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Sample records for gas fission products

  1. New Results on Helium and Tritium Gas Production From Ternary Fission

    NASA Astrophysics Data System (ADS)

    Serot, O.; Wagemans, C.; Heyse, J.

    2005-05-01

    Ternary fission constitutes an important source of helium and tritium gas production in nuclear reactors and in used fuel elements. Data related to this production are therefore requested by nuclear industry. In the present paper, we report results from measurements of the 4He and 3H emission probabilities (denoted LRA/B and t/B, respectively). These measurements concern both thermal neutron-induced fission reactions as well as spontaneous fission decays. For spontaneous fission, data are reported for nuclides ranging from 238Pu up to 252Cf. For thermal neutron-induced fission, results cover target nuclei between 229Th and 251Cf. Based on these and other results, semi-empirical relations are proposed. These correlations are only valid if spontaneous fission data and neutron-induced fission data are considered separately, which shows the impact of the fissioning nucleus-excitation energy on the ternary particle-emission process. In this way, t/B and LRA/B values could be evaluated for fissioning systems not investigated so far. These results could be used for the ternary fission-yield evaluation of the JEFF3.1 library.

  2. Fission gas detection system

    DOEpatents

    Colburn, Richard P.

    1985-01-01

    A device for collecting fission gas released by a failed fuel rod which device uses a filter to pass coolant but which filter blocks fission gas bubbles which cannot pass through the filter due to the surface tension of the bubble.

  3. Selective Trapping of Volatile Fission Products with an Off-Gas Treatment System

    SciTech Connect

    B.R. Westphal; J.J. Park; J.M. Shin; G.I. Park; K.J. Bateman; D.L. Wahlquist

    2008-07-01

    A head-end processing step, termed DEOX for its emphasis on decladding via oxidation, is being developed for the treatment of spent oxide fuel by pyroprocessing techniques. The head-end step employs high temperatures to oxidize UO2 to U3O8 resulting in the separation of fuel from cladding and the removal of volatile fission products. Development of the head-end step is being performed in collaboration with the Korean Atomic Energy Research Institute (KAERI) through an International Nuclear Energy Research Initiative. Following the initial experimentation for the removal of volatile fission products, an off-gas treatment system was designed in conjunction with KAERI to collect specific fission gases. The primary volatile species targeted for trapping were iodine, technetium, and cesium. Each species is intended to be collected in distinct zones of the off-gas system and within those zones, on individual filters. Separation of the volatile off-gases is achieved thermally as well as chemically given the composition of the filter media. A description of the filter media and a basis for its selection will be given along with the collection mechanisms and design considerations. In addition, results from testing with the off-gas treatment system will be presented.

  4. Transport of fission products with a helium gas-jet at TRIGA-SPEC

    NASA Astrophysics Data System (ADS)

    Eibach, M.; Beyer, T.; Blaum, K.; Block, M.; Eberhardt, K.; Herfurth, F.; Geppert, C.; Ketelaer, J.; Ketter, J.; Krämer, J.; Krieger, A.; Knuth, K.; Nagy, Sz.; Nörtershäuser, W.; Smorra, C.

    2010-02-01

    A helium gas-jet system for the transport of fission products from the research reactor TRIGA Mainz has been developed, characterized and tested within the TRIGA-SPEC experiment. For the first time at TRIGA Mainz carbon aerosol particles have been used for the transport of radionuclides from a target chamber with high efficiency. The radionuclides have been identified by means of γ-spectroscopy. Transport time, efficiency as well as the absolute number of transported radionuclides for several species have been determined. The design and the characterization of the gas-jet system are described and discussed.

  5. Fission Product Monitoring and Release Data for the Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John B. Walter; Jason M. Harp; Mark W. Drigert; Edward L. Reber

    2010-10-01

    The AGR-1 experiment is a fueled multiple-capsule irradiation experiment that was irradiated in the Advanced Test Reactor (ATR) from December 26, 2006 until November 6, 2009 in support of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Fuel Development and Qualification program. An important measure of the fuel performance is the quantification of the fission product releases over the duration of the experiment. To provide this data for the inert fission gasses(Kr and Xe), a fission product monitoring system (FPMS) was developed and implemented to monitor the individual capsule effluents for the radioactive species. The FPMS continuously measured the concentrations of various krypton and xenon isotopes in the sweep gas from each AGR-1 capsule to provide an indicator of fuel irradiation performance. Spectrometer systems quantified the concentrations of Kr-85m, Kr-87, Kr-88, Kr-89, Kr-90, Xe-131m, Xe-133, Xe 135, Xe 135m, Xe-137, Xe-138, and Xe-139 accumulated over repeated eight hour counting intervals.-. To determine initial fuel quality and fuel performance, release activity for each isotope of interest was derived from FPMS measurements and paired with a calculation of the corresponding isotopic production or birthrate. The release activities and birthrates were combined to determine Release-to-Birth ratios for the selected nuclides. R/B values provide indicators of initial fuel quality and fuel performance during irradiation. This paper presents a brief summary of the FPMS, the release to birth ratio data for the AGR-1 experiment and preliminary comparisons of AGR-1 experimental fuels data to fission gas release models.

  6. FIssion Product Prompt γ-ray spectrometer: Development of an instrumented gas-filled magnetic spectrometer at the ILL

    NASA Astrophysics Data System (ADS)

    Blanc, A.; Chebboubi, A.; Faust, H.; Jentschel, M.; Kessedjian, G.; Köster, U.; Materna, T.; Panebianco, S.; Sage, C.; Urban, W.

    2013-12-01

    Accurate thermal neutron-induced fission data are important for applications in reactor physics as well as for fundamental nuclear physics. FIPPS is the new FIssion Product Prompt γ-ray Spectrometer being developed at the Institut Laue Langevin for neutron-induced fission studies. FIPPS is based on the combination of a large Germanium detector array surrounding a fission target, a Time-Of-Flight detector and a Gas-Filled Magnet (GFM) to identify mass, nuclear charge and kinetic energy of one of the fission fragments. The GFM will be instrumented with a Time-Projection Chamber (TPC) for individual 3D tracking of the fragments. A conceptual design study of the new spectrometer is presented.

  7. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John b. Walter

    2010-10-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  8. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K Hartwell; John B. Walter

    2008-09-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  9. Gas emission from the UO2 samples, containing fission products and burnable absorber

    NASA Astrophysics Data System (ADS)

    Kopytin, V. P.; Baranov, V. G.; Burlakova, M. A.; Tenishev, A. V.; Kuzmin, R. S.; Pokrovskiy, S. A.; Mikhalchik, V. V.

    2016-04-01

    The process gas released from the fuel pellets of uranium fuel during fuel burn-up reduces the thermal conductivity of the rod-shell gap, enhances hydrogen embrittlement of the cladding material, causes it's carbonization, as well as transport processes in the fuel. In this study a technique of investigating the thermal desorption of gases from the UO2 fuel material were perfected in the temperature range 300-2000 K for uniform sample heating rate of 15 K/min in vacuum. The characteristic kinetic dependences are acquired for the gas emission from UO2 samples, containing simulators of fission products (SFP) and the burnable neutron absorber (BNA). Depending on the amount of SFP and BNA contained in the sample thermal desorption gas spectra (TDGS) vary. The composition of emitted gas varies, as well as the number of peaks in the TDGS and the peaks shift to higher temperatures. This indicates that introduction of SFPs and BNA alters the sample material structure and cause the creation of so- called traps which have different bonding energies to the gases. The traps can be a grid of dislocations, voids, and contained in the UO2 matrix SFP and BNA. Similar processes will occur in the fuel pellets in the real conditions of the Nuclear Power Plant as well.

  10. Fission-product retention in HTGR fuels

    SciTech Connect

    Homan, F.J.; Kania, M.J.; Tiegs, T.N.

    1982-01-01

    Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed.

  11. Production of fissioning uranium plasma to approximate gas-core reactor conditions

    NASA Technical Reports Server (NTRS)

    Lee, J. H.; Mcfarland, D. R.; Hohl, F.; Kim, K. H.

    1974-01-01

    The intense burst of neutrons from the d-d reaction in a plasma-focus apparatus is exploited to produce a fissioning uranium plasma. The plasma-focus apparatus consists of a pair of coaxial electrodes and is energized by a 25 kJ capacitor bank. A 15-g rod of 93% enriched U-235 is placed in the end of the center electrode where an intense electron beam impinges during the plasma-focus formation. The resulting uranium plasma is heated to about 5 eV. Fission reactions are induced in the uranium plasma by neutrons from the d-d reaction which were moderated by the polyethylene walls. The fission yield is determined by evaluating the gamma peaks of I-134, Cs-138, and other fission products, and it is found that more than 1,000,000 fissions are induced in the uranium for each focus formation, with at least 1% of these occurring in the uranium plasma.

  12. Mass spectrometry studies of fission product behavior: 2, Gas phase species

    SciTech Connect

    Blackburn, P.E.; Johnson, C.E.

    1987-01-01

    Revaporization of fission products from reactor system surfaces has become a complicating factor in source term definition. Critical to this phenomena is understanding the nature and behavior of the vapor phase species. This study characterizes the stability of the CsI . CsOH vapor phase complex. Vapor pressures were measured with a mass spectrometer. Thermodynamic data were obtained for CsOH(g), Cs/sub 2/(OH)/sub 2/(g), CsI(g), Cs/sub 2/I/sub 2/(g) and CsI . CsOH(g). Activity coefficients were derived for the CsI-CsOH system. The relative ionization cross section of CsOH is about ten times the cross section of CsI(g). CsI . CsOH fragments to Cs/sub 2/OH/sup +/ and an iodine atom. 17 refs., 4 figs., 6 tabs.

  13. TREATMENT OF FISSION PRODUCT WASTE

    DOEpatents

    Huff, J.B.

    1959-07-28

    A pyrogenic method of separating nuclear reactor waste solutions containing aluminum and fission products as buring petroleum coke in an underground retort, collecting the easily volatile gases resulting as the first fraction, he uminum chloride as the second fraction, permitting the coke bed to cool and ll contain all the longest lived radioactive fission products in greatly reduced volume.

  14. Gas-leaking fuel elements number and fission gas product coolant volumetric activities assessment in the VVER-440 nuclear power plant

    NASA Astrophysics Data System (ADS)

    Szuta, Marcin

    1992-07-01

    In a nuclear power plant it is required to monitor continuously the number of gas-leaking fuel elements and the contamination level of the primary coolant by fission gas products. It is proposed to use the radiation monitoring system equipped with the computer technics provided with the suitable program package for fulfilment this requirements. The input data to start up the program consists of the 88Kr volumetric activity measured by the radiation monitoring system and three actual technological parameters: coolant temperature at inlet, thermal power and coolant flow rate.

  15. Correlation of recent fission product release data

    SciTech Connect

    Kress, T.S.; Lorenz, R.A.; Nakamura, T.; Osborne, M.F.

    1989-01-01

    For the calculation of source terms associated with severe accidents, it is necessary to model the release of fission products from fuel as it heats and melts. Perhaps the most definitive model for fission product release is that of the FASTGRASS computer code developed at Argonne National Laboratory. There is persuasive evidence that these processes, as well as additional chemical and gas phase mass transport processes, are important in the release of fission products from fuel. Nevertheless, it has been found convenient to have simplified fission product release correlations that may not be as definitive as models like FASTGRASS but which attempt in some simple way to capture the essence of the mechanisms. One of the most widely used such correlation is called CORSOR-M which is the present fission product/aerosol release model used in the NRC Source Term Code Package. CORSOR has been criticized as having too much uncertainty in the calculated releases and as not accurately reproducing some experimental data. It is currently believed that these discrepancies between CORSOR and the more recent data have resulted because of the better time resolution of the more recent data compared to the data base that went into the CORSOR correlation. This document discusses a simple correlational model for use in connection with NUREG risk uncertainty exercises. 8 refs., 4 figs., 1 tab.

  16. [Fission product yields of 60 fissioning reactions]. Final report

    SciTech Connect

    Rider, B.F.

    1995-05-01

    In keeping with the statement of work, I have examined the fission product yields of 60 fissioning reactions. In co-authorship with the UTR (University Technical Representative) Talmadge R. England ``Evaluation and Compilation of Fission Product Yields 1993,`` LA-UR-94-3106(ENDF-349) October, (1994) was published. This is an evaluated set of fission product Yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set.

  17. PRODUCING ENERGY AND RADIOACTIVE FISSION PRODUCTS

    DOEpatents

    Segre, E.; Kennedy, J.W.; Seaborg, G.T.

    1959-10-13

    This patent broadly discloses the production of plutonium by the neutron bombardment of uranium to produce neptunium which decays to plutonium, and the fissionability of plutonium by neutrons, both fast and thermal, to produce energy and fission products.

  18. The SPIDER fission fragment spectrometer for fission product yield measurements

    SciTech Connect

    Meierbachtol, K.; Tovesson, F.; Shields, D.; Arnold, C.; Blakeley, R.; Bredeweg, T.; Devlin, M.; Hecht, A. A.; Heffern, L. E.; Jorgenson, J.; Laptev, A.; Mader, D.; O׳Donnell, J. M.; Sierk, A.; White, M.

    2015-04-01

    The SPectrometer for Ion DEtermination in fission Research (SPIDER) developed 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 utilized, with the goal of measuring the mass of the fission products with an average resolution of 1 atomic mass unit (amu). The SPIDER instrument, consisting of detector components for time-of-flight, trajectory, and energy measurements, assembled and tested using 229Th and 252Cf radioactive decay sources. For commissioning, the fully assembled system measured fission products from spontaneous fission of 252Cf. Finally, 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). These mass yield results measured from 252Cf spontaneous fission products are reported from an E–v measurement.

