.; Rochman, D.; Sonzogni, A. A.; Lubitz, C. R.; Trumbull, T. H.; Weinman, J. P.; Brown, D. A.; Cullen, D. E.; Heinrichs, D. P.; McNabb, D. P.; Derrien, H.; Dunn, M. E.; Larson, N. M.; Leal, L. C.; Carlson, A. D.; Block, R. C.; Briggs, J. B.; Cheng, E. T.; Huria, H. C.; Zerkle, M. L.; Kozier, K. S.; Courcelle, A.; Pronyaev, V.; van der Marck, S. C.

2006-12-01

We describe the next generation general purpose Evaluated Nuclear Data File, ENDF/B-VII.0, of recommended nuclear data for advanced nuclear science and technology applications. The library, released by the U.S. Cross Section Evaluation Working Group (CSEWG) in December 2006, contains data primarily for reactions with incident neutrons, protons, and photons on almost 400 isotopes, based on experimental data and theory predictions. The principal advances over the previous ENDF/B-VI library are the following: (1) New cross sections for U, Pu, Th, Np and Am actinide isotopes, with improved performance in integral validation criticality and neutron transmission benchmark tests; (2) More precise standard cross sections for neutron reactions on H, 6Li, 10B, Au and for 235,238U fission, developed by a collaboration with the IAEA and the OECD/NEA Working Party on Evaluation Cooperation (WPEC); (3) Improved thermal neutron scattering; (4) An extensive set of neutron cross sections on fission products developed through a WPEC collaboration; (5) A large suite of photonuclear reactions; (6) Extension of many neutron- and proton-induced evaluations up to 150 MeV; (7) Many new light nucleus neutron and proton reactions; (8) Post-fission beta-delayed photon decay spectra; (9) New radioactive decay data; (10) New methods for uncertainties and covariances, together with covariance evaluations for some sample cases; and (11) New actinide fission energy deposition. The paper provides an overview of this library, consisting of 14 sublibraries in the same ENDF-6 format as the earlier ENDF/B-VI library. We describe each of the 14 sublibraries, focusing on neutron reactions. Extensive validation, using radiation transport codes to simulate measured critical assemblies, show major improvements: (a) The long-standing underprediction of low enriched uranium thermal assemblies is removed; (b) The 238U and 208Pb reflector biases in fast systems are largely removed; (c) ENDF/B-VI.8 good agreement for simulations of thermal high-enriched uranium assemblies is preserved; (d) The underprediction of fast criticality of 233,235U and 239Pu assemblies is removed; and (e) The intermediate spectrum critical assemblies are predicted more accurately. We anticipate that the new library will play an important role in nuclear technology applications, including transport simulations supporting national security, nonproliferation, advanced reactor and fuel cycle concepts, criticality safety, fusion, medicine, space applications, nuclear astrophysics, and nuclear physics facility design. The ENDF/B-VII.0 library is archived at the National Nuclear Data Center, BNL, and can be retrieved from www.nndc.bnl.gov.

Photofission cross-section ratio measurement of 235U/238U using monoenergetic photons in the energy range of 9.0-16.6 MeV

NASA Astrophysics Data System (ADS)

Krishichayan; Bhike, Megha; Finch, S. W.; Howell, C. R.; Tonchev, A. P.; Tornow, W.

2017-05-01

Photofission cross-section ratios of 235U and 238U have been measured using monoenergetic photon beams at the HIγS facility of TUNL. These measurements have been performed in small energy steps between 9.0 and 16.6 MeV using a dual-fission ionization chamber. Measured cross-section ratios are compared with the previous experimental data as well as with the recent evaluated nuclear data library ENDF.

LDEF Experiment P0006 Linear Energy Transfer Spectrum Measurement (LETSME) quick look report

NASA Technical Reports Server (NTRS)

1990-01-01

A preliminary analysis of the various passive radiation detector materials included in the P0006 LETSME experiment flown on LDEF (Long Duration Exposure Facility) is presented. It consists of four tasks: (1) readout and analysis of thermoluminescent detectors (TLD); (2) readout and analysis of fission foil/mica detectors; (3) readout and analysis of (6)LiF/CR-39 detectors; and (4) preliminary processing and readout of CR-39 and polycarbonate plastic nuclear track detectors (PNTD).

The fiftieth anniversary of the first public announcement of the successful test of fission: Proceedings

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This report contains comments and discussions on the history of fission. The following people comments and lectures are discussed in this report: Remarks and introduction of Maxine F. Singer; president's message, Maxine F. Singer; introduction of Stephen Joel Trachtenberg; President's message, Stephen Joel Trachtenberg; introduction of Frederick Seitz; lecture: Nuclear Science: Promises and Perceptions, '' Frederick Seitz; introduction of K. Alex Mueller; lecture: High Temperature Ferroelectricity and Superconductivity,'' introduction of Edward Teller; and lecture: Toward a More Secure World,'' Edward Teller.

Evaluation of cryogenic insulation materials and composites for use in nuclear radiation environments

NASA Technical Reports Server (NTRS)

Bullock, R. E.

1972-01-01

The following subjects are studied: (1) composite materials tests; (2) test of liquid level sensors and fission couples; (3) test of valve-seal materials; (4) boron epoxy composites; (5) radiation analysis of explosive materials and bifuels for RNS applications; and (6) test of thermal insulation.

Heatpipe power system and heatpipe bimodal system design and development options

NASA Technical Reports Server (NTRS)

Houts, M. G.; Poston, D. I.; Emrich, W. J., Jr.

1997-01-01

The Heatpipe Power System (HPS) is a potential, near-term, low-cost space fission power system. The Heatpipe Bimodal System (HBS) is a potential, near-term, low-cost space fission power and/or propulsion system. Both systems will be composed of independent modules, and all components operate within the existing databases. The HPS and HBS have relatively few system integration issues; thus, the successful development of a module is a significant step toward verifying system feasibility and performance estimates. A prototypic HPS module is being fabricated, and testing is scheduled to begin in November 1996. A successful test will provide high confidence that the HPS can achieve its predicted performance.

MODELING AND ANALYSIS OF FISSION PRODUCT TRANSPORT IN THE AGR-3/4 EXPERIMENT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Humrickhouse, Paul W.; Collin, Blaise P.; Hawkes, Grant L.