  19. The SPIDER fission fragment spectrometer for fission product yield measurements

    DOE PAGES

    Meierbachtol, K.; Tovesson, F.; Shields, D.; Arnold, C.; Blakeley, R.; Bredeweg, T.; Devlin, M.; Hecht, A. A.; Heffern, L. E.; Jorgenson, J.; et al

    2015-04-01

    The SPectrometer for Ion DEtermination in fission Research (SPIDER) developed 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 utilized, with the goal of measuring the mass of the fission products with an average resolution of 1 atomic mass unit (amu). The SPIDER instrument, consisting of detector components for time-of-flight, trajectory, and energy measurements, assembled and tested using 229Th and 252Cf radioactive decay sources. For commissioning, the fully assembled system measured fission products from spontaneous fission of 252Cf. Finally,more » 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). These mass yield results measured from 252Cf spontaneous fission products are reported from an E–v measurement.« less

  20. 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.

  1. Antiproton Powered Gas Core Fission Rocket

    NASA Astrophysics Data System (ADS)

    Kammash, T.

    Extensive research in recent years has demonstrated that “at rest” annihilation of antiprotons in the uranium isotope U238 leads to fission at nearly 100% efficiency. The resulting highly-ionizing, energetic fission fragments can heat a suitable medium to very high temperatures, making such a process particularly suitable for space propulsion applications. Such an ionized medium, which would serve as a propellant, can be confined by a magnetic field during the heating process, and subsequently ejected through a magnetic nozzle to generate thrust. The gasdynamic mirror (GDM) magnetic configuration is especially suited for this application since the underlying confinement principle is that the plasma be of such density and temperature as to make the ion-ion collision mean free path shorter than the plasma length. Under these conditions the plasma behaves like a fluid, and its escape from the system is analogous to the flow of a gas into vacuum from a vessel with a hole. For the system we propose we envisage radially injecting atomic or U238 plasma beam at a pre-determined position and axially pulsing an antiproton beam which upon interaction with the uranium target gives rise to near isotropic ejection of fission fragments with a total mass of 212 amu and total energy of about 160 MeV. These particles, along with the annihilation products (i.e. pions and muons) will heat the background U238 gas - inserted into the chamber just prior to the release of the antiproton - to one keV temperature. Preliminary analysis reveals that such a propulsion system can produce a specific impulse of about 3000 seconds at a thrust of about 50 kN. When applied to a round trip Mars mission, we find that such a journey can be accomplished in about 142 days with 2 days of thrusting and requiring only one gram of antiprotons to achieve it.

  2. Antiproton Powered Gas Core Fission Rocket

    NASA Astrophysics Data System (ADS)

    Kammash, Terry

    2005-02-01

    Extensive research in recent years has demonstrated that "at rest" annihilation of antiprotons in the uranium isotope U238 leads to fission at nearly 100% efficiency. The resulting highly-ionizing, energetic fission fragments can heat a suitable medium to very high temperatures, making such a process particularly suitable for space propulsion applications. Such an ionized medium, which would serve as a propellant, can be confined by a magnetic field during the heating process, and subsequently ejected through a magnetic nozzle to generate thrust. The gasdynamic mirror (GDM) magnetic configuration is especially suited for this application since the underlying confinement principle is that the plasma be of such density and temperature as to make the ion-ion collision mean free path shorter than the plasma length. Under these conditions the plasma behaves like a fluid, and its escape from the system is analogous to the flow of a gas into vacuum from a vessel with a hole. For the system we propose we envisage radially injecting atomic or U238 plasma beam at a pre-determined position and axially pulsing an antiproton beam which upon interaction with the uranium target gives rise to near isotropic ejection of fission fragments with a total mass of 212 amu and total energy of about 160 MeV. These particles, along with the annihilation products (i.e. pions and muons) will heat the background U238 gas — inserted into the chamber just prior to the release of the antiproton — to one keV temperature. Preliminary analysis reveals that such a propulsion system can produce a specific impulse of about 3000 seconds at a thrust of about 50 kN. When applied to a round trip Mars mission, we find that such a journey can be accomplished in about 142 days with 2 days of thrusting and requiring only one gram of antiprotons to achieve it.

  3. Antiproton Powered Gas Core Fission Rocket

    SciTech Connect

    Kammash, Terry

    2005-02-06

    Extensive research in recent years has demonstrated that 'at rest' annihilation of antiprotons in the uranium isotope U238 leads to fission at nearly 100% efficiency. The resulting highly-ionizing, energetic fission fragments can heat a suitable medium to very high temperatures, making such a process particularly suitable for space propulsion applications. Such an ionized medium, which would serve as a propellant, can be confined by a magnetic field during the heating process, and subsequently ejected through a magnetic nozzle to generate thrust. The gasdynamic mirror (GDM) magnetic configuration is especially suited for this application since the underlying confinement principle is that the plasma be of such density and temperature as to make the ion-ion collision mean free path shorter than the plasma length. Under these conditions the plasma behaves like a fluid, and its escape from the system is analogous to the flow of a gas into vacuum from a vessel with a hole. For the system we propose we envisage radially injecting atomic or U238 plasma beam at a pre-determined position and axially pulsing an antiproton beam which upon interaction with the uranium target gives rise to near isotropic ejection of fission fragments with a total mass of 212 amu and total energy of about 160 MeV. These particles, along with the annihilation products (i.e. pions and muons) will heat the background U238 gas - inserted into the chamber just prior to the release of the antiproton - to one keV temperature. Preliminary analysis reveals that such a propulsion system can produce a specific impulse of about 3000 seconds at a thrust of about 50 kN. When applied to a round trip Mars mission, we find that such a journey can be accomplished in about 142 days with 2 days of thrusting and requiring only one gram of antiprotons to achieve it.

  4. RECOVERY OF ALUMINUM FROM FISSION PRODUCTS

    DOEpatents

    Blanco, R.E.; Higgins, I.R.

    1962-11-20

    A method is given for recovertng aluminum values from aqueous solutions containing said values together with fission products. A mixture of Fe/sub 2/O/ sub 3/ and MnO/sub 2/ is added to a solution containing aluminum and fission products. The resulting aluminum-containing supernatant is then separated from the fission product-bearing metal oxide precipitate and is contacted with a cation exchange resin. The aluminum sorbed on the resin is then eluted and recovered. (AEC)

  5. FISSION PRODUCT REMOVAL FROM ORGANIC SOLUTIONS

    DOEpatents

    Moore, R.H.

    1960-05-10

    The decontamination of organic solvents from fission products and in particular the treatment of solvents that were used for the extraction of uranium and/or plutonium from aqueous acid solutions of neutron-irradiated uranium are treated. The process broadly comprises heating manganese carbonate in air to a temperature of between 300 and 500 deg C whereby manganese dioxide is formed; mixing the manganese dioxide with the fission product-containing organic solvent to be treated whereby the fission products are precipitated on the manganese dioxide; and separating the fission product-containing manganese dioxide from the solvent.

  6. Heavy element fission products on earth

    NASA Astrophysics Data System (ADS)

    Shukoliukov, Iu. A.

    Current data on the products of spontaneous fission in radioactive minerals, lithospheric rocks, and atmosphere are presented. Methods of nuclear geochronology are discussed together with the role of Pu-244 in the isotopic balance of the earth. Natural chain fission reactions are examined with particular reference to the Oklo phenomenon. The discussion covers geological and chemical features of the Oklo deposits, evaluation of the Oklo fission-product data, and prospects for discovering other natural reactors of this type.

  7. METHOD FOR SEPARATING PLUTONIUM AND FISSION PRODUCTS EMPLOYING AN OXIDE AS A CARRIER FOR FISSION PRODUCTS

    DOEpatents

    Davies, T.H.

    1961-07-18

    Carrier precipitation processes for separating plutonium values from uranium fission products are described. Silicon dioxide or titanium dioxide in a finely divided state is added to an acidic aqueous solution containing hexavalent plutonium ions together with ions of uranium fission products. The supernatant solution containing plutonium ions is then separated from the oxide and the fission products associated therewith.

  8. PROCESS FOR SEPARATING URANIUM FISSION PRODUCTS

    DOEpatents

    Spedding, F.H.; Butler, T.A.; Johns, I.B.

    1959-03-10

    The removal of fission products such as strontium, barium, cesium, rubidium, or iodine from neutronirradiated uranium is described. Uranium halide or elemental halogen is added to melted irradiated uranium to convert the fission products to either more volatile compositions which vaporize from the melt or to higher melting point compositions which separate as solids.

  9. Fission Product Sorptivity in Graphite

    SciTech Connect

    Tompson, Jr., Robert V.; Loyalka, Sudarshan; Ghosh, Tushar; Viswanath, Dabir; Walton, Kyle; Haffner, Robert

    2015-04-01

    Both adsorption and absorption (sorption) of fission product (FP) gases on/into graphite are issues of interest in very high temperature reactors (VHTRs). In the original proposal, we proposed to use packed beds of graphite particles to measure sorption at a variety of temperatures and to use an electrodynamic balance (EDB) to measure sorption onto single graphite particles (a few μm in diameter) at room temperature. The use of packed beds at elevated temperature is not an issue. However, the TPOC requested revision of this initial proposal to included single particle measurements at elevated temperatures up to 1100 °C. To accommodate the desire of NEUP to extend the single particle EDB measurements to elevated temperatures it was necessary to significantly revise the plan and the budget. These revisions were approved. In the EDB method, we levitate a single graphite particle (the size, surface characteristics, morphology, purity, and composition of the particle can be varied) or agglomerate in the balance and measure the sorption of species by observing the changes in mass. This process involves the use of an electron stepping technique to measure the total charge on a particle which, in conjunction with the measured suspension voltages for the particle, allows for determinations of mass and, hence, of mass changes which then correspond to measurements of sorption. Accommodating elevated temperatures with this type of system required a significant system redesign and required additional time that ultimately was not available. These constraints also meant that the grant had to focus on fewer species as a result. Overall, the extension of the original proposed single particle work to elevated temperatures added greatly to the complexity of the proposed project and added greatly to the time that would eventually be required as well. This means that the bulk of the experimental progress was made using the packed bed sorption systems. Only being able to recruit one

  10. SOURCE OF PRODUCTS OF NUCLEAR FISSION

    DOEpatents

    Harteck, P.; Dondes, S.

    1960-03-15

    A source of fission product recoil energy suitable for use in radiation chemistry is reported. The source consists of thermal neutron irradiated glass wool having a diameter of 1 to 5 microns and containing an isotope fissionable by thermal neutrons, such as U/sup 235/.

  11. Modeling Fission Product Sorption in Graphite Structures

    SciTech Connect

    Szlufarska, Izabela; Morgan, Dane; Allen, Todd

    2013-04-08

    The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high- temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributions of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission

  12. Multiscale development of a fission gas thermal conductivity model: Coupling atomic, meso and continuum level simulations

    NASA Astrophysics Data System (ADS)

    Tonks, Michael R.; Millett, Paul C.; Nerikar, Pankaj; Du, Shiyu; Andersson, David; Stanek, Christopher R.; Gaston, Derek; Andrs, David; Williamson, Richard

    2013-09-01

    Fission gas production and evolution significantly impact the fuel performance, causing swelling, a reduction in the thermal conductivity and fission gas release. However, typical empirical models of fuel properties treat each of these effects separately and uncoupled. Here, we couple a fission gas release model to a model of the impact of fission gas on the fuel thermal conductivity. To quantify the specific impact of grain boundary (GB) bubbles on the thermal conductivity, we use atomistic and mesoscale simulations. Atomistic molecular dynamic simulations were employed to determine the GB thermal resistance. These values were then used in mesoscale heat conduction simulations to develop a mechanistic expression for the effective GB thermal resistance of a GB containing gas bubbles, as a function of the percentage of the GB covered by fission gas. The coupled fission gas release and thermal conductivity model was implemented in Idaho National Laboratory's BISON fuel performance code to model the behavior of a 10-pellet LWR fuel rodlet, showing how the fission gas impacts the UO2 thermal conductivity. Furthermore, additional BISON simulations were conducted to demonstrate the impact of average grain size on both the fuel thermal conductivity and the fission gas release.

  13. Fission product release from irradiated LWR fuel under accident conditions

    SciTech Connect

    Strain, R.V.; Sanecki, J.E.; Osborne, M.F.

    1984-01-01

    Fission product release from irradiated LWR fuel is being studied by heating fuel rod segments in flowing steam and an inert carrier gas to simulate accident conditions. Fuels with a range of irradiation histories are being subjected to several steam flow rates over a wide range of temperatures. Fission product release during each test is measured by gamma spectroscopy and by detailed examination of the collection apparatus after the test has been completed. These release results are complemented by a detailed posttest examination of samples of the fuel rod segment. Results of release measurements and fuel rod characterizations for tests at 1400 through 2000/sup 0/C are presented in this paper.

  14. USA/FRG umbrella agreement for cooperation in GCR [Gas Cooled Reactor] development: Fuel, fission products and graphite subprogram. Part 1, Management meeting report: Part 2, Revised subprogram plan, Revision 10

    SciTech Connect

    1986-05-01

    This Subprogram Plan describes cooperative work in the areas of HTR fuel and graphite development and fission product studies that is being carried out under US/FRG/Swiss Implementing Agreement for cooperation in Gas Cooled Reactor development. Only bilateral US/FRG cooperation is included, since it is the only active work in this subprogram area at this time. The cooperation has been in progress since February 1977. A number of Project Work Statements have been developed in each of the major areas of the subprogram, and work on many of them is in progress. The following specific areas are included in the scope of this plan: fuel development; graphite development; fission product release; and fission product behavior outside the fuel elements.

  15. Fission Product Decay Heat Calculations for Neutron Fission of 232Th

    NASA Astrophysics Data System (ADS)

    Son, P. N.; Hai, N. X.

    2016-06-01

    Precise information on the decay heat from fission products following times after a fission reaction is necessary for safety designs and operations of nuclear-power reactors, fuel storage, transport flasks, and for spent fuel management and processing. In this study, the timing distributions of fission products' concentrations and their integrated decay heat as function of time following a fast neutron fission reaction of 232Th were exactly calculated by the numerical method with using the DHP code.

  16. 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.

  17. Fission gas release restrictor for breached fuel rod

    DOEpatents

    Kadambi, N. Prasad; Tilbrook, Roger W.; Spencer, Daniel R.; Schwallie, Ambrose L.

    1986-01-01

    In the event of a breach in the cladding of a rod in an operating liquid metal fast breeder reactor, the rapid release of high-pressure gas from the fission gas plenum may result in a gas blanketing of the breached rod and rods adjacent thereto which impairs the heat transfer to the liquid metal coolant. In order to control the release rate of fission gas in the event of a breached rod, the substantial portion of the conventional fission gas plenum is formed as a gas bottle means which includes a gas pervious means in a small portion thereof. During normal reactor operation, as the fission gas pressure gradually increases, the gas pressure interiorly of and exteriorly of the gas bottle means equalizes. In the event of a breach in the cladding, the gas pervious means in the gas bottle means constitutes a sufficient restriction to the rapid flow of gas therethrough that under maximum design pressure differential conditions, the fission gas flow through the breach will not significantly reduce the heat transfer from the affected rod and adjacent rods to the liquid metal heat transfer fluid flowing therebetween.

  18. Rapid separation of fresh fission products (draft)

    SciTech Connect

    Dry, D. E.; Bauer, E.; Petersen, L. A.

    2003-01-01

    The fission of highly eruiched uranium by thermal neutrons creates dozens of isotopic products. The Isotope and Nuclear Chemistry Group participates in programs that involve analysis of 'fiesh' fission products by beta counting following radiochemical separations. This is a laborious and time-consuming process that can take several days to generate results. Gamma spectroscopy can provide a more immediate path to isolopic activities, however short-lived, high-yield isotopes can swamp a gamma spectrum, making difficult the identification and quantification of isotopes on the wings and valley of the fission yield curve. The gamma spectrum of a sample of newly produced fission products is dominated by the many emissions of a very few high-yield isotopes. Specilkally, {sup 132}Te (3.2 d), its daughter, {sup 132}I(2 .28 h), {sup 140}Ba (12.75 d), and its daughter {sup 140}La (1.68 d) emit at least 18 gamma rays above 100 keV that are greater than 5% abundance. Additionally, the 1596 keV emission fiom I4'La imposes a Compton background that hinders the detection of isotopes that are neither subject to matrix dependent fractionation nor gaseous or volatile recursors. Some of these isotopes of interest are {sup 111}Ag, {sup 115}Cd, and the rare earths, {sup 153}Sm, {sup 154}Eu, {sup 156}Eu, and {sup 160}Tb. C-INC has performed an HEU irradiation and also 'cold' carrier analyses by ICP-AES to determine methods for rapid and reliable separations that may be used to detect and quantify low-yield fission products by gamma spectroscopy. Results and progress will be presented.

  19. Sensitivity analysis of the fission gas behavior model in BISON.

    SciTech Connect

    Swiler, Laura Painton; Pastore, Giovanni; Perez, Danielle; Williamson, Richard

    2013-05-01

    This report summarizes the result of a NEAMS project focused on sensitivity analysis of a new model for the fission gas behavior (release and swelling) in the BISON fuel performance code of Idaho National Laboratory. Using the new model in BISON, the sensitivity of the calculated fission gas release and swelling to the involved parameters and the associated uncertainties is investigated. The study results in a quantitative assessment of the role of intrinsic uncertainties in the analysis of fission gas behavior in nuclear fuel.

  20. REMOVAL OF FISSION PRODUCTS FROM WATER

    DOEpatents

    Rosinski, J.

    1961-12-19

    A process is given for precipitating fission products from a body of water having a pH of above 6.5. Calcium permanganate and ferrous sulfate are added in a molar ratio of l: 3, whereby a mixed precipitate of manganese dioxide, ferric hydroxide and calcium sulfate is formed; the precipitate carries the fisston products and settles to the bottom of the body of water. (AEC)

  1. SPIDER Progress Towards High Resolution Correlated Fission Product Data

    NASA Astrophysics Data System (ADS)

    Shields, Dan; Meierbachtol, Krista; Tovesson, Fredrik; Arnold, Charles; Blackeley, Rick; Bredeweg, Todd; Devlin, Matt; Hecht, Adam; Jandel, Marian; Jorgenson, Justin; Nelson, Ron; White, Morgan; Spider Team

    2014-09-01

    The SPIDER detector (SPectrometer for Ion DEtermination in fission Research) is under development with the goal of obtaining high-resolution, high-efficiency, correlated fission product data needed for many applications including the modeling of next generation nuclear reactors, stockpile stewardship, and the fundamental understanding of the fission process. SPIDER simultaneously measures velocity and energy of both fission products to calculate fission product yields (FPYs), neutron multiplicity (ν), and total kinetic energy (TKE). A detailed description of the prototype SPIDER detector components will be presented. Characterization measurements with alpha and spontaneous fission sources will also be discussed. LA-UR-14-24875. The SPIDER detector (SPectrometer for Ion DEtermination in fission Research) is under development with the goal of obtaining high-resolution, high-efficiency, correlated fission product data needed for many applications including the modeling of next generation nuclear reactors, stockpile stewardship, and the fundamental understanding of the fission process. SPIDER simultaneously measures velocity and energy of both fission products to calculate fission product yields (FPYs), neutron multiplicity (ν), and total kinetic energy (TKE). A detailed description of the prototype SPIDER detector components will be presented. Characterization measurements with alpha and spontaneous fission sources will also be discussed. LA-UR-14-24875. This work is in part supported by LANL Laboratory Directed Research and Development Projects 20110037DR and 20120077DR.

  2. Nuclear Fission and Fission{minus}Product Spectroscopy: Second International Workshop. Proceedings

    SciTech Connect

    Fioni, G.; Faust, H.; Oberstedt, S.; Hambsch, F.

    1998-10-01

    These proceedings represent papers presented at the Second International Workshop on Nuclear Fission and Fission{minus}Product Spectroscopy held in Seyssins, France in April, 1998. The objective was to bring together the specialists in the field to overview the situation and to assess our present understanding of the fission process. The topics presented at the conference included nuclear waste management, incineration, neutron driven transmutation, leakage etc., radioactive beams, neutron{minus}rich nuclei, neutron{minus}induced and spontaneous fission, ternary fission phenomena, angular momentum, parity and time{minus}reversal phenomena, and nuclear fission at higher excitation energy. Modern spectroscopic tools for gamma spectroscopy as applied to fission were also discussed. There were 53 papers presented at the conference,out of which 3 have been abstracted for the Energy,Science and Technology database.(AIP)

  3. Modeling of Fission Gas Release in UO2

    SciTech Connect

    MH Krohn

    2006-01-23

    A two-stage gas release model was examined to determine if it could provide a physically realistic and accurate model for fission gas release under Prometheus conditions. The single-stage Booth model [1], which is often used to calculate fission gas release, is considered to be oversimplified and not representative of the mechanisms that occur during fission gas release. Two-stage gas release models require saturation at the grain boundaries before gas is release, leading to a time delay in release of gases generated in the fuel. Two versions of a two-stage model developed by Forsberg and Massih [2] were implemented using Mathcad [3]. The original Forsbers and Massih model [2] and a modified version of the Forsberg and Massih model that is used in a commercially available fuel performance code (FRAPCON-3) [4] were examined. After an examination of these models, it is apparent that without further development and validation neither of these models should be used to calculate fission gas release under Prometheus-type conditions. There is too much uncertainty in the input parameters used in the models. In addition. the data used to tune the modified Forsberg and Massih model (FRAPCON-3) was collected under commercial reactor conditions, which will have higher fission rates relative to Prometheus conditions [4].

  4. 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.

  5. SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS

    DOEpatents

    Nicholls, C.M.; Wells, I.; Spence, R.

    1959-10-13

    The separation of uranium and plutonium from neutronirradiated uranium is described. The neutron-irradiated uranium is dissolved in nitric acid to provide an aqueous solution 3N in nitric acid. The fission products of the solution are extruded by treating the solution with dibutyl carbitol substantially 1.8N in nitric acid. The organic solvent phase is separated and neutralized with ammonium hydroxide and the plutonium reduced with hydroxylamine base to the trivalent state. Treatment of the mixture with saturated ammonium nitrate extracts the reduced plutonium and leaves the uranium in the organic solvent.

  6. Data summary report for fission product release test VI-6

    SciTech Connect

    Osborne, M.F.; Lorenz, R.A.; Travis, J.R.; Webster, C.S.; Collins, J.L.

    1994-03-01

    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 directly 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.

  7. Ceramic Hosts for Fission Products Immobilization

    SciTech Connect

    Peter C Kong

    2010-07-01

    Natural spinel, perovskite and zirconolite rank among the most leach resistant of mineral forms. They also have a strong affinity for a large number of other elements and including actinides. Specimens of natural perovskite and zirconolite were radioisotope dated and found to have survived at least 2 billion years of natural process while still remain their loading of uranium and thorium . Developers of the Synroc waste form recognized and exploited the capability of these minerals to securely immobilize TRU elements in high-level waste . However, the Synroc process requires a relatively uniform input and hot pressing equipment to produce the waste form. It is desirable to develop alternative approaches to fabricate these durable waste forms to immobilize the radioactive elements. One approach is using a high temperature process to synthesize these mineral host phases to incorporate the fission products in their crystalline structures. These mineral assemblages with immobilized fission products are then isolated in a durable high temperature glass for periods measured on a geologic time scale. This is a long term research concept and will begin with the laboratory synthesis of the pure spinel (MgAl2O4), perovskite (CaTiO3) and zirconolite (CaZrTi2O7) from their constituent oxides. High temperature furnace and/or thermal plasma will be used for the synthesis of these ceramic host phases. Nonradioactive strontium oxide will be doped into these ceramic phases to investigate the development of substitutional phases such as Mg1-xSrxAl2O4, Ca1-xSrxTiO3 and Ca1-xSrxZrTi2O7. X-ray diffraction will be used to establish the crystalline structures of the pure ceramic hosts and the substitution phases. Scanning electron microscopy and energy dispersive X-ray analysis (SEM-EDX) will be performed for product morphology and fission product surrogates distribution in the crystalline hosts. The range of strontium doping is planned to reach the full substitution of the divalent

  8. 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.

  9. SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS

    DOEpatents

    Boyd, G.E.; Adamson, A.W.; Schubert, J.; Russell, E.R.

    1958-10-01

    A chromatographic adsorption process is presented for the separation of plutonium from other fission products formed by the irradiation of uranium. The plutonium and the lighter element fission products are adsorbed on a sulfonated phenol-formaldehyde resin bed from a nitric acid solution containing the dissolved uranium. Successive washes of sulfuric, phosphoric, and nitric acids remove the bulk of the fission products, then an eluate of dilute phosphoric and nitric acids removes the remaining plutonium and fission products. The plutonium is selectively removed by passing this solution through zirconium phosphate, from which the plutonium is dissolved with nitric acid. This process provides a convenient and efficient means for isolating plutonium.

  10. DESIGN OF AN ON-LINE, MULTI-SPECTROMETER FISSION PRODUCT MONITORING SYSTEM (FPMS) TO SUPPORT ADVANCED GAS REACTOR (AGR) FUEL TESTING AND QUALIFICATION IN THE ADVANCED TEST REACTOR

    SciTech Connect

    J. K. Hartwell; D. M. Scates; M. W. Drigert

    2005-11-01

    The US Department of Energy (DOE) is embarking on a series of tests of coated-particle reactor fuel for the Advanced Gas Reactor (AGR). As one part of this fuel development program a series of eight (8) fuel irradiation tests are planned for the Idaho National Laboratory’s (INL’s) Advanced Test Reactor (ATR). The first test in this series (AGR-1) will incorporate six separate “capsules” irradiated simultaneously, each containing about 51,000 TRISO-coated fuel particles supported in a graphite matrix and continuously swept with inert gas during irradiation. The effluent gas from each of the six capsules must be independently monitored in near real time and the activity of various fission gas nuclides determined and reported. A set of seven heavily-shielded high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based total radiation detectors have been designed, and are being configured and tested for use during the AGR-1 experiment. The AGR-1 test specification requires that the AGR-1 fission product measurement system (FPMS) have sufficient sensitivity to detect the failure of a single coated fuel particle and sufficient range to allow it to “count” multiple (up to 250) successive particle failures. This paper describes the design and expected performance of the AGR-1 FPMS.

  11. Fission product plateout/liftoff/washoff test plan. Revision 1

    SciTech Connect

    Acharya, R.; Hanson, D.

    1988-05-01

    A test program is planned in the COMEDIE loop of the Commissariat a l`Energy Atomique (CEA), Grenoble, France, to generate integral test data for the validation of computer codes used to predict fission product transport and core corrosion in the Modular High Temperature Gas-Cooled Reactor (MHTGR). The inpile testing will be performed by the CEA under contract from the US Department of Energy (DOE); the contract will be administered by Oak Ridge National Laboratory (ORNL). The primary purpose of this test plan is to provide an overview of the proposed program in terms of the overall scope and schedule. 8 refs, 3 figs.

  12. Acoustic Sensors for Fission Gas Characterization in MTR Harsh Environment

    NASA Astrophysics Data System (ADS)

    Very, F.; Rosenkrantz, E.; Fourmentel, D.; Destouches, C.; Villard, J. F.; Combette, P.; Ferrandis, J. Y.