In this work we describe the ongoing modeling and analysis efforts in support of the AGR-3/4 experiment. AGR-3/4 is intended to provide data to assess fission product retention and transport (e.g., diffusion coefficients) in fuel matrix and graphite materials. We describe a set of pre-test predictions that incorporate the results of detailed thermal and fission product release models into a coupled 1D radial diffusion model of the experiment, using diffusion coefficients reported in the literature for Ag, Cs, and Sr. We make some comparisons of the predicted Cs profiles to preliminary measured data for Cs and find these to bemore » reasonable, in most cases within an order of magnitude. Our ultimate objective is to refine the diffusion coefficients using AGR-3/4 data, so we identify an analytical method for doing so and demonstrate its efficacy via a series of numerical experiments using the model predictions. Finally, we discuss development of a post-irradiation examination plan informed by the modeling effort and simulate some of the heating tests that are tentatively planned.« less

Small Scale Characterization of the Presence of the Explosive Octahydro-1,3,5,7-tetranitro- 1,3,5,7 tetrazocine (HMX) Near Former Naval Sites on Vieques Island, Puerto Rico

NASA Astrophysics Data System (ADS)

Simmons, C. C.; Carvalho-Knighton, K. M.; Pyrtle, A. J.

2007-12-01

Octahydro-1,3,5,7-tetranitro-1,3,5,7 tetrazocine (HMX) is a synthetic energetic compounds that has been commonly used in military munitions. The presence and movement of HMX through the environment is of growing concern because of potential civilian exposure and impacts on human health. HMX remains in the environment unreactive with little degradation. It can be transported great distances in water thus having the possibility for migrating into groundwater. The former naval sites in Vieques were used for weapons training and housed several disposal sites. Previous studies around these sites indicate the presence of radioactive materials produced through thermal fission, such as Cs-137. Since HMX was primarily used to implode fissionable materials in nuclear devices, evaluating the release of HMX and consequent movement through the environment at these sites is essential. Surface water and soil samples as well as core and pore water samples were collected from two sites in Vieques; Kiani Lagoon and Mosquito Bay. All samples were extracted using EPA method 8330 and analyzed using RP-HPLC analysis with a C-18 column. HMX was undetected in samples collected from both Kiani Lagoon and Mosquito Bay. The development of a model that studies the flow rates and fate of water runoff in these areas of interest, coupled with data on groundwater testing inside the actual former naval facilities, is being explored for further sample collection and analysis.

Spent Nuclear Fuel Disposition

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagner, John C.

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

Spent Nuclear Fuel Disposition

DOE PAGES

Wagner, John C.

2016-05-22

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

Critical experiments at Sandia National Laboratories : technical meeting on low-power critical facilities and small reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harms, Gary A.; Ford, John T.; Barber, Allison Delo

2010-11-01

Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-IIImore » is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide benchmark reactor physics data to support validation of the reactor physics codes used to design commercial reactor fuel elements in an enrichment range above the current 5% enrichment cap. A first set of critical experiments in the 7uPCX has been completed. More experiments are planned in the 7uPCX series. The critical experiments at Sandia National Laboratories are currently funded by the US Department of Energy Nuclear Criticality Safety Program (NCSP). The NCSP has committed to maintain the critical experiment capability at Sandia and to support the development of a critical experiments training course at the facility. The training course is intended to provide hands-on experiment experience for the training of new and re-training of practicing Nuclear Criticality Safety Engineers. The current plans are for the development of the course to continue through the first part of fiscal year 2011 with the development culminating is the delivery of a prototype of the course in the latter part of the fiscal year. The course will be available in fiscal year 2012.« less

Research Activities in Fission Chamber Modeling in Support of the Nuclear Energy Industry

NASA Astrophysics Data System (ADS)

Jammes, C.; Filliatre, P.; Geslot, B.; Oriol, L.; Berhouet, F.; Villard, J.-F.; Vermeeren, L.

2010-12-01

Fission chambers are widely used in the nuclear industry. As an example, they play a major role in the control of any fission reactor and are thus regarded as a key component for ensuring their safety. They are also employed in the material testing reactors for monitoring irradiations. We have recently started a research program, the objective of which is to improve the performance of those neutron detectors in terms of lifetime, calibration, and online diagnosis. In this paper, we present several studies carried out in order to model the signal delivered by a fission chamber. First, the simulation of the deposit evolution allowed us to select the most appropriate fissile material for a given spectrum and fluence. Second, we studied the impact of the bias voltage and filling gas characteristics on the charge collection time. Finally, the simulation of a pulse signal prior to amplification showed how it is important to have a satisfactory knowledge of the energy for creating ion pairs to accurately assess the signal in current or Campbelling mode.

SOPHAEROS code development and its application to falcon tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lajtha, G.; Missirlian, M.; Kissane, M.

1996-12-31

One of the key issues in source-term evaluation in nuclear reactor severe accidents is determination of the transport behavior of fission products released from the degrading core. The SOPHAEROS computer code is being developed to predict fission product transport in a mechanistic way in light water reactor circuits. These applications of the SOPHAEROS code to the Falcon experiments, among others not presented here, indicate that the numerical scheme of the code is robust, and no convergence problems are encountered. The calculation is also very fast being three times longer on a Sun SPARC 5 workstation than real time and typicallymore » {approx} 10 times faster than an identical calculation with the VICTORIA code. The study demonstrates that the SOPHAEROS 1.3 code is a suitable tool for prediction of the vapor chemistry and fission product transport with a reasonable level of accuracy. Furthermore, the fexibility of the code material data bank allows improvement of understanding of fission product transport and deposition in the circuit. Performing sensitivity studies with different chemical species or with different properties (saturation pressure, chemical equilibrium constants) is very straightforward.« less

FROM THE HISTORY OF PHYSICS: The development of the first Soviet atomic bomb

NASA Astrophysics Data System (ADS)

Goncharov, German A.; Ryabev, Lev D.

2001-01-01

In the late 1930s and early 1940s, two remarkable physical phenomena — the fission of heavy nuclei and the chain fission reaction — were discovered, implying that a new powerful source of energy (nuclear fission energy) might become a practical possibility for mankind. At that time, however, the political situation in the world made the development of the atomic bomb the main objective of nuclear energy research in the countries involved. The first atomic bombs, notoriously used in the war against Japan, were produced by the United States of America only six and a half years after the discovery of fission. Four years later, the first Soviet atomic bomb was tested. This was a major step toward the establishment of nuclear parity which led to stability and global peace and thus greatly influenced the destiny of human kind. Based on documentary materials covering the period from 1939 to 1949, this paper traces the origin and evolution of the physical ideas behind the first Soviet atomic bomb and discusses the most important events associated with the project.

Accelerator-driven Transmutation of Waste

NASA Astrophysics Data System (ADS)

Venneri, Francesco

1998-04-01

Nuclear waste from commercial power plants contains large quantities of plutonium, other fissionable actinides, and long-lived fission products that are potential proliferation concerns and create challenges for the long-term storage. Different strategies for dealing with nuclear waste are being followed by various countries because of their geologic situations and their views on nuclear energy, reprocessing and non-proliferation. The current United States policy is to store unprocessed spent reactor fuel in a geologic repository. Other countries are opting for treatment of nuclear waste, including partial utilization of the fissile material contained in the spent fuel, prior to geologic storage. Long-term uncertainties are hampering the acceptability and eventual licensing of a geologic repository for nuclear spent fuel in the US, and driving up its cost. The greatest concerns are with the potential for radiation release and exposure from the spent fuel for tens of thousands of years and the possible diversion and use of the actinides contained in the waste for weapons construction. Taking advantage of the recent breakthroughs in accelerator technology and of the natural flexibility of subcritical systems, the Accelerator-driven Transmutation of Waste (ATW) concept offers the United States and other countries the possibility to greatly reduce plutonium, higher actinides and environmentally hazardous fission products from the waste stream destined for permanent storage. ATW does not eliminate the need for, but instead enhances the viability of permanent waste repositories. Far from being limited to waste destruction, the ATW concept also brings to the table new technologies that could be relevant for next-generation power producing reactors. In the ATW concept, spent fuel would be shipped to the ATW site where the plutonium, transuranics and selected long-lived fission products would be destroyed by fission or transmutation in their first and only pass through the facility, using an accelerator-driven subcritical burner cooled by liquid lead/bismuth and limited pyrochemical treatment of the spent fuel and residual waste. This approach contrasts with the present-day practices of aqueous reprocessing (Europe and Japan), in which high purity plutonium is produced and used in the fabrication of fresh mixed oxide fuel (MOX) that is shipped off-site for use in light water reactors.

Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and implementation of a virtual reactivity feedback control loop; testing and thermal hydraulic characterization of the coolant flow paths for two space reactor concepts; and analysis of heat pipe operation during start-up and steady state operation.

John von Neumann and Klaus Fuchs: an Unlikely Collaboration

NASA Astrophysics Data System (ADS)

Bernstein, Jeremy

2010-03-01

I discuss the origin of the idea of making a fusion (hydrogen) bomb and the physics involved in it, and then turn to the design proposed for one by the unlikely collaborators John von Neumann and Klaus Fuchs in a patent application they filed at Los Alamos in May 1946, which Fuchs passed on to the Russians in March 1948, and which with substantial modifications was tested on the island of Eberiru on the Eniwetok atoll in the South Pacific on May 8, 1951. This test showed that the fusion of deuterium and tritium nuclei could be ignited, but that the ignition would not propagate because the heat produced was rapidly radiated away. Meanwhile, Stanislaw Ulam and C.J. Everett had shown that Edward Teller’s Classical Super could not work, and at the end of December 1950, Ulam had conceived the idea of super compression, using the energy of a fission bomb to compress the fusion fuel to such a high density that it would be opaque to the radiation produced. Once Teller understood this, he invented a greatly improved, new method of compression using radiation, which then became the heart of the Ulam-Teller bomb design, which was tested, also in the South Pacific, on November 1, 1952. The Russians have freely acknowledged that Fuchs gave them the fission bomb, but they have insisted that no one gave them the fusion bomb, which grew out of design involving a fission bomb surrounded by alternating layers of fusion and fission fuels, and which they tested on November 22, 1955. Part of the irony of this story is that neither the American nor the Russian hydrogen-bomb programs made any use of the brilliant design that von Neumann and Fuchs had conceived as early as 1946, which could have changed the entire course of development of both programs.

Scientific Design of the New Neutron Radiography Facility (SANRAD) at SAFARI-1 for South Africa

NASA Astrophysics Data System (ADS)

de Beer, F. C.; Gruenauer, F.; Radebe, J. M.; Modise, T.; Schillinger, B.

The final scientific design for an upgraded neutron radiography/tomography facility at beam port no.2 of the SAFARI-1 nuclear research reactor has been performed through expert advice from Physics Consulting, FRMII in Germany and IPEN, Brazil. A need to upgrade the facility became apparent due to the identification of various deficiencies of the current SANRAD facility during an IAEA-sponsored expert mission of international scientists to Necsa, South Africa. A lack of adequate shielding that results in high neutron background on the beam port floor, a mismatch in the collimator aperture to the core that results in a high gradient in neutron flux on the imaging plane and due to a relative low L/D the quality of the radiographs are poor, are a number of deficiencies to name a few.The new design, based on results of Monte Carlo (MCNP-X) simulations of neutron- and gamma transport from the reactor core and through the new facility, is being outlined. The scientific design philosophy, neutron optics and imaging capabilities that include the utilization of fission neutrons, thermal neutrons, and gamma-rays emerging from the core of SAFARI-1 are discussed.

UCx target preparations and characterizations

NASA Astrophysics Data System (ADS)

Andrighetto, Alberto; Corradetti, Stefano; Manzolaro, Mattia; Scarpa, Daniele; Monetti, Alberto; Rossignoli, Massimo; Borgna, Francesca; Ballan, Michele; Agostini, Mattia; D'Agostini, Fabio; Ferrari, Matteo; Zenoni, Aldo

2018-05-01

The Target-Ion Source unit is the core of an ISOL-RIB facility. Many international ISOL facilities have chosen different layouts of this unit. Many research groups are involved in research and development of targets capable of dissipating high power and, at the same time, be able to have a fast isotope release. This is mandatory in order to produce beams of short half-life isotopes. The research of new materials with advanced microstructural features is crucial in this field. The design of a proper target is indeed strictly related to the obtainment of porous refractory materials, which are capable to work under extreme conditions (temperatures up to 2000 °C in high vacuum) with a high release efficiency. For SPES, the second generation Italian ISOL-RIB Facility, the target will be made of uranium carbide (UCx) in which, by fission induced by a proton beam of 40 MeV of energy (8 kW of power), isotopes in the 60-160 amu mass region are produced. The current technological developments are also crucial in the study of third generation ISOL facilities.