    Our group is now working for more than 15 years, in a close partnership with CEA, on the development of acoustic sensors devoted to the characterization of fission gas release for in-pile experiments in Material Testing Reactor. First of all, we will present the main principle of the method and the result of a first succeed experiment called REMORA 3 used to differentiate helium and fission gas released kinetics under transient operating condition [1]. Then we will present our new researches involving thick film transducers produced by screen-printing process in order to propose piezoelectric structures for harsh temperature and irradiation measurements in new MTR reactor.

  13. Transmission electron microscopy characterization of the fission gas bubble superlattice in irradiated U-7wt% Mo dispersion fuels

    SciTech Connect

    B.D. Miller; J. Gan; D.D. Keiser Jr.; A.B. Robinson; J.-F. Jue; J.W. Madden; P.G. Medvedev

    2015-03-01

    Transmission electron microscopy characterization of irradiated U-7wt% Mo dispersion fuel was performed on various samples to understand the effect of irradiation parameters (fission density, fission rate, and temperature) on the self-organized fission-gas-bubble superlattice that forms in the irradiated U-Mo fuel. The bubble superlattice was seen to form a face-centered cubic structure coherent with the host U-7wt% Mo body centered cubic structure. At a fission density between 3.0 and 4.5 x 1021 fiss/cm3, the superlattice bubbles appear to have reached a saturation size with additional fission gas associated with increasing burnup predominately accumulating along grain boundaries. At a fission density of ~4.5x1021 fiss/cm3, the U-7wt% Mo microstructure undergoes grain subdivision and can no longer support the ordered bubble superlattice. The fuel grains are primarily less than 500 nm in diameter with micron-size fission-gas bubbles present on the grain boundaries. Solid fission products decorate the inside surface of the micron-sized fission-gas bubbles. Residual superlattice bubbles are seen in areas where fuel grains remain micron sized. Potential mechanisms of the formation and collapse of the bubble superlattice are discussed.

  14. (Fuel, fission product, and graphite technology)

    SciTech Connect

    Stansfield, O.M.

    1990-07-25

    Travel to the Forschungszentrum (KFA) -- Juelich described in this report was for the purpose of participating in the annual meeting of subprogram managers for the US/DOE Umbrella Agreement for Fuel, Fission Product, and Graphite Technology. At this meeting the highlights of the cooperative exchange were reviewed for the time period June 1989 through June 1990. The program continues to contribute technology in an effective way for both countries. Revision 15 of the Subprogram Plan will be issued as a result of the meeting. There was interest expressed by KFA management in the level of support received from the NPR program and in potential participation in the COMEDIE loop experiment being conducted at the CEA.

  15. Immobilization of fission products in phosphate ceramic waste forms

    SciTech Connect

    Singh, D.

    1996-10-01

    The goal of this project is to develop and demonstrate the feasibility of a novel low-temperature solidification/stabilization (S/S) technology for immobilizing waste streams containing fission products such as cesium, strontium, and technetium in a chemically bonded phosphate ceramic. This technology can immobilize partitioned tank wastes and decontaminate waste streams containing volatile fission products.

  16. REGENERATION OF FISSION-PRODUCT-CONTAINING MAGNESIUM-THORIUM ALLOYS

    DOEpatents

    Chiotti, P.

    1964-02-01

    A process of regenerating a magnesium-thorium alloy contaminated with fission products, protactinium, and uranium is presented. A molten mixture of KCl--LiCl-MgCl/sub 2/ is added to the molten alloy whereby the alkali, alkaline parth, and rare earth fission products (including yttrium) and some of the thorium and uranium are chlorinated and

  17. Chemical factors affecting fission product transport in severe LMFBR accidents

    SciTech Connect

    Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

    1984-10-01

    This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly.

  18. SEPARATION OF FISSION PRODUCTS FROM PLUTONIUM BY PRECIPITATION

    DOEpatents

    Seaborg, G.T.; Thompson, S.G.; Davidson, N.R.

    1959-09-01

    Fission product separation from hexavalent plutonium by bismuth phosphate precipitation of the fission products is described. The precipitation, according to this invention, is improved by coprecipitating ceric and zirconium phosphates (0.05 to 2.5 grams/liter) with the bismuth phosphate.

  19. Installation and Final Testing of an On-Line, Multi-Spectrometer Fission Product Monitoring System (FPMS) to Support Advanced Gas Reactor (AGR) Fuel Testing and Qualification in the Advanced Test Reactor

    SciTech Connect

    J. K. Hartwell; D. M. Scates; M. W. Drigert; J. B. Walter

    2006-10-01

    The US Department of Energy (DOE) is initiating tests of reactor fuel for use in an Advanced Gas Reactor (AGR). The AGR will use helium coolant, a low-power-density ceramic core, and coated-particle fuel. A series of eight (8) fuel irradiation tests are planned for the Idaho National Laboratory’s (INL’s) Advanced Test Reactor (ATR). One important measure of fuel performance in these tests is quantification of the fission gas releases over the nominal 2-year duration of each irradiation experiment. This test objective will be met using the AGR Fission Product Monitoring System (FPMS) which includes seven (7) on-line detection stations viewing each of the six test capsule effluent lines (plus one spare). Each station incorporates both a heavily-shielded high-purity germanium (HPGe) gamma-ray spectrometer for quantification of the isotopic releases, and a NaI(Tl) scintillation detector to monitor the total count rate and identify the timing of the releases. The AGR-1 experiment will begin irradiation after October 1, 2006. To support this experiment, the FPMS has been completely assembled, tested, and calibrated in a laboratory at the INL, and then reassembled and tested in its final location in the ATR reactor basement. This paper presents the details of the equipment performance, the control and acquisition software, the test plan for the irradiation monitoring, and the installation in the ATR basement. Preliminary on-line data may be available by the Conference date.

  20. Bulk and surface controlled diffusion of fission gas atoms

    SciTech Connect

    Andersson, Anders D.

    2012-08-09

    Fission gas retention and release impact nuclear fuel performance by, e.g., causing fuel swelling leading to mechanical interaction with the clad, increasing the plenum pressure and reducing the gap thermal conductivity. All of these processes are important to understand in order to optimize operating conditions of nuclear reactors and to simulate accident scenarios. Most fission gases have low solubility in the fuel matrix, which is especially pronounced for large fission gas atoms such as Xe and Kr, and as a result there is a significant driving force for segregation of gas atoms to extended defects such as grain boundaries or dislocations and subsequently for nucleation of gas bubbles at these sinks. Several empirical or semi-empirical models have been developed for fission gas release in nuclear fuels, e.g. [1-6]. One of the most commonly used models in fuel performance codes was published by Massih and Forsberg [3,4,6]. This model is similar to the early Booth model [1] in that it applies an equivalent sphere to separate bulk UO{sub 2} from grain boundaries represented by the sphere circumference. Compared to the Booth model, it also captures trapping at grain boundaries, fission gas resolution and it describes release from the boundary by applying timedependent boundary conditions to the circumference. In this work we focus on the step where fission gas atoms diffuse from the grain interior to the grain boundaries. The original Massih-Forsberg model describes this process by applying an effective diffusivity divided into three temperature regimes. In this report we present results from density functional theory calculations (DFT) that are relevant for the high (D{sub 3}) and intermediate (D{sub 2}) temperature diffusivities of fission gases. The results are validated by making a quantitative comparison to Turnbull's [8-10] and Matzke's data [12]. For the intrinsic or high temperature regime we report activation energies for both Xe and Kr diffusion in UO

  1. (COMEDIE program review and fission product transport in MHTGR reactor)

    SciTech Connect

    Stansfield, O.M.

    1990-03-15

    The subcontract between Martin Marietta Energy Systems, Inc., and the CEA provides for the refurbishment of the high pressure COMEDIE test loop in the SILOE reactor and a series of experiments to characterize fission product lift-off from MHTGR heat exchanger surfaces under several depressurization accident scenarios. The data will contribute to the validation of models and codes used to predict fission product transport in the MHTGR. In the meeting at CEA headquarters in Paris the program schedule and preparation for the DCAA and Quality Assurance audits were discussed. Long-range interest in expanded participation in the gas-cooled reactor technology Umbrella Agreement was also expressed by the CEA. At the CENG, in Grenoble, technical details on the loop design, fabrication components, development of test procedures, and preparation for the DOE quality assurance (QA) audit in May were discussed. After significant delays in CY 1989 it appears that good progress is being made in CY 1990 and the first major test will be initiated by December. An extensive list of agreements and commitments was generated to facilitate the coordination and planning of future work. 2 figs., 2 tabs.

  2. Chemistry of fission product iodine under nuclear reactor accident conditions

    SciTech Connect

    Malinauskas, A.P.; Bell, J.T.

    1986-01-01

    The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs.

  3. Fission Product Transmutation in Mixed Radiation Fields

    SciTech Connect

    Harmon, Frank; Burgett, Erick; Starovoitova, Valeriia; Tsveretkov, Pavel

    2015-01-15

    Work under this grant addressed a part of the challenge facing the closure of the nuclear fuel cycle; reducing the radiotoxicity of lived fission products (LLFP). It was based on the possibility that partitioning of isotopes and accelerator-based transmutation on particular LLFP combined with geological disposal may lead to an acceptable societal solution to the problem of management. The feasibility of using photonuclear processes based on the excitation of the giant dipole resonance (GDR) by bremsstrahlung radiation as a cost effective transmutation method was accessed. The nuclear reactions of interest: (γ,xn), (n,γ), (γ,p) can be induced by bremsstrahlung radiation produced by high power electron accelerators. The driver of these processes would be an accelerator that produces a high energy and high power electron beam of ~ 100 MeV. The major advantages of such accelerators for this purpose are that they are essentially available “off the shelf” and potentially would be of reasonable cost for this application. Methods were examined that used photo produced neutrons or the bremsstrahlung photons only, or use both photons and neutrons in combination for irradiations of selected LLFP. Extrapolating the results to plausible engineering scale transmuters it was found that the energy cost for 129I and 99Tc transmutation by these methods are about 2 and 4%, respectively, of the energy produced from 1000MWe.

  4. Fission gas bubble percolation on crystallographically consistent grain boundary networks

    NASA Astrophysics Data System (ADS)

    Sabogal-Suárez, Daniel; David Alzate-Cardona, Juan; Restrepo-Parra, Elisabeth

    2016-07-01

    Fission gas release in nuclear fuels can be modeled in the framework of percolation theory, where each grain boundary is classified as open or closed to the release of the fission gas. In the present work, two-dimensional grain boundary networks were assembled both at random and in a crystallographically consistent manner resembling a general textured microstructure. In the crystallographically consistent networks, grain boundaries were classified according to its misorientation. The percolation behavior of the grain boundary networks was evaluated as a function of radial cracks and radial thermal gradients in the fuel pellet. Percolation thresholds tend to shift to the left with increasing length and number of cracks, especially in the presence of thermal gradients. In general, the topology and percolation behavior of the crystallographically consistent networks differs from those of the random network.

  5. Determination of 140La fission product interference factor for INAA

    NASA Astrophysics Data System (ADS)

    Ribeiro, Iberê S., Jr.; Genezini, Frederico A.; Saiki, Mitiko; Zahn, Guilherme S.

    2014-11-01

    Instrumental Neutron Activation Analysis (INAA) is a technique widely used to determine the concentration of several elements in several kinds of matrices. However if the sample of interest has higher relative uranium concentration the obtained results can be interfered by the uranium fission products. One of these cases that is affected by interference due to U fission is the 140La , because this radioisotope used in INAA for the determination of concentration the La is also produced by the -β of 140Ba , an uranium fission product. The 140La interference factor was studied in this work and a factor to describe its time dependence was obtained.

  6. The Oklo natural reactor: Cumulative fission yields and retentivity of the symmetric mass region fission products

    NASA Astrophysics Data System (ADS)

    De Laeter, J. R.; Rosman, K. J. R.; Smith, C. L.

    1980-10-01

    Solid source mass spectrometry has been used to determine the relative cumulative fission yields of five elements in three samples of uranium ore from reactor zones in the Oklo mine site. Eighteen fission chains covering the mass range from 105 ≤ A ≤ 130 have been measured for Pd, Ag, Cd, Sn and Te. These measurements have enabled a number of nuclear parameters to be calculated including the relative proportions of 235U, 238U and 239Pu involved in the fission process. The concentration of the five elements in the Oklo samples have also been measured using the stable isotope dilution technique. These values have then been compared to the estimates of the amount of these elements produced by fission under the conditions that are appropriate to the three samples. This procedure enables the retentivity of the elements in the reactor zones to be evaluated. Our work confirms the fact that Pd and Te are retained almost in their entirety in the samples, whereas the other three elements have been partially lost from the reactor site. Almost all the Cd fission products have been lost, and more than 50% of the Ag and Sn fission-produced material has been removed.

  7. Thermodynamics of fission products in UO2+-x

    SciTech Connect

    Nerikar, Pankaj V

    2009-01-01

    The stabilities of selected fission products - Xe, Cs, and Sr - are investigated as a function of non-stoichiometry x in UO{sub 2{+-}x}. In particular, density functional theory (OFT) is used to calculate the incorporation and solution energies of these fission products at the anion and cation vacancy sites, at the divacancy, and at the bound Schottky defect. In order to reproduce the correct insulating state of UO{sub 2}, the DFT calculations are performed using spin polarization and with the Hubbard U tenn. In general, higher charge defects are more soluble in the fuel matrix and the solubility of fission products increases as the hyperstoichiometry increases. The solubility of fission product oxides is also explored. CS{sub 2}O is observed as a second stable phase and SrO is found to be soluble in the UO{sub 2} matrix for all stoichiometries. These observations mirror experimentally observed phenomena.

  8. Fission gas bubble identification using MATLAB's image processing toolbox

    DOE PAGES

    Collette, R.; King, J.; Keiser, Jr., D.; Miller, B.; Madden, J.; Schulthess, J.