The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

NASA Astrophysics Data System (ADS)

Scarlat, Raluca O.; Peterson, Per F.

2014-01-01

The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

Development of DEMO-FNS tokamak for fusion and hybrid technologies

NASA Astrophysics Data System (ADS)

Kuteev, B. V.; Azizov, E. A.; Alexeev, P. N.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

2015-07-01

The history of fusion-fission hybrid systems based on a tokamak device as an extremely efficient DT-fusion neutron source has passed through several periods of ample research activity in the world since the very beginning of fusion research in the 1950s. Recently, a new roadmap of the hybrid program has been proposed with the goal to build a pilot hybrid plant (PHP) in Russia by 2030. Development of the DEMO-FNS tokamak for fusion and hybrid technologies, which is planned to be built by 2023, is the key milestone on the path to the PHP. This facility is in the phase of conceptual design aimed at providing feasibility studies for a full set of steady state tokamak technologies at a fusion energy gain factor Q ˜ 1, fusion power of ˜40 MW and opportunities for testing a wide range of hybrid technologies with the emphasis on continuous nuclide processing in molten salts. This paper describes the project motivations, its current status and the key issues of the design.

Study of the Adsorption of Fission Products by the Soil of Ezeiza. Report No. 35; ESTUDIO DE LA ADSORCION DE PRODUCTOS DE FISION POR TIERRA DE EZEIZA. INFORME NO. 35

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anghileri, L.J.

1960-01-01

A study was made of the adsorptive properties of Ezeiza soil for fission products using an adsorption column technique and adsorption on suspensions. The tests showed that the upper soil level in the zone of Ezeiza is a good adsorber. For fission products in the presence of U, adsorption was over 75% of the activity, the fixation being dependent on the soil concentration, pH of the solution to be decontaminated, and the contact time. For Sr/sup 90/ the values were close to 99% with concentrations of the order of 25 g of soil/100 cc of solution. For Cs/sup 137/ themore » adsorption is almost complete (99%) with 15 g/cc. (J.S.R.)« less

Correlated prompt fission data in transport simulations

DOE PAGES

Talou, P.; Vogt, R.; Randrup, J.; ...

2018-01-24

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n -n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ raysmore » from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. Here, this review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Lastly, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.« less

Correlated prompt fission data in transport simulations

NASA Astrophysics Data System (ADS)

Talou, P.; Vogt, R.; Randrup, J.; Rising, M. E.; Pozzi, S. A.; Verbeke, J.; Andrews, M. T.; Clarke, S. D.; Jaffke, P.; Jandel, M.; Kawano, T.; Marcath, M. J.; Meierbachtol, K.; Nakae, L.; Rusev, G.; Sood, A.; Stetcu, I.; Walker, C.

2018-01-01

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n - n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ rays from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. This review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Finally, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.

Correlated prompt fission data in transport simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Talou, P.; Vogt, R.; Randrup, J.

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n -n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ raysmore » from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. Here, this review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Lastly, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.« less

The ties that bind: genetic relatedness predicts the fission and fusion of social groups in wild African elephants.

PubMed

Archie, Elizabeth A; Moss, Cynthia J; Alberts, Susan C

2006-03-07

Many social animals live in stable groups. In contrast, African savannah elephants (Loxodonta africana) live in unusually fluid, fission-fusion societies. That is, 'core' social groups are composed of predictable sets of individuals; however, over the course of hours or days, these groups may temporarily divide and reunite, or they may fuse with other social groups to form much larger social units. Here, we test the hypothesis that genetic relatedness predicts patterns of group fission and fusion among wild, female African elephants. Our study of a single Kenyan population spans 236 individuals in 45 core social groups, genotyped at 11 microsatellite and one mitochondrial DNA (mtDNA) locus. We found that genetic relatedness predicted group fission; adult females remained with their first order maternal relatives when core groups fissioned temporarily. Relatedness also predicted temporary fusion between social groups; core groups were more likely to fuse with each other when the oldest females in each group were genetic relatives. Groups that shared mtDNA haplotypes were also significantly more likely to fuse than groups that did not share mtDNA. Our results suggest that associations between core social groups persist for decades after the original maternal kin have died. We discuss these results in the context of kin selection and its possible role in the evolution of elephant sociality.

OPERATIONAL CHARACTERISTICS OF THE ARMOUR FISSION GAS GAMMA FACILITY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Terrell, C.W.; McElroy, W.N.

1958-10-31

As the reactor power level is changed frequently, the radiation levels in the gamma facility fluctuate. Data are presented to show the power dependency of the gamma dose rate and the manner of growth and decay. Additional data show the dependercy of the equilibrium gamma activity on the foel temperature and total system pressure. The final phase of the work is directed toward determining an average gamma energy by attenuation measurements with various thicknesses of several materials. The neutrou flux associated with the gas phase activity is determined by foil measurement. From the measurements of dose rate and average gammamore » energy, calculations to determine the number of curies of gas phase decay gamma activity per watt of reactor power are presented. (auth)« less

Curved Waveguide Based Nuclear Fission for Small, Lightweight Reactors

NASA Technical Reports Server (NTRS)

Coker, Robert; Putnam, Gabriel

2012-01-01

The focus of the presented work is on the creation of a system of grazing incidence, supermirror waveguides for the capture and reuse of fission sourced neutrons. Within research reactors, neutron guides are a well known tool for directing neutrons from the confined and hazardous central core to a more accessible testing or measurement location. Typical neutron guides have rectangular, hollow cross sections, which are crafted as thin, mirrored waveguides plated with metal (commonly nickel). Under glancing angles with incoming neutrons, these waveguides can achieve nearly lossless transport of neutrons to distant instruments. Furthermore, recent developments have created supermirror surfaces which can accommodate neutron grazing angles up to four times as steep as nickel. A completed system will form an enclosing ring or spherical resonator system to a coupled neutron source for the purpose of capturing and reusing free neutrons to sustain and/or accelerate fission. While grazing incidence mirrors are a known method of directing and safely using neutrons, no method has been disclosed for capture and reuse of neutrons or sustainment of fission using a circular waveguide structure. The presented work is in the process of fabricating a functional, highly curved, neutron supermirror using known methods of Ni-Ti layering capable of achieving incident reflection angles up to four times steeper than nickel alone. Parallel work is analytically investigating future geometries, mirror compositions, and sources for enabling sustained fission with applicability to the propulsion and energy goals of NASA and other agencies. Should research into this concept prove feasible, it would lead to development of a high energy density, low mass power source potentially capable of sustaining fission with a fraction of the standard critical mass for a given material and a broadening of feasible materials due to reduced rates of release, absorption, and non-fission for neutrons. This advance could be applied to direct propulsion through guided fission products or as a secondary energy source for high impulse electric propulsion. It would help meet national needs for highly efficient energy sources with limited dependence on fossil fuels or conflict materials, and it would improve the use of low grade fissile materials which would help reduce national stockpiles and waste.