    2016-06-08

    Automated image processing routines have the potential to aid in the fuel performance evaluation process by eliminating bias in human judgment that may vary from person-to-person or sample-to-sample. In addition, this study presents several MATLAB based image analysis routines designed for fission gas void identification in post-irradiation examination of uranium molybdenum (U–Mo) monolithic-type plate fuels. Frequency domain filtration, enlisted as a pre-processing technique, can eliminate artifacts from the image without compromising the critical features of interest. This process is coupled with a bilateral filter, an edge-preserving noise removal technique aimed at preparing the image for optimal segmentation. Adaptive thresholding provedmore » to be the most consistent gray-level feature segmentation technique for U–Mo fuel microstructures. The Sauvola adaptive threshold technique segments the image based on histogram weighting factors in stable contrast regions and local statistics in variable contrast regions. Once all processing is complete, the algorithm outputs the total fission gas void count, the mean void size, and the average porosity. The final results demonstrate an ability to extract fission gas void morphological data faster, more consistently, and at least as accurately as manual segmentation methods.« less

  9. Post-irradiation Examination and Fission Product Inventory Analysis of AGR-1 Irradiation Capsules

    SciTech Connect

    J M Harp; P D Demkowicz; S A Ploger

    2012-10-01

    The AGR-1 experiment was the first in a series of Advanced Gas Reactor (AGR) experiments designed to test TRISO fuel under High Temperature Gas Reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post-irradiation examination (PIE) at INL’s Materials and Fuels Complex (MFC). The inventory and distribution of fission products, especially Ag-110m, was assessed and analyzed for all the components of the AGR-1 capsules. This data should help inform the study of fission product migration in coated particle fuel. Gamma spectrometry was used to measure the activity of various different fission products in the different components of the AGR-1 test train. Each capsule contained: 12 fuel compacts, a graphite holder that kept the fuel compacts in place, graphite spacers that were above and below the graphite holders and fuel compacts, gas lines through which a helium neon gas mixture flowed in and out of each capsule, and the stainless steel shell that contained the experiment. Gamma spectrometry results and the experimental techniques used to capture these results will be presented for all the capsule components. The components were assayed to determine the total activity of different fission products present in or on them. These totals are compared to the total expected activity of a particular fission product in the capsule based on predictions from physics simulation. Based on this metric, a significant fraction of the Ag-110m was detected outside the fuel compacts, but the amount varied highly between the 6 capsules. Very small fractions of Cs-137 (<2E-5), Cs-134 (<1e-5), and Eu-154 (<4e-4) were detected outside of the fuel compacts. Additionally, the distribution of select fission products in some of the components including the fuel compacts and the graphite holders were measured and will be discussed.

  10. Gaseous fission product management for molten salt reactors and vented fuel systems

    SciTech Connect

    Messenger, S. J.; Forsberg, C.; Massie, M.

    2012-07-01

    Fission gas disposal is one of the unresolved difficulties for Molten Salt Reactors (MSRs) and advanced reactors with vented fuel systems. As these systems operate, they produce many radioactive isotopes of xenon and krypton (e.g. {sup 135}Xe t{sub 1/2} = 9.14 hours and {sup 85}Kr t{sub 1/2}= 10.73 years). Removing these gases proves vital to the success of such reactor designs for two reasons. First, the gases act as large neutron sinks which decrease reactivity and must be counterbalanced by increasing fuel loading. Second, for MSRs, inert fission product gases naturally separate quickly from high temperature salts, thus creating high vapor pressure which poses safety concerns. For advanced reactors with solid vented fuel, the gases are allowed to escape into an off-gas system and thus must be managed. Because of time delays in transport of fission product gases in vented fuel systems, some of the shorter-lived radionuclides will decay away thereby reducing the fission gas source term relative to an MSR. To calculate the fission gas source term of a typical molten salt reactor, we modeled a 1000 MWe graphite moderated thorium MSR similar to that detailed in Mathieu et al. [1]. The fuel salt used in these calculations was LiF (78 mole percent) - (HN)F 4 (22 mole percent) with a heavy nuclide composition of 3.86% {sup 233}U and 96.14% {sup 232}Th by mass. Before we can remove the fission product gases produced by this reactor configuration, we must first develop an appropriate storage mechanism. The gases could be stored in pressurized containers but then one must be concerned about bottle failure. Methods to trap noble gases in matrices are expensive and complex. Alternatively, there are direct storage/disposal options: direct injection into the Earth or injecting a grout-based product into the Earth. Advances in drilling technologies, hydro fracture technologies, and methods for the sequestration of carbon dioxide from fossil fuel plants are creating new options

  11. Development of a multiscale thermal conductivity model for fission gas in UO2

    NASA Astrophysics Data System (ADS)

    Tonks, Michael R.; Liu, Xiang-Yang; Andersson, David; Perez, Danielle; Chernatynskiy, Aleksandr; Pastore, Giovanni; Stanek, Christopher R.; Williamson, Richard

    2016-02-01

    Accurately predicting changes in the thermal conductivity of light water reactor UO2 fuel throughout its lifetime in reactor is an essential part of fuel performance modeling. However, typical thermal conductivity models from the literature are empirical. In this work, we begin to develop a mechanistic thermal conductivity model by focusing on the impact of gaseous fission products, which is coupled to swelling and fission gas release. The impact of additional defects and fission products will be added in future work. The model is developed using a combination of atomistic and mesoscale simulation, as well as analytical models. The impact of dispersed fission gas atoms is quantified using molecular dynamics simulations corrected to account for phonon-spin scattering. The impact of intragranular bubbles is accounted for using an analytical model that considers phonon scattering. The impact of grain boundary bubbles is determined using a simple model with five thermal resistors that are parameterized by comparing to 3D mesoscale heat conduction results. When used in the BISON fuel performance code to model four reactor experiments, it produces reasonable predictions without having been fit to fuel thermocouple data.

  12. Compilation of fission product yields Vallecitos Nuclear Center

    SciTech Connect

    Rider, B.F.

    1980-01-01

    This document is the ninth in a series of compilations of fission yield data made at Vallecitos Nuclear Center in which fission yield measurements reported in the open literature and calculated charge distributions have been utilized to produce a recommended set of yields for the known fission products. The original data with reference sources, as well as the recommended yields are presented in tabular form for the fissionable nuclides U-235, Pu-239, Pu-241, and U-233 at thermal neutron energies; for U-235, U-238, Pu-239, and Th-232 at fission spectrum energies; and U-235 and U-238 at 14 MeV. In addition, U-233, U-236, Pu-240, Pu-241, Pu-242, Np-237 at fission spectrum energies; U-233, Pu-239, Th-232 at 14 MeV and Cf-252 spontaneous fission are similarly treated. For 1979 U234F, U237F, Pu249H, U234He, U236He, Pu238F, Am241F, Am243F, Np238F, and Cm242F yields were evaluated. In 1980, Th227T, Th229T, Pa231F, Am241T, Am241H, Am242Mt, Cm245T, Cf249T, Cf251T, and Es254T are also evaluated.

  13. Analysis of Fission Products on the AGR-1 Capsule Components

    SciTech Connect

    Paul A. Demkowicz; Jason M. Harp; Philip L. Winston; Scott A. Ploger

    2013-03-01

    The components of the AGR-1 irradiation capsules were analyzed to determine the retained inventory of fission products in order to determine the extent of in-pile fission product release from the fuel compacts. This includes analysis of (i) the metal capsule components, (ii) the graphite fuel holders, (iii) the graphite spacers, and (iv) the gas exit lines. The fission products most prevalent in the components were Ag-110m, Cs 134, Cs 137, Eu-154, and Sr 90, and the most common location was the metal capsule components and the graphite fuel holders. Gamma scanning of the graphite fuel holders was also performed to determine spatial distribution of Ag-110m and radiocesium. Silver was released from the fuel components in significant fractions. The total Ag-110m inventory found in the capsules ranged from 1.2×10 2 (Capsule 3) to 3.8×10 1 (Capsule 6). Ag-110m was not distributed evenly in the graphite fuel holders, but tended to concentrate at the axial ends of the graphite holders in Capsules 1 and 6 (located at the top and bottom of the test train) and near the axial center in Capsules 2, 3, and 5 (in the center of the test train). The Ag-110m further tended to be concentrated around fuel stacks 1 and 3, the two stacks facing the ATR reactor core and location of higher burnup, neutron fluence, and temperatures compared with Stack 2. Detailed correlation of silver release with fuel type and irradiation temperatures is problematic at the capsule level due to the large range of temperatures experienced by individual fuel compacts in each capsule. A comprehensive Ag 110m mass balance for the capsules was performed using measured inventories of individual compacts and the inventory on the capsule components. For most capsules, the mass balance was within 11% of the predicted inventory. The Ag-110m release from individual compacts often exhibited a very large range within a particular capsule.

  14. Evaluation and compilation of fission product yields 1993

    SciTech Connect

    England, T.R.; Rider, B.F.

    1995-12-31

    This document is the latest in a series of compilations of fission yield data. Fission yield measurements reported in the open literature and calculated charge distributions have been used to produce a recommended set of yields for the fission products. The original data with reference sources, and the recommended yields axe presented in tabular form. These include many nuclides which fission by neutrons at several energies. These energies include thermal energies (T), fission spectrum energies (F), 14 meV High Energy (H or HE), and spontaneous fission (S), in six sets of ten each. Set A includes U235T, U235F, U235HE, U238F, U238HE, Pu239T, Pu239F, Pu241T, U233T, Th232F. Set B includes U233F, U233HE, U236F, Pu239H, Pu240F, Pu241F, Pu242F, Th232H, Np237F, Cf252S. Set C includes U234F, U237F, Pu240H, U234HE, U236HE, Pu238F, Am241F, Am243F, Np238F, Cm242F. Set D includes Th227T, Th229T, Pa231F, Am241T, Am241H, Am242MT, Cm245T, Cf249T, Cf251T, Es254T. Set E includes Cf250S, Cm244S, Cm248S, Es253S, Fm254S, Fm255T, Fm256S, Np237H, U232T, U238S. Set F includes Cm243T, Cm246S, Cm243F, Cm244F, Cm246F, Cm248F, Pu242H, Np237T, Pu240T, and Pu242T to complete fission product yield evaluations for 60 fissioning systems in all. This report also serves as the primary documentation for the second evaluation of yields in ENDF/B-VI released in 1993.

  15. The behavior of fission products during nuclear rocket reactor tests

    SciTech Connect

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    The experience base regarding fission product behavior developed during the Rover program, the nuclear rocket development program of 1955--1972, will be useful in planning a renewed nuclear rocket program. During the Rover program, 20 reactors were tested at the Nuclear Rocket Development Station in Nevada. Nineteen of these discharged effluent directly into the atmosphere; the last reactor tested, a non-flight-prototypic, fuel-element-testing reactor called the Nuclear Furnace (NF-1) was connected to an effluent cleanup system that removed fission products before the hydrogen coolant (propellant) was discharged to the atmosphere. In general, we are able to increase both test duration and fuel temperature during the test series. Therefore fission product data from the later part of the program are more interesting and more applicable to future reactors. We have collected fission product retention (and release) data reported in both formal and informal publications for six of the later reactor tests; five of these were Los Alamos reactors that were firsts of a kind in configuration or operating conditions. We have also, with the cooperation of Westinghouse, included fission product data from the NRX-A6 reactor, the final member of series of developmental reactors with the same basic geometry, but with significant design and fabrication improvements as the series continued. Table 1 lists the six selected reactors and the test parameters for each.

  16. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

    2007-10-01

    The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

  17. Comparison of Fission Product Yields and Their Impact

    SciTech Connect

    S. Harrison

    2006-02-01

    This memorandum describes the Naval Reactors Prime Contractor Team (NRPCT) Space Nuclear Power Program (SNPP) interest in determining the expected fission product yields from a Prometheus-type reactor and assessing the impact of these species on materials found in the fuel element and balance of plant. Theoretical yield calculations using ORIGEN-S and RACER computer models are included in graphical and tabular form in Attachment, with focus on the desired fast neutron spectrum data. The known fission product interaction concerns are the corrosive attack of iron- and nickel-based alloys by volatile fission products, such as cesium, tellurium, and iodine, and the radiological transmutation of krypton-85 in the coolant to rubidium-85, a potentially corrosive agent to the coolant system metal piping.

  18. Fission Product Data Measured at Los Alamos for Fission Spectrum and Thermal Neutrons on 239Pu, 235U, 238U

    NASA Astrophysics Data System (ADS)

    Selby, H. D.; Mac Innes, M. R.; Barr, D. W.; Keksis, A. L.; Meade, R. A.; Burns, C. J.; Chadwick, M. B.; Wallstrom, T. C.