Design and Testing for a New Thermosyphon Irradiation Vehicle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Felde, David K.; Carbajo, Juan J.; McDuffee, Joel Lee

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) requires most materials and all fuel experiments to be placed in a pressure containment vessel to ensure that internal contaminants such as fission products cannot be released into the primary coolant. It also requires that all experiments be capable of withstanding various accident conditions (e.g., loss of coolant) without generating vapor bubbles on the surface of the experiment in the primary coolant. These requirements are intended to artificially increase experiment temperatures by introducing a barrier between the experimental materials and the HFIR coolant, and by reducing heatmore » loads to the HFIR primary coolant, thus ensuring that no boiling can occur. A proposed design for materials irradiation would remove these limitations by providing the required primary containment with an internal cooling flow. This would allow for experiments to be irradiated without concern for coolant contamination (e.g., from cladding failure of advanced fuel pins) or for specimen heat load. This report describes a new materials irradiation experiment design that uses a thermosyphon cooling system to allow experimental materials direct access to a liquid coolant. The new design also increases the range of conditions that can be tested in HFIR. This design will provide a unique capability to validate the performance of current and advanced fuels and materials. Because of limited supporting data for this kind of irradiation vehicle, a test program was initiated to obtain operating data that can be used to (1) qualify the vehicle for operation in HFIR and (2) validate computer models used to perform design- and safety-basis calculations. This report also describes the test facility and experimental data, and it provides a comparison of the experimental data to computer simulations. A total of 51 tests have been completed: four tests with pure steam, 12 tests with argon, and 35 tests with helium. A total of 10 tests were performed at subatmospheric pressure, and four of these were performed with pure steam. One test was conducted at a high power of 92.7 kW, six tests were HFIR startups, and two tests were HFIR loss of offsite power (LOOP). Pressures up to 10 MPa, vapor temperatures up to 583 K (310°C), and heater temperatures above 600 K (327°C) have been reached in these tests. Two computer programs, RELAP5-3D and TRACE, have been used to simulate the tests. The TRACE code has shown good agreement with the test data and has been used to model a variety of tests. This experimental facility has been very useful in demonstrating the viability of this new type of irradiation facility.« less

In-Pile Sub-Miniature Fission Chambers Testing in BR2

NASA Astrophysics Data System (ADS)

Vermeeren, L.; Wéber, M.; Blandin, Ch.; Breaud, S.

2003-06-01

Three innovative sub-miniature fission chambers (SMFC), designed and manufactured at the Nuclear Measurement Systems Laboratory (LSMN) of CEA/Cadarache, were extensively tested in the BR2 research reactor at SCK•CEN, Mol. We present the experimental results for the (thermal) neutron sensitivity, the gamma-induced signal, the signal due to activation, the current picked up by the signal cable, the global current/voltage characteristics and the long term behaviour up to a thermal neutron fluence of 2.7·1021 n/cm2. We also compare the data with results from calculations with our FCD computer code. The onset of the saturation domain is well predicted by FCD; the neutron sensitivities can be accounted for perfectly after a refinement of the FCD model.

Progress In Developing Laser Based Post Irradiation Examination Infrastructure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, James A.; Scott, Clark L.; Benefiel, Brad C.

To be able to understand the performance of reactor fuels and materials, irradiated materials must be characterized effectively and efficiently in a high rad environment. The characterization work must be performed remotely and in an environment hostile to instrumentation. Laser based characterization techniques provide the ability to be remote and robust in a hot-cell environment. Laser based instrumentation also can provide high spatial resolution suitable for scanning and imaging large areas. The INL is currently developing three laser based Post Irradiation Examination (PIE) stations for the Hot Fuel Examination Facility at the INL. These laser based systems will characterize irradiatedmore » materials and fuels. The characterization systems are the following: Laser Shock Laser based ultrasonic C-scan system Gas Assay, Sample, and Recharge system (GASR, up-grade to an existing system). The laser shock technique will characterize material properties and failure loads/mechanisms in various materials such as LWR fuel, plate fuel, and next generation fuel forms, for PIE in high radiation areas. The laser shock-technique induces large amplitude shock waves to mechanically characterize interfaces such as the fuel-clad bond. The shock wave travels as a compression wave through the material to the free (unconfined) back surface and reflects back through the material under test as a rarefaction (tensile) wave. This rarefaction wave is the physical mechanism that produces internal de-lamination failure. As part of the laser shock system, a laser-based ultrasonic C-scan system will be used to detect and characterize debonding caused by the laser shock technique. The laser ultrasonic system will be fully capable of performing classical non-destructive evaluation testing and imaging functions such as microstructure characterization, flaw detection and dimensional metrology in complex components. The purpose of the GASR is to measure the pressure/volume of the plenum of an irradiated fuel element and obtain fission gas samples for analysis. The study of pressure and volume in the plenum of an irradiated fuel element and the analysis of fission gases released from the fuel is important to understanding the performance of reactor fuels and materials. This system may also be used to measure the pressure/volume of other components (such as control blades) and obtain gas samples from these components for analysis. The main function of the laser in this application is to puncture the fuel element to allow the fission gas to escape and if necessary to weld the spot close. The GASR station will have the inherent capability to perform cutting welding and joining functions within a hot-cell.« less

Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Chang, G. S.; Lillo, M. A.

2009-08-01

The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y-Z mini-plate fuel model was developed. The Y-Z model divides each fuel plate into 30 equal intervals in both the Y and Z directions. The MCNP-calculated results and the detailed Y-Z fission power mapping were used to help design the AFIP fuel test assembly to demonstrate that the AFIP test assembly thermal-hydraulic limits will not exceed the ATR safety limits.

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Technical Reports Server (NTRS)

Martin, James; Salvail, Pat

2003-01-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final "wet in". A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/- 1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to less than10(exp -10) std cc/sec helium and vacuum conditioned at 250 C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Astrophysics Data System (ADS)

Martin, James; Salvail, Pat

2004-02-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final ``wet in''. A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/-1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to <10-10 std cc/sec helium and vacuum conditioned at 250 °C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 °C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

The fiftieth anniversary of the first public announcement of the successful test of fission: Proceedings

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This report contains comments and discussions on the history of fission. The following people comments and lectures are discussed in this report: Remarks and introduction of Maxine F. Singer; president`s message, Maxine F. Singer; introduction of Stephen Joel Trachtenberg; President`s message, Stephen Joel Trachtenberg; introduction of Frederick Seitz; lecture: ``Nuclear Science: Promises and Perceptions, `` Frederick Seitz; introduction of K. Alex Mueller; lecture: ``High Temperature Ferroelectricity and Superconductivity,`` introduction of Edward Teller; and lecture: ``Toward a More Secure World,`` Edward Teller.

Radiochemistry and the Study of Fission

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rundberg, Robert S.

These are slides from a lecture given at UC Berkeley. Radiochemistry has been used to study fission since its discovery. Radiochemical methods are used to determine cumulative mass yields. These measurements have led to the two-mode fission hypothesis to model the neutron energy dependence of fission product yields. Fission product yields can be used for the nuclear forensics of nuclear explosions. The mass yield curve depends on both the fuel and the neutron spectrum of a device. Recent studies have shown that the nuclear structure of the compound nucleus can affect the mass yield distribution. The following topics are covered:more » In the beginning: the discovery of fission; forensics using fission products: what can be learned from fission products, definitions of R-values and Q-values, fission bases, K-factors and fission chambers, limitations; the neutron energy dependence of the mass yield distribution (the two mode fission hypothesis); the influence of nuclear structure on the mass yield distribution. In summary: Radiochemistry has been used to study fission since its discovery. Radiochemical measurement of fission product yields have provided the highest precision data for developing fission models and for nuclear forensics. The two-mode fission hypothesis provides a description of the neutron energy dependence of the mass yield curve. However, data is still rather sparse and more work is needed near second and third chance fission. Radiochemical measurements have provided evidence for the importance of nuclear states in the compound nucleus in predicting the mass yield curve in the resonance region.« less

Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collin, Blaise P.; Petti, David A.; Demkowicz, Paul A.

Safety tests were conducted on fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800 °C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during 15 of these safety tests. Comparisons between PARFUME predictions and post-irradiation examination results of the safety tests were conducted on two types of AGR-1 compacts: compactsmore » containing only intact particles and compacts containing one or more particles whose SiC layers failed during safety testing. In both cases, PARFUME globally over-predicted the experimental release fractions by several orders of magnitude: more than three (intact) and two (failed SiC) orders of magnitude for silver, more than three and up to two orders of magnitude for strontium, and up to two and more than one orders of magnitude for krypton. The release of cesium from intact particles was also largely over-predicted (by up to five orders of magnitude) but its release from particles with failed SiC was only over-predicted by a factor of about 3. These over-predictions can be largely attributed to an over-estimation of the diffusivities used in the modeling of fission product transport in TRISO-coated particles. The integral release nature of the data makes it difficult to estimate the individual over-estimations in the kernel or each coating layer. Nevertheless, a tentative assessment of correction factors to these diffusivities was performed to enable a better match between the modeling predictions and the safety testing results. The method could only be successfully applied to silver and cesium. In the case of strontium, correction factors could not be assessed because potential release during the safety tests could not be distinguished from matrix content released during irradiation. Furthermore, in the case of krypton, all the coating layers are partly retentive and the available data did not allow the level of retention in individual layers to be determined, hence preventing derivation of any correction factors.« less

Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

DOE PAGES

Collin, Blaise P.; Petti, David A.; Demkowicz, Paul A.; ...

2016-04-07

Safety tests were conducted on fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800 °C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during 15 of these safety tests. Comparisons between PARFUME predictions and post-irradiation examination results of the safety tests were conducted on two types of AGR-1 compacts: compactsmore » containing only intact particles and compacts containing one or more particles whose SiC layers failed during safety testing. In both cases, PARFUME globally over-predicted the experimental release fractions by several orders of magnitude: more than three (intact) and two (failed SiC) orders of magnitude for silver, more than three and up to two orders of magnitude for strontium, and up to two and more than one orders of magnitude for krypton. The release of cesium from intact particles was also largely over-predicted (by up to five orders of magnitude) but its release from particles with failed SiC was only over-predicted by a factor of about 3. These over-predictions can be largely attributed to an over-estimation of the diffusivities used in the modeling of fission product transport in TRISO-coated particles. The integral release nature of the data makes it difficult to estimate the individual over-estimations in the kernel or each coating layer. Nevertheless, a tentative assessment of correction factors to these diffusivities was performed to enable a better match between the modeling predictions and the safety testing results. The method could only be successfully applied to silver and cesium. In the case of strontium, correction factors could not be assessed because potential release during the safety tests could not be distinguished from matrix content released during irradiation. Furthermore, in the case of krypton, all the coating layers are partly retentive and the available data did not allow the level of retention in individual layers to be determined, hence preventing derivation of any correction factors.« less

Phase 1 Methyl Iodide Deep-Bed Adsorption Tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soelberg, Nick; Watson, Tony

2014-08-22

Nuclear fission results in the production of fission products (FPs) and activation products including iodine-129, which could evolve into used fuel reprocessing facility off-gas systems, and could require off-gas control to limit air emissions to levels within acceptable emission limits. Research, demonstrations, and some reprocessing plant experience have indicated that diatomic iodine can be captured with efficiencies high enough to meet regulatory requirements. Research on the capture of organic iodides has also been performed, but to a lesser extent [Jubin 2012b]. Several questions remain open regarding the capture of iodine bound in organic compounds. Deep-bed methyl iodide adsorption testing hasmore » progressed according to a multi-laboratory methyl iodide adsorption test plan. This report summarizes the first phase of methyl iodide adsorption work performed according to this test plan using the deep-bed iodine adsorption test system at the Idaho National Laboratory (INL), performed during Fiscal Year (FY) 2013 and early FY-2014. Testing has been performed to address questions posed in the test plan, and followed the testing outline in the test plan. Tests established detection limits, developed procedures for sample analysis with minimal analytical interferences, and confirmed earlier results that show that the methyl iodide reacts when in contact with the AgZ sorbent, and not significantly in the gas flow upstream of the sorbent. The reaction(s) enable separation of the iodine from the organic moiety, so that the iodine can chemisorb onto the sorbent. The organic moiety can form other compounds, some of which are organic compounds that are detected and can be tentatively identified using GC-FID and GCMS. Test results also show that other gas constituents (NOx and/or H2O) can affect the methyl iodide reactions. With NOx and H2O present in the gas stream, the majority of uncaptured iodine exiting iodine-laden sorbent beds is in the form of I2 or HI, species that are soluble in NaOH scrubbing solution for iodine analysis. But when NOx and H2O are not present, then the majority of the uncaptured iodine exiting iodine-laden sorbent is in the form of methyl iodide. Methyl iodide adsorption efficiencies have been high enough so that initial DFs exceed 1,000 to 10,000. The methyl iodide mass transfer zone depths are estimated at 4-8 inches, possibly deeper than mass transfer zone depths estimated for I2 adsorption on AgZ. Additional deep-bed testing and analyses are recommended to (a) expand the data base for methyl iodide adsorption under various conditions specified in the methyl iodide test plan, and (b) provide more data for evaluating organic iodide reactions and reaction byproducts for different potential adsorption conditions.« less