    2010-12-01

    We describe measurements of fission product data at Los Alamos that are important for determining the number of fissions that have occurred when neutrons are incident on plutonium and uranium isotopes. The fission-spectrum measurements were made using a fission chamber designed by the National Institute for Standards and Technology (NIST) in the BIG TEN critical assembly, as part of the Inter-laboratory Liquid Metal Fast Breeder Reactor (LMFBR) Reaction Rate (ILRR) collaboration. The thermal measurements were made at Los Alamos' Omega West Reactor. A related set of measurements were made of fission-product ratios (so-called R-values) in neutron environments provided by a number of Los Alamos critical assemblies that range from having average energies causing fission of 400-600 keV (BIG TEN and the outer regions of the Flattop-25 assembly) to higher energies (1.4-1.9 MeV) in the Jezebel, and in the central regions of the Flattop-25 and Flattop-Pu, critical assemblies. From these data we determine ratios of fission product yields in different fuel and neutron environments (Q-values) and fission product yields in fission spectrum neutron environments for 99Mo, 95Zr, 137Cs, 140Ba, 141,143Ce, and 147Nd. Modest incident-energy dependence exists for the 147Nd fission product yield; this is discussed in the context of models for fission that include thermal and dynamical effects. The fission product data agree with measurements by Maeck and other authors using mass-spectrometry methods, and with the ILRR collaboration results that used gamma spectroscopy for quantifying fission products. We note that the measurements also contradict earlier 1950s historical Los Alamos estimates by ˜5-7%, most likely owing to self-shielding corrections not made in the early thermal measurements. Our experimental results provide a confirmation of the England-Rider ENDF/B-VI evaluated fission-spectrum fission product yields that were carried over to the ENDF/B-VII.0 library, except for 99Mo

  19. FASTGRASS implementation in BISON and Fission gas behavior characterization in UO2 and connection to validating MARMOT

    SciTech Connect

    Yun, Di; Mo, Kun; Ye, Bei; Jamison, Laura M.; Miao, Yinbin; Lian, Jie; Yao, Tiankei

    2015-09-30

    This activity is supported by the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Product Line (FPL). Two major accomplishments in FY 15 are summarized in this report: (1) implementation of the FASTGRASS module in the BISON code; and (2) a Xe implantation experiment for large-grained UO2. Both BISON AND MARMOT codes have been developed by Idaho National Laboratory (INL) to enable next generation fuel performance modeling capability as part of the NEAMS Program FPL. To contribute to the development of the Moose-Bison-Marmot (MBM) code suite, we have implemented the FASTGRASS fission gas model as a module in the BISON code. Based on rate theory formulations, the coupled FASTGRASS module in BISON is capable of modeling LWR oxide fuel fission gas behavior and fission gas release. In addition, we conducted a Xe implantation experiment at the Argonne Tandem Linac Accelerator System (ATLAS) in order to produce the needed UO2 samples with desired bubble morphology. With these samples, further experiments to study the fission gas diffusivity are planned to provide validation data for the Fission Gas Release Model in MARMOT codes.

  20. A Covariance Generation Methodology for Fission Product Yields

    NASA Astrophysics Data System (ADS)

    Terranova, N.; Serot, O.; Archier, P.; Vallet, V.; De Saint Jean, C.; Sumini, M.

    2016-03-01

    Recent safety and economical concerns for modern nuclear reactor applications have fed an outstanding interest in basic nuclear data evaluation improvement and completion. It has been immediately clear that the accuracy of our predictive simulation models was strongly affected by our knowledge on input data. Therefore strong efforts have been made to improve nuclear data and to generate complete and reliable uncertainty information able to yield proper uncertainty propagation on integral reactor parameters. Since in modern nuclear data banks (such as JEFF-3.1.1 and ENDF/BVII.1) no correlations for fission yields are given, in the present work we propose a covariance generation methodology for fission product yields. The main goal is to reproduce the existing European library and to add covariance information to allow proper uncertainty propagation in depletion and decay heat calculations. To do so, we adopted the Generalized Least Square Method (GLSM) implemented in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation), developed at CEA-Cadarache. Theoretical values employed in the Bayesian parameter adjustment are delivered thanks to a convolution of different models, representing several quantities in fission yield calculations: the Brosa fission modes for pre-neutron mass distribution, a simplified Gaussian model for prompt neutron emission probability, theWahl systematics for charge distribution and the Madland-England model for the isomeric ratio. Some results will be presented for the thermal fission of U-235, Pu-239 and Pu-241.

  1. Data summary report for fission product release test VI-5

    SciTech Connect

    Osborne, M.F.; Lorenz, R.A.; Travis, J.R.; Webster, C.S.; Collins, J.L. )

    1991-10-01

    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 in 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.

  2. Applications for fission product data to problems in stellar nucleosynthesis

    SciTech Connect

    Mathews, G.J.

    1983-10-01

    A general overview of the nucleosynthesis mechanisms for heavy (A greater than or equal to 70) nuclei is presented with particular emphasis on critical data needs. The current state of the art in nucleosynthesis models is described and areas in which fission product data may provide useful insight are proposed. 33 references, 10 figures.

  3. Fission properties and production mechanisms for the heaviest known elements

    SciTech Connect

    Hoffman, D.C.

    1981-01-01

    Mass yields of the spontaneous fission of Fm isotopes, Cf isotopes, and /sup 259/Md are discussed. Actinide yields were measured for bombardments of /sup 248/Cm with /sup 16/O, /sup 18/O, /sup 20/Ne, and /sup 22/Ne. A superheavy product might be produced by bombarding /sup 248/Cm with /sup 48/Ca ions. 12 figures. (DLC)

  4. SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS BY ADSORPTION

    DOEpatents

    Seaborg, G.T.; Willard, J.E.

    1958-01-01

    A method is presented for the separation of plutonium from solutions containing that element in a valence state not higher than 41 together with uranium ions and fission products. This separation is accomplished by contacting the solutions with diatomaceous earth which preferentially adsorbs the plutonium present. Also mentioned as effective for this adsorbtive separation are silica gel, filler's earth and alumina.

  5. Recent MELCOR and VICTORIA Fission Product Research at the NRC

    SciTech Connect

    Bixler, N.E.; Cole, R.K.; Gauntt, R.O.; Schaperow, J.H.; Young, M.F.

    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 in 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

  6. Analysis of fission gas release kinetics by on-line mass spectrometry

    SciTech Connect

    Zerega, Y.; Reynard-Carette, C.; Parrat, D.; Carette, M.; Brkic, B.; Lyoussi, A.; Bignan, G.; Janulyte, A.; Andre, J.; Pontillon, Y.; Ducros, G.; Taylor, S.

    2011-07-01

    a gamma spectrometer in the fission product laboratory of the future Jules Horowitz Material Test Reactor. An intermediate step will consist of testing first equipment on an existing experimental facility in the LECA-STAR Hot Cell Laboratory of the CEA Cadarache. This paper presents the scientific and operational stakes linked to fission gas issues, resumes the current state of art for analyzing them in nuclear facilities, then presents the skills gathered through this collaboration to overcome technological bottlenecks. Finally it describes the implementation strategy in nuclear research facilities of the CEA Cadarache. (authors)

  7. Measurement of fission gas release from irradiated Usbnd Mo dispersion fuel samples

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2016-09-01

    The uranium-molybdenum (Usbnd Mo) alloy dispersed in an Alsbnd Si matrix has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. In this paper, two irradiated samples containing 53.9 vol% U-7wt% Mo fuel particles dispersed in an Al-2wt% Si matrix were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Measurements revealed three distinct fission gas release events for the samples from 400 to 700 °C, as well as a number of minor fission gas releases below and above this temperature range. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature with exceptional agreement.

  8. Comparison of fission product release predictions using PARFUME with results from the AGR-1 irradiation experiment

    SciTech Connect

    Blaise Collin

    2014-09-01

    This report documents comparisons between post-irradiation examination measurements and model predictions of silver (Ag), cesium (Cs), and strontium (Sr) release from selected tristructural isotropic (TRISO) fuel particles and compacts during the first irradiation test of the Advanced Gas Reactor program that occurred from December 2006 to November 2009 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The modeling was performed using the particle fuel model computer code PARFUME (PARticle FUel ModEl) developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact) but it can be assessed at the particle level by adjusting the diffusivity in the fuel matrix to very high values. Furthermore, the diffusivity of each layer can be individually set to a high value (typically 10-6 m2/s) to simulate a failed layer with no capability of fission product retention. In this study, the comparison to PIE focused on fission product release and because of the lack of failure in the irradiation, the probability of particle failure was not calculated. During the AGR-1 irradiation campaign, the fuel kernel produced and released fission products, which migrated through the successive

  9. Design of pellet surface grooves for fission gas plenum

    SciTech Connect

    Carter, T.J.; Jones, L.R.; Macici, N.; Miller, G.C.

    1986-01-01

    In the Canada deuterium uranium pressurized heavy water reactor, short (50-cm) Zircaloy-4 clad bundles are fueled on-power. Although internal void volume within the fuel rods is adequate for the present once-through natural uranium cycle, the authors have investigated methods for increasing the internal gas storage volume needed in high-power, high-burnup, experimental ceramic fuels. This present work sought to prove the methodology for design of gas storage volume within the fuel pellets - specifically the use of grooves pressed or machined into the relatively cool pellet/cladding interface. Preanalysis and design of pellet groove shape and volume was accomplished using the TRUMP heat transfer code. Postirradiation examination (PIE) was used to check the initial design and heat transfer assumptions. Fission gas release was found to be higher for the grooved pellet rods than for the comparison rods with hollow or unmodified pellets. This had been expected from the initial TRUMP thermal analyses. The ELESIM fuel modeling code was used to check in-reactor performance, but some modifications were necessary to accommodate the loss of heat transfer surface to the grooves. It was concluded that for plenum design purposes, circumferential pellet grooves could be adequately modeled by the codes TRUMP and ELESIM.

  10. Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    SciTech Connect

    Dawn Scates

    2010-10-01

    A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

  11. Characterization of intergranular fission gas bubbles in U-Mo fuel.

    SciTech Connect

    Kim, Y. S.; Hofman, G.; Rest, J.; Shevlyakov, G. V.; Nuclear Engineering Division; SSCR RIAR

    2008-04-14

    plates were also performed and compared with the calculated quantities. Bubble density per unit grain boundary surface area was calculated by using the density per unit grain boundary length data. These data were used as input for mechanistic modeling described in section 4. Volumetric bubble density was calculated by using density per unit grain boundary surface area. Based on these data, bubble volumetric fraction was calculated. Bubble volume fraction was also calculated by using the density per unit cross section area. Bubble volume fraction was also directly measured for some plates. These three results are comparable although the direct measurement data are slightly larger than the others. Bubble volume fraction increased as a function of burnup, reaching {approx}2% of fuel volume at 3 x 10{sup 21} f/cm{sup 3}. Fission gas bubble swelling is minor compared to that of solid fission product swelling.

  12. The coupled kinetics of grain growth and fission product behavior in nuclear fuel under degraded-core accident conditions

    NASA Astrophysics Data System (ADS)

    Rest, J.

    1985-04-01

    The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, and cesium release from (1) irradiated high-burnup LWR fuel in a flowing steam atmosphere during high-temperature, in-cell heating tests (performed at Oak Ridge National Laboratory) and (2) trace-irradiated LWR fuel during severe-fuel-damage (SFD) tests (performed in the PBF reactor in Idaho). A theory of grain boundary sweeping of gas bubbles has been included within the FASTGRASS-VFP formalism. This theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges, and provides a means of determining whether gas bubbles are caught up and moved along by a moving grain boundary or whether the grain boundary is only temporarily retarded by the bubbles and then breaks away. In addition, as FASTGRASS-VFP provides for a mechanistic calculation of ultra- and intergranular fission product behavior, the coupled calculation between fission gas behavior and grain growth is kinetically comprehensive. Results of the analyses demonstrate that intragranular fission product behavior during both types of tests can be interpreted in terms of a grain-growth/grain-boundary-sweeping mechanism that enhances the flow of fission products from within the grains to the grain boundaries. The effect of fuel oxidation by steam on fission product and grain growth behavior is also considered. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in high-burnup fuel are highlighted.

  13. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

    DOE PAGES

    Pastore, Giovanni; Swiler, L. P.; Hales, Jason D.; Novascone, Stephen R.; Perez, Danielle M.; Spencer, Benjamin W.; Luzzi, Lelio; Uffelen, Paul Van; Williamson, Richard L.

    2014-10-12

    The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code and a recently implemented physics-based model for the coupled fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information from the open literature. The study leads to an initial quantitative assessment of the uncertaintymore » in fission gas behavior modeling with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.« less

  14. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

    NASA Astrophysics Data System (ADS)

    Pastore, Giovanni; Swiler, L. P.; Hales, J. D.; Novascone, S. R.; Perez, D. M.; Spencer, B. W.; Luzzi, L.; Van Uffelen, P.; Williamson, R. L.

    2015-01-01

    The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code with a recently implemented physics-based model for fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information in the open literature. The study leads to an initial quantitative assessment of the uncertainty in fission gas behavior predictions with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, significantly higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.

  15. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

    SciTech Connect

    Pastore, Giovanni; Swiler, L. P.; Hales, Jason D.; Novascone, Stephen R.; Perez, Danielle M.; Spencer, Benjamin W.; Luzzi, Lelio; Uffelen, Paul Van; Williamson, Richard L.

    2014-10-12

    The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code and a recently implemented physics-based model for the coupled fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information from the open literature. The study leads to an initial quantitative assessment of the uncertainty in fission gas behavior modeling with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.

  16. Measurement of fission gas release from irradiated U-Mo monolithic fuel samples

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine J.; Pool, Karl N.

    2015-06-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of heating post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium-molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1000 °C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature.

  17. Measurement of fission gas release from irradiated U–Mo monolithic fuel samples

    SciTech Connect

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine J.; Pool, Karl N.

    2015-06-01

    The uranium–molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of heating post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium–molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1000 °C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature.

  18. Measurement of Fission Gas Release from Irradiated U-Mo Monolithic Fuel Samples

    SciTech Connect

    Burkes, Douglas; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine; Pool, Karl N.

    2015-06-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of annealing post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium-molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1050 C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in literature.

  19. 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.

  20. Radiation re-solution of fission gas in non-oxide nuclear fuel

    NASA Astrophysics Data System (ADS)

    Matthews, Christopher; Schwen, Daniel; Klein, Andrew C.

    2015-02-01

    Renewed interest in fast nuclear reactors is creating a need for better understanding of fission gas bubble behavior in non-oxide fuels to support very long fuel lifetimes. Collisions between fission fragments and their subsequent cascades can knock fission gas atoms out of bubbles and back into the fuel lattice. We showed that these collisions can be treated as using the so-called "homogenous" atom-by-atom re-solution theory and calculated using the Binary Collision Approximation code 3DOT. The calculations showed that there is a decrease in the re-solution parameter as bubble radius increases until about 50 nm, at which the re-solution parameter stays nearly constant. Furthermore, our model shows ion cascades created in the fuel result in many more implanted fission gas atoms than collisions directly with fission fragments. This calculated re-solution parameter can be used to find a re-solution rate for future bubble simulations.

  1. Measurements for the JASPER Program fission gas plenum experiment

    SciTech Connect

    Muckenthaler, F.J.

    1987-06-01

    The Fission Gas Plenum Experiment was conducted at the Oak Ridge National Laboratory Tower Shielding Facility during FY 1987 to: provide data for verification of the assumptions and calculational methods used to determine the neutron leakage from the plenum, and provide an uncertainty evaluation associated with the calculations. The Tower Shielding Reactor source was modified to represent the neutron spectrum leaving a typical liquid-metal-cooled reactor core along its axis. The experimental configurations resulted from the insertion of either a homogeneous or homogeneous-heterogeneous gas plenum combination into the iris of a concrete slab, with the only variable being the thickness of the plenum. Integral neutron fluxes were measured behind each of the configurations at specified locations, and neutron spectra were obtained behind selected mockups. The experimental data are presented in both tabular and graphical form. This experiment is the second in a series of six experiments to be performed as part of a cooperative effort between the United States Department of Energy and the Japan Power Reactor and Nuclear Fuel Development Corporation. The research program is intended to provide support for the development of advanced sodium-cooled reactors.

  2. Analysis of fission product release behavior during the TMI-2 accident

    SciTech Connect

    Petti, D. A.; Adams, J. P.; Anderson, J. L.; Hobbins, R. R.

    1987-01-01

    An analysis of fission product release during the Three Mile Island Unit 2 (TMI-2) accident has been initiated to provide an understanding of fission product behavior that is consistent with both the best estimate accident scenario and fission product results from the ongoing sample acquisition and examination efforts. ''First principles'' fission product release models are used to describe release from intact, disrupted, and molten fuel. Conclusions relating to fission product release, transport, and chemical form are drawn. 35 refs., 12 figs., 7 tabs.

  3. Superabsorbing gel for actinide, lanthanide, and fission product decontamination

    DOEpatents

    Kaminski, Michael D.; Mertz, Carol J.

    2016-06-07

    The present invention provides an aqueous gel composition for removing actinide ions, lanthanide ions, fission product ions, or a combination thereof from a porous surface contaminated therewith. The composition comprises a polymer mixture comprising a gel forming cross-linked polymer and a linear polymer. The linear polymer is present at a concentration that is less than the concentration of the cross-linked polymer. The polymer mixture is at least about 95% hydrated with an aqueous solution comprising about 0.1 to about 3 percent by weight (wt %) of a multi-dentate organic acid chelating agent, and about 0.02 to about 0.6 molar (M) carbonate salt, to form a gel. When applied to a porous surface contaminated with actinide ions, lanthanide ions, and/or other fission product ions, the aqueous gel absorbs contaminating ions from the surface.

  4. Transmutation of selected fission products in a fast reactor

    SciTech Connect

    Wootan, D.W.; Nelson, J.V.

    1993-05-01

    A small fast reactor such as the Fast Flux Test Facility can be an effective device for destroying long-lived fission products such as {sup 99}Tc and {sup 129}I. There are several potential core configuration options using both fast and moderated neutron spectra. The calculated Doppler reactivity coefficient for {sup 99}Tc was 60% of the value for {sup 238}U on a per atom basis. Replacing {sup 238}U with {sup 99}Tc in waste burn applications has the positive attributes of reduced parasitic capture in uranium and enhanced fission product destruction, while retaining a substantial Doppler effect. A modular liquid metal reactor system could support about 8 to 10 comparably sized conventional light water reactors.

  5. NEANDC specialists meeting on yields and decay data of fission product nuclides

    SciTech Connect

    Chrien, R.E.; Burrows, T.W.

    1983-01-01

    Separate abstracts were prepared for the 29 papers presented. Workshop reports on decay heat, fission yields, beta- and gamma-ray spectroscopy, and delayed neutrons are included. An appendix contains a survey of the most recent compilations and evaluations containing fission product yield, fission product decay data, and delayed neutron yield information. (WHK)

  6. Fission product ion exchange between zeolite and a molten salt

    NASA Astrophysics Data System (ADS)

    Gougar, Mary Lou D.

    The electrometallurgical treatment of spent nuclear fuel (SNF) has been developed at Argonne National Laboratory (ANL) and has been demonstrated through processing the sodium-bonded SNF from the Experimental Breeder Reactor-II in Idaho. In this process, components of the SNF, including U and species more chemically active than U, are oxidized into a bath of lithium-potassium chloride (LiCl-KCl) eutectic molten salt. Uranium is removed from the salt solution by electrochemical reduction. The noble metals and inactive fission products from the SNF remain as solids and are melted into a metal waste form after removal from the molten salt bath. The remaining salt solution contains most of the fission products and transuranic elements from the SNF. One technique that has been identified for removing these fission products and extending the usable life of the molten salt is ion exchange with zeolite A. A model has been developed and tested for its ability to describe the ion exchange of fission product species between zeolite A and a molten salt bath used for pyroprocessing of spent nuclear fuel. The model assumes (1) a system at equilibrium, (2) immobilization of species from the process salt solution via both ion exchange and occlusion in the zeolite cage structure, and (3) chemical independence of the process salt species. The first assumption simplifies the description of this physical system by eliminating the complications of including time-dependent variables. An equilibrium state between species concentrations in the two exchange phases is a common basis for ion exchange models found in the literature. Assumption two is non-simplifying with respect to the mathematical expression of the model. Two Langmuir-like fractional terms (one for each mode of immobilization) compose each equation describing each salt species. The third assumption offers great simplification over more traditional ion exchange modeling, in which interaction of solvent species with each other

  7. 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

  8. [Biological effects of nuclear fission products. Repeated exposures].

    PubMed

    Vasilenko, I Ia

    1994-01-01

    The results of experimental studies on the repeated exposure to radioiodine (131I) and nuclear fission products (NFP) are presented, the doses used being equal to those resulted in radiation disease under first and second input. The animals satisfactory withstood the repeated exposure. The residual injuries appeared slightly. The animals' state was satisfactory during 5 years. Blastomogenic effect of NFP was revealed in remote periods.

  9. Molten salt extraction of transuranic and reactive fission products from used uranium oxide fuel

    DOEpatents

    Herrmann, Steven Douglas

    2014-05-27

    Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium/transuranic product recovery and fission product immobilization in engineered waste forms.

  10. Fusion-Fission Hybrid for Fissile Fuel Production without Processing

    SciTech Connect

    Fratoni, M; Moir, R W; Kramer, K J; Latkowski, J F; Meier, W R; Powers, J J

    2012-01-02

    Two scenarios are typically envisioned for thorium fuel cycles: 'open' cycles based on irradiation of {sup 232}Th and fission of {sup 233}U in situ without reprocessing or 'closed' cycles based on irradiation of {sup 232}Th followed by reprocessing, and recycling of {sup 233}U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE1) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuel composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in the critical reactors

  11. Fission track astrology of three Apollo 14 gas-rich breccias

    NASA Technical Reports Server (NTRS)

    Graf, H.; Shirck, J.; Sun, S.; Walker, R.

    1973-01-01

    The three Apollo 14 breccias 14301, 14313, and 14318 all show fission xenon due to the decay of Pu-244. To investigate possible in situ production of the fission gas, an analysis was made of the U-distribution in these three breccias. The major amount of the U lies in glass clasts and in matrix material and no more than 25% occurs in distinct high-U minerals. The U-distribution of each breccia is discussed in detail. Whitlockite grains in breccias 14301 and 14318 found with the U-mapping were etched and analyzed for fission tracks. The excess track densities are much smaller than indicated by the Xe-excess. Because of a preirradiation history documented by very high track densities in feldspar grains, however, it is impossible to attribute the excess tracks to the decay of Pu-244. A modified track method has been developed for measuring average U-concentrations in samples containing a heterogeneous distribution of U in the form of small high-U minerals. The method is briefly discussed, and results for the rocks 14301, 14313, 14318, 68815, 15595, and the soil 64421 are given.

  12. Fuel Performance Experience, Analysis and Modeling: Deformations, Fission Gas Release and Pellet-Clad Interaction

    SciTech Connect

    Zhou, G.; Hallstadius, L.; Helmersson, S.; Massih, A.R.; Schrire, D.; Kaellstroem, R.; Wikmark, G.; Hellwig, C.

    2007-07-01

    Some basic attributes of light water reactor fuel performance, determined by measurements, are evaluated. In particular, data on fuel volume swelling, cladding creep/growth, fission product gas release and cladding deformation due to pellet-clad mechanical interaction of rods irradiated in power reactors to rod burnups up to about 70 MWd/kgU are presented and appraised. A thermal fuel matrix swelling caused by fission products shows a linear increase in the fuel volume fraction with burnup up to 70 MWd/kgU with a mean rate of 0.76% per 10 MWd/kgU at a best-estimate level. Cladding hoop strain data due to in-reactor creep as a function of burnup from 15 to 70 MWd/kgU for pressurized water reactor (PWR) rods and from 5 to 50 MWd/kgU for boiling water reactor (BWR) rods are presented. The maximum measured cladding creep-down hoop strain in the considered BWR rods is {epsilon}{sub {theta}} {approx_equal} -0.5% and in the PWR rods {epsilon}{sub {theta}} {approx_equal} -1.25%. Rod growth data on BWR and PWR rods as a function of burnup are presented and discussed. Rod internal free volume data, measured and calculated as a function of burnup, are presented. Recent high burnup (52-70 MWd/kgU) fission product gas release data obtained by destructive methods are evaluated with the STAV7 computer code. Finally, slow power ramp experiments conducted at the Studsvik R2 reactor are simulated with the STAV7 code and it is observed that by accounting the contribution of fuel thermal gaseous swelling, the code describes the clad diameter increase due to pellet-clad mechanical interaction under the power bump satisfactorily. (authors)

  13. Aggregate Fission-Product Decay Data Based on ENDF/B-IV and -V.

    1982-10-12

    Version 02 The ENDF/B-IV fission-product files contain neutron cross sections, decay constants, decay energies, and other decay data for 824 important fission products. They also contain fission yields for these fission products produced by one or more fission-neutron energies (14 MeV, fast, and thermal fission). Also, spectral data exist for the most important decay-heat contributors among the 824 nuclides. Because the spectra are based on fission pulses, the libraries have a general utility. The exponentialmore » fits, for example, can be folded into any power (fission) history that can be described analytically or by a histogram representation. The effects of neutron absorption are also treated and approximately accounted for in the methodology.« less

  14. PolyPole-1: An accurate numerical algorithm for intra-granular fission gas release

    NASA Astrophysics Data System (ADS)

    Pizzocri, D.; Rabiti, C.; Luzzi, L.; Barani, T.; Van Uffelen, P.; Pastore, G.

    2016-09-01

    The transport of fission gas from within the fuel grains to the grain boundaries (intra-granular fission gas release) is a fundamental controlling mechanism of fission gas release and gaseous swelling in nuclear fuel. Hence, accurate numerical solution of the corresponding mathematical problem needs to be included in fission gas behaviour models used in fuel performance codes. Under the assumption of equilibrium between trapping and resolution, the process can be described mathematically by a single diffusion equation for the gas atom concentration in a grain. In this paper, we propose a new numerical algorithm (PolyPole-1) to efficiently solve the fission gas diffusion equation in time-varying conditions. The PolyPole-1 algorithm is based on the analytic modal solution of the diffusion equation for constant conditions, combined with polynomial corrective terms that embody the information on the deviation from constant conditions. The new algorithm is verified by comparing the results to a finite difference solution over a large number of randomly generated operation histories. Furthermore, comparison to state-of-the-art algorithms used in fuel performance codes demonstrates that the accuracy of PolyPole-1 is superior to other algorithms, with similar computational effort. Finally, the concept of PolyPole-1 may be extended to the solution of the general problem of intra-granular fission gas diffusion during non-equilibrium trapping and resolution, which will be the subject of future work.

  15. Grain Boundary Percolation Modeling of Fission Gas Release in Oxide Fuels

    SciTech Connect

    Paul C. Millett; Michael R. Tonks; S. B. Biner

    2012-05-01

    We present a new approach to fission gas release modeling in oxide fuels based on grain boundary network percolation. The method accounts for variability in the bubble growth and coalescence rates on individual grain boundaries, and the resulting effect on macroscopic fission gas release. Two-dimensional representa- tions of fuel pellet microstructures are considered, and the resulting gas release rates are compared with traditional two-stage Booth models, which do not account for long-range percolation on grain boundary net- works. The results show that the requirement of percolation of saturated grain boundaries can considerably reduce the total gas release rates, particularly when gas resolution is considered.

  16. Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations

    SciTech Connect

    Gauld, I.C.

    2005-08-12

    U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses that credit fission products in two respects: (1) the microscopic cross sections determine the reactivity worth of the fission products in spent fuel and (2) the cross sections determine the reaction rates during irradiation and thus influence the accuracy of predicted final concentrations of the fission products in the spent fuel. This report evaluates and quantifies the importance of the fission product cross sections in predicting concentrations of fission products proposed for use in burnup credit. The study includes an assessment of the major fission products in burnup credit and their production precursors. Finally, the cross-section importances, or sensitivities, are combined with the importance of each major fission product to the system eigenvalue (k{sub eff}) to determine the net importance of cross sections to k{sub eff}. The importances established the following fission products, listed in descending order of priority, that are most likely to benefit burnup credit when their cross-section uncertainties are reduced: {sup 151}Sm, {sup 103}Rh, {sup 155}Eu, {sup 150}Sm, {sup 152}Sm, {sup 153}Eu, {sup 154}Eu, and {sup 143}Nd.

  17. Venting of fission products and shielding in thermionic nuclear reactor systems

    NASA Technical Reports Server (NTRS)

    Salmi, E. W.

    1972-01-01

    Most thermionic reactors are designed to allow the fission gases to escape out of the emitter. A scheme to allow the fission gases to escape is proposed. Because of the low activity of the fission products, this method should pose no radiation hazards.

  18. Methods to Collect, Compile, and Analyze Observed Short-lived Fission Product Gamma Data

    SciTech Connect

    Finn, Erin C.; Metz, Lori A.; Payne, Rosara F.; Friese, Judah I.; Greenwood, Lawrence R.; Kephart, Jeremy D.; Pierson, Bruce D.; Ellis, Tere A.

    2011-09-29

    A unique set of fission product gamma spectra was collected at short times (4 minutes to 1 week) on various fissionable materials. Gamma spectra were collected from the neutron-induced fission of uranium, neptunium, and plutonium isotopes at thermal, epithermal, fission spectrum, and 14-MeV neutron energies. This report describes the experimental methods used to produce and collect the gamma data, defines the experimental parameters for each method, and demonstrates the consistency of the measurements.

  19. Engineering Report on the Fission Gas Getter Concept

    SciTech Connect

    Ecker, Lynne; Ghose, Sanjit; Gill, Simerjeet; Thallapally, Praveen K.; Strachan, Denis M.

    2012-11-01

    In 2010, the Department of Energy (DOE) requested that a Brookhaven National Laboratory (BNL)-led team research the possibility of using a getter material to reduce the pressure in the plenum region of a light water reactor fuel rod. During the first two years of the project, several candidate materials were identified and tested using a variety of experimental techniques, most with xenon as a simulant for fission products. Earlier promising results for candidate getter materials were found to be incorrect, caused by poor experimental techniques. In May 2012, it had become clear that none of the initial materials had demonstrated the ability to adsorb xenon in the quantities and under the conditions needed. Moreover, the proposed corrective action plan could not meet the schedule needed by the project manager. BNL initiated an internal project review which examined three questions: 1. Which materials, based on accepted materials models, might be capable of absorbing xenon? 2. Which experimental techniques are capable of not only detecting if xenon has been absorbed but also determine by what mechanism and the resulting molecular structure? 3. Are the results from the previous techniques useable now and in the future? As part of the second question, the project review team evaluated the previous experimental technique to determine why incorrect results were reported in early 2012. This engineering report is a summary of the current status of the project review, description of newly recommended experiments and results from feasibility studies at the National Synchrotron Light Source (NSLS).

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

    SciTech Connect

    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 materials 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.

  1. Recent thermochemical research on reactor materials and fission products

    NASA Astrophysics Data System (ADS)

    Cordfunke, E. H. P.; Konings, R. J. M.; Westrum, E. F.

    1989-09-01

    By adiabatic calorimetric measurements from 5 to 350 K and enthalpy increment determinations above the ambient temperature the thermophysical properties of such uranium compounds as UF 3, UC1 3, UBr 3, URu 3, URh 3, UPd 3 and fission product combinations such as RuO 2, RuSe 2, and CsBO 2 have been obtained. In addition, the enthalpies of formation of these substances have been determined by EMF and enthalpy of solution measurements. By combining these measurements the formation properties have been derived as a basis for modeling, critical evaluation and prediction. Some examples of these applications are given.

  2. Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

    SciTech Connect

    Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

    2005-05-24

    The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

  3. Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

    NASA Astrophysics Data System (ADS)

    Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

    2005-05-01

    The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

  4. PURIFICATION OF PLUTONIUM USING A CERIUM PRECIPITATE AS A CARRIER FOR FISSION PRODUCTS

    DOEpatents

    Faris, B.F.; Olson, C.M.

    1961-07-01

    Bismuth phosphate carrier precipitation processes are described for the separation of plutonium from fission products wherein in at least one step bismuth phosphate is precipitated in the presence of hexavalent plutonium thereby carrying a portion of the fission products from soluble plu tonium values. In this step, a cerium phosphate precipitate is formed in conjunction with the bismuth phosphate precipitate, thereby increasing the amount of fission products removed from solution.

  5. Thermal stability of fission gas bubble superlattice in irradiated U–10Mo fuel

    SciTech Connect

    Gan, J.; Keiser, D. D.; Miller, B. D.; Robinson, A. B.; Wachs, D. M.; Meyer, M. K.

    2015-09-01

    To investigate the thermal stability of the fission gas bubble superlattice, a key microstructural feature in both irradiated U-7Mo dispersion and U-10Mo monolithic fuel plates, a FIB-TEM sample of the irradiated U-10Mo fuel with a local fission density of 3.5×1021 fissions/cm3 was used for an in-situ heating TEM experiment. The temperature of the heating holder was raised at a ramp rate of approximately 10 ºC/min up to ~700 ºC, kept at that temperature for about 34 min, continued to 850 ºC with a reduced rate of 5 ºC/min. The result shows a high thermal stability of the fission gas bubble superlattice. The implication of this observation on the fuel microstructural evolution and performance under irradiation is discussed.

  6. A proposed standard on medical isotope production in fission reactors

    SciTech Connect

    Schenter, R. E.; Brown, G. J.; Holden, C. S.

    2006-07-01

    Authors Robert E. Sehenter, Garry Brown and Charles S. Holden argue that a Standard for 'Medical Isotope Production' is needed. Medical isotopes are becoming major components of application for the diagnosis and treatment of all the major diseases including all forms of cancer, heart disease, arthritis, Alzheimer's, among others. Current nuclear data to perform calculations is incomplete, dated or imprecise or otherwise flawed for many isotopes that could have significant applications in medicine. Improved data files will assist computational analyses to design means and methods for improved isotope production techniques in the fission reactor systems. Initial focus of the Standard is expected to be on neutron cross section and branching data for both fast and thermal reactor systems. Evaluated and reviewed tables giving thermal capture cross sections and resonance integrals for the major target and product medical isotopes would be the expected 'first start' for the 'Standard Working Group'. (authors)

  7. Immobilization of fission products in phosphate ceramic waste forms

    SciTech Connect

    Singh, D.; Wagh, A.

    1997-10-01

    Chemically bonded phosphate ceramics (CBPCs) have several advantages that make them ideal candidates for containing radioactive and hazardous wastes. In general, phosphates have high solid-solution capacities for incorporating radionuclides, as evidenced by several phosphates (e.g., monazites and apatites) that are natural analogs of radioactive and rare-earth elements. The phosphates have high radiation stability, are refractory, and will not degrade in the presence of internal heating by fission products. Dense and hard CBPCs can be fabricated inexpensively and at low temperature by acid-base reactions between an inorganic oxide/hydroxide powder and either phosphoric acid or an acid-phosphate solution. The resulting phosphates are extremely insoluble in aqueous media and have excellent long-term durability. CBPCs offer the dual stabilization mechanisms of chemical fixation and physical encapsulation, resulting in superior waste forms. The goal of this task is develop and demonstrate the feasibility of CBPCs for S/S of wastes containing fission products. The focus of this work is to develop a low-temperature CBPC immobilization system for eluted {sup 99}Tc wastes from sorption processes.

  8. Preliminary investigation of a technique to separate fission noble metals from fission product mixtures

    SciTech Connect

    Mellinger, G.B.; Jensen, G.A.

    1982-08-01

    A variation of the gold-ore fire assay technique was examined as a method for recovering Pd, Rh and Ru from fission products. The mixture of fission product oxides is combined with glass-forming chemicals, a metal oxide such as PbO (scavenging agent), and a reducing agent such as charcoal. When this mixture is melted, a metal button is formed which extracts the noble metals. The remainder cools to form a glass for nuclear waste storage. Recovery depended only on reduction of the scavenger oxide to metal. When such reduction was achieved, no difference in noble metal recovery efficiency was found among the scavengers studied (PbO, SnO, CuO, Bi/sub 2/O/sub 3/, Sb/sub 2/O/sub 3/). Not all reducing agents studied, however, were able to reduce all scavenger oxides to metal. Only graphite would reduce SnO and CuO and allow noble metal recovery. The scavenger oxides Sb/sub 2/O/sub 3/, Bi/sub 2/O/sub 3/, and PbO, however, were reduced by all of the reducing agents tested. Similar noble metal recovery was found with each. Lead oxide was found to be the most promising of the potential scavengers. It was reduced by all of the reducing agents tested, and its higher density may facilitate the separation. Use of lead oxide also appeared to have no deterimental effect on the glass quality. Charcoal was identified as the preferred reducing agent. As long as a separable metal phase was formed in the melt, noble metal recovery was not dependent on the amount of reducing agent and scavenger oxide. High glass viscosities inhibited separation of the molten scavenger, while low viscosities allowed volatile loss of RuO/sub 4/. A viscosity of approx. 20 poise at the processing temperature offered a good compromise between scavenger separation and Ru recovery. Glasses in which PbO was used as the scavenging agent were homogeneous in appearance. Resistance to leaching was close to that of certain waste glasses reported in the literature. 12 figures. 7 tables.

  9. Assessment of selected fission products in the Savannah River Site environment

    SciTech Connect

    Carlton, W.H.; Denham, M.

    1997-04-01

    Most of the radioactivity produced by the operation of a nuclear reactor results from the fission process, during which the nucleus of a fissionable atom (such as 235U) splits into two or more nuclei, which typically are radioactive. The Radionuclide Assessment Program (RAP) has reported on fission products cesium, strontium, iodine, and technetium. Many other radionuclides are produced by the fission process. Releases of several additional fission products that result in dose to the offsite population are discussed in this publication. They are 95Zr, 95Nb, 103Ru, 106Ru, 141Ce, and 144Ce. This document will discuss the production, release, migration, and dose to humans for each of these selected fission products.

  10. On-site gamma-ray spectroscopic measurements of fission gas release in irradiated nuclear fuel.

    PubMed

    Matsson, I; Grapengiesser, B; Andersson, B

    2007-01-01

    An experimental, non-destructive in-pool, method for measuring fission gas release (FGR) in irradiated nuclear fuel has been developed. Using the method, a significant number of experiments have been performed in-pool at several nuclear power plants of the BWR type. The method utilises the 514 keV gamma-radiation from the gaseous fission product (85)Kr captured in the fuel rod plenum volume. A submergible measuring device (LOKET) consisting of an HPGe-detector and a collimator system was utilised allowing for single rod measurements on virtually all types of BWR fuel. A FGR database covering a wide range of burn-ups (up to average rod burn-up well above 60 MWd/kgU), irradiation history, fuel rod position in cross section and fuel designs has been compiled and used for computer code benchmarking, fuel performance analysis and feedback to reactor operators. Measurements clearly indicate the low FGR in more modern fuel designs in comparison to older fuel types.

  11. Experimental Measurements of Short-Lived Fission Products from Uranium, Neptunium, Plutonium and Americium

    SciTech Connect

    Metz, Lori A.; Payne, Rosara F.; Friese, Judah I.; Greenwood, Lawrence R.; Kephart, Jeremy D.; Pierson, Bruce D.

    2009-11-01

    Fission yields are especially well characterized for long-lived fission products. Modeling techniques incorporate numerous assumptions and can be used to deduce information about the distribution of short-lived fission products. This work is an attempt to gather experimental (model-independent) data on the short-lived fission products. Fissile isotopes of uranium, neptunium, plutonium and americium were irradiated under pulse conditions at the Washington State University 1 MW TRIGA reactor to achieve ~108 fissions. The samples were placed on a HPGe (high purity germanium) detector to begin counting in less than 3 minutes post irradiation. The samples were counted for various time intervals ranging from 5 minutes to 1 hour. The data was then analyzed to determine which radionuclides could be quantified and compared to the published fission yield data.

  12. Fission Product Yields from Fission Spectrum n+{sup 239}Pu for ENDF/B-VII.1

    SciTech Connect

    Chadwick, M.B.; Kawano, T.; Barr, D.W.; Mac Innes, M.R.; Kahler, A.C.; Graves, T.; Selby, H.; Burns, C.J.; Inkret, W.C.; Keksis, A.L.; Lestone, J.P.; Sierk, A.J.; Talou, P.

    2010-12-15

    We describe a new cumulated fission product yield (FPY) evaluation for fission spectrum neutrons on plutonium that updates the ENDF/B-VI evaluation by England and Rider, for the forthcoming ENDF/B-VII.1 database release. We focus on FPs that are needed for high accuracy burnup assessments; that is, for inferring the number of fissions in a neutron environment. Los Alamos conducted an experiment in the 1970s in the Bigten fast critical assembly to determine fission product yields as part of the Interlaboratory Reaction Rate (ILRR) collaboration, and this has defined the Laboratory's fission standard to this day. Our evaluation includes use of the LANL-ILRR measurements (not previously available to evaluators) as well as other Laboratory FPY measurements published in the literature, especially the high-accuracy mass spectrometry data from Maeck and others. Because the measurement database for some of the FPs is small - especially for {sup 99}Mo - we use a meta-analysis that incorporates insights from other accurately-measured benchmark FP data, using R-value ratio measurements. The meta-analysis supports the FP measurements from the LANL-ILRR experiment. Differences between our new evaluations and ENDF/B-VI are small for some FPs (less than 1-2%-relative for {sup 95}Zr, {sup 140}Ba, {sup 144}Ce), but are larger for {sup 99}Mo (4%-relative) and {sup 147}Nd (5%-relative, at 1.5 MeV) respectively. We present evidence for an incident neutron energy dependence to the {sup 147}Nd fission product yield that accounts for observed differences in the FPY at a few-hundred keV average energy in fast reactors versus measurements made at higher average neutron energies in Los Alamos' fast critical assemblies. Accounting for such FPY neutron energy dependencies is important if one wants to reach a goal of determining the number of fissions to accuracies of 1-2%. An evaluation of the energy-dependence of fission product yields is given for all A values based on systematical