Development of a “Fission-proxy” Method for the Measurement of 14-MeV Neutron Fission Yields at CAMS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gharibyan, Narek

2016-10-25

Relative fission yield measurements were made for 50 fission products from 25.6±0.5 MeV alpha-induced fission of Th-232. Quantitative comparison of these experimentally measured fission yields with the evaluated fission yields from 14-MeV neutron-induced fission of U-235 demonstrates the feasibility of the proposed fission-proxy method. This new technique, based on the Bohr-independence hypothesis, permits the measurement of fission yields from an alternate reaction pathway (Th-232 + 25.6 MeV α → U-236* vs. U-235 + 14-MeV n → U-236*) given that the fission process associated with the same compound nucleus is independent of its formation. Other suitable systems that can potentially bemore » investigated in this manner include (but are not limited to) Pu-239 and U-237.« less

Dynamic high energy density plasma environments at the National Ignition Facility for nuclear science research

NASA Astrophysics Data System (ADS)

Cerjan, Ch J.; Bernstein, L.; Berzak Hopkins, L.; Bionta, R. M.; Bleuel, D. L.; Caggiano, J. A.; Cassata, W. S.; Brune, C. R.; Frenje, J.; Gatu-Johnson, M.; Gharibyan, N.; Grim, G.; Hagmann, Chr; Hamza, A.; Hatarik, R.; Hartouni, E. P.; Henry, E. A.; Herrmann, H.; Izumi, N.; Kalantar, D. H.; Khater, H. Y.; Kim, Y.; Kritcher, A.; Litvinov, Yu A.; Merrill, F.; Moody, K.; Neumayer, P.; Ratkiewicz, A.; Rinderknecht, H. G.; Sayre, D.; Shaughnessy, D.; Spears, B.; Stoeffl, W.; Tommasini, R.; Yeamans, Ch; Velsko, C.; Wiescher, M.; Couder, M.; Zylstra, A.; Schneider, D.

2018-03-01

The generation of dynamic high energy density plasmas in the pico- to nano-second time domain at high-energy laser facilities affords unprecedented nuclear science research possibilities. At the National Ignition Facility (NIF), the primary goal of inertial confinement fusion research has led to the synergistic development of a unique high brightness neutron source, sophisticated nuclear diagnostic instrumentation, and versatile experimental platforms. These novel experimental capabilities provide a new path to investigate nuclear processes and structural effects in the time, mass and energy density domains relevant to astrophysical phenomena in a unique terrestrial environment. Some immediate applications include neutron capture cross-section evaluation, fission fragment production, and ion energy loss measurement in electron-degenerate plasmas. More generally, the NIF conditions provide a singular environment to investigate the interplay of atomic and nuclear processes such as plasma screening effects upon thermonuclear reactivity. Achieving enhanced understanding of many of these effects will also significantly advance fusion energy research and challenge existing theoretical models.

Investigation of materials for fusion power reactors

NASA Astrophysics Data System (ADS)

Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

2014-06-01

The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

A target development program for beamhole spallation neutron sources in the megawatt range

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bauer, G.S.; Atchison, F.

1995-10-01

Spallation sources as an alternative to fission neutron sources have been operating successfully up to 160 kW of beam power. With the next generation of these facilities aiming at the medium power range between 0.5 and 5 MW, loads on the targets will be high enough to make present experience of little relevance. With the 0.6 MW continuous facility SINQ under construction, and a 5 MW pulsed facility (ESS) under study in Europe, a research and development program is about to be started which aimes at assessing the limits of stationary and moving solid targets and the feasibility and potentialmore » benefits of flowing liquid metal targets. Apart from theoretical work and examination of existing irradiated material, including used targets from ISIS, it is intended to take advantage of the SINQ solid rod target design to improve the relevant data base by building the target in such a way that individual rods can be equipped as irradiation capsules.« less

Synthesis of Actinide Materials for the Study of Basic Actinide Science and Rapid Separation of Fission Products

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dorhout, Jacquelyn Marie

This dissertation covers several distinct projects relating to the fields of nuclear forensics and basic actinide science. Post-detonation nuclear forensics, in particular, the study of fission products resulting from a nuclear device to determine device attributes and information, often depends on the comparison of fission products to a library of known ratios. The expansion of this library is imperative as technology advances. Rapid separation of fission products from a target material, without the need to dissolve the target, is an important technique to develop to improve the library and provide a means to develop samples and standards for testing separations.more » Several materials were studied as a proof-of-concept that fission products can be extracted from a solid target, including microparticulate (< 10 μm diameter) dUO 2, porous metal organic frameworks (MOFs) synthesized from depleted uranium (dU), and other organicbased frameworks containing dU. The targets were irradiated with fast neutrons from one of two different neutron sources, contacted with dilute acids to facilitate the separation of fission products, and analyzed via gamma spectroscopy for separation yields. The results indicate that smaller particle sizes of dUO 2 in contact with the secondary matrix KBr yield higher separation yields than particles without a secondary matrix. It was also discovered that using 0.1 M HNO 3 as a contact acid leads to the dissolution of the target material. Lower concentrations of acid were used for future experiments. In the case of the MOFs, a larger pore size in the framework leads to higher separation yields when contacted with 0.01 M HNO 3. Different types of frameworks also yield different results.« less

ARTIST: An International Project Investigating Aerosol Retention in a Ruptured Steam Generator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guentay, S.; Dehbi, A.; Suckow, D.

2002-07-01

Steam generator tube ruptures (SGTR) with a concurrent stuck open safety relief valve are counted among the risk dominant accident sequences because of the potential for radioactive products to bypass the containment. Owing to the absence of relevant empirical data and the complexity of the geometry and controlling processes, the aerosol removal in the steam generator (SG) tubes and in the secondary side is not well understood. Therefore, little or no credit is usually taken for aerosol retention due to natural processes in the various components of a SG. To help reduce the uncertainties associated with fission product release followingmore » an SGTR sequence, the Paul Scherrer Institut has initiated an international experimental project to be performed in the ARTIST (AeRosol Trapping In a Steam generaTor) facility in the time period from 2002 to 2007. The ARTIST test section is a scaled model of a real SG, and is comprised of a 264-tube bundle with a maximum height of 3.8 m, as well as one full-size droplet separator and one full-size steam dryer. The ARTIST facility is capable of producing soluble and insoluble aerosols and entrain them at sonic gas flow rates (up to 0.25 kg/s, thus matching comparable values predicted by the codes. In addition, aerosols can be generated at prototypical concentrations (up to 5 g/m{sup 3}) and sizes (0.2-5 mm AMMD). State of the art instrumentation is used (Low-pressure impactors, photometers, on-line particle sizer, online droplet sizer, etc.). The ARTIST project will simulate the flow and retention of aerosol-borne fission products in the SG, and provide a unique database to support safety assessments and analytical models. The project is foreseen in seven phases: 1) Aerosol retention in the tube under dry secondary side conditions, 2) Aerosol retention in the near field close to break under dry conditions, 3) Aerosol retention in the bundle far field under dry conditions, 4) Aerosol retention in the separator and dryer under dry conditions, 5) Aerosol retention in the bundle section under wet conditions, 6) Droplet retention in separator and dryer sections and 7) Integral tests to examine overall retention in the SG unit. The project will investigate phenomena at the separate effect and integral levels, and will also address selected accident management (AM) issues. The kick-off experiments are scheduled for the first half of 2002, and some early results will be summarized at the meeting. (authors)« less

Computer program FPIP-REV calculates fission product inventory for U-235 fission

NASA Technical Reports Server (NTRS)

Brown, W. S.; Call, D. W.

1967-01-01

Computer program calculates fission product inventories and source strengths associated with the operation of U-235 fueled nuclear power reactor. It utilizes a fission-product nuclide library of 254 nuclides, and calculates the time dependent behavior of the fission product nuclides formed by fissioning of U-235.

Facile manipulation of protein localization in fission yeast through binding of GFP-binding protein to GFP.

PubMed

Chen, Ying-Hui; Wang, Gao-Yuan; Hao, Hao-Chao; Chao, Chun-Jiang; Wang, Yamei; Jin, Quan-Wen

2017-03-01

GFP-binding protein (or GBP) has been recently developed in various systems and organisms as an efficient tool to purify GFP-fusion proteins. Due to the high affinity between GBP and GFP or GFP variants, this GBP-based approach is also ideally suited to alter the localization of functional proteins in live cells. In order to facilitate the wide use of the GBP-targeting approach in the fission yeast Schizosaccharomyces pombe , we developed a set of pFA6a-, pJK148- and pUC119-based vectors containing GBP- or GBP-mCherry-coding sequences and variants of inducible nmt1 or constitutive adh1 promoters that result in different levels of expression. The GBP or GBP-mCherry fragments can serve as cassettes for N- or C-terminal genomic tagging of genes of interest. We illustrated the application of these vectors in the construction of yeast strains with Dma1 or Cdc7 tagged with GBP-mCherry and efficient targeting of Dma1- or Cdc7-GBP-mCherry to the spindle pole body by Sid4-GFP. This series of vectors should help to facilitate the application of the GBP-targeting approach in manipulating protein localization and the analysis of gene function in fission yeast, at the level of single genes, as well as at a systematic scale. © 2017. Published by The Company of Biologists Ltd.

International challenge to predict the impact of radioxenon releases from medical isotope production on a comprehensive nuclear test ban treaty sampling station.

PubMed

Eslinger, Paul W; Bowyer, Ted W; Achim, Pascal; Chai, Tianfeng; Deconninck, Benoit; Freeman, Katie; Generoso, Sylvia; Hayes, Philip; Heidmann, Verena; Hoffman, Ian; Kijima, Yuichi; Krysta, Monika; Malo, Alain; Maurer, Christian; Ngan, Fantine; Robins, Peter; Ross, J Ole; Saunier, Olivier; Schlosser, Clemens; Schöppner, Michael; Schrom, Brian T; Seibert, Petra; Stein, Ariel F; Ungar, Kurt; Yi, Jing

2016-06-01

The International Monitoring System (IMS) is part of the verification regime for the Comprehensive Nuclear-Test-Ban-Treaty Organization (CTBTO). At entry-into-force, half of the 80 radionuclide stations will be able to measure concentrations of several radioactive xenon isotopes produced in nuclear explosions, and then the full network may be populated with xenon monitoring afterward. An understanding of natural and man-made radionuclide backgrounds can be used in accordance with the provisions of the treaty (such as event screening criteria in Annex 2 to the Protocol of the Treaty) for the effective implementation of the verification regime. Fission-based production of (99)Mo for medical purposes also generates nuisance radioxenon isotopes that are usually vented to the atmosphere. One of the ways to account for the effect emissions from medical isotope production has on radionuclide samples from the IMS is to use stack monitoring data, if they are available, and atmospheric transport modeling. Recently, individuals from seven nations participated in a challenge exercise that used atmospheric transport modeling to predict the time-history of (133)Xe concentration measurements at the IMS radionuclide station in Germany using stack monitoring data from a medical isotope production facility in Belgium. Participants received only stack monitoring data and used the atmospheric transport model and meteorological data of their choice. Some of the models predicted the highest measured concentrations quite well. A model comparison rank and ensemble analysis suggests that combining multiple models may provide more accurate predicted concentrations than any single model. None of the submissions based only on the stack monitoring data predicted the small measured concentrations very well. Modeling of sources by other nuclear facilities with smaller releases than medical isotope production facilities may be important in understanding how to discriminate those releases from releases from a nuclear explosion. Published by Elsevier Ltd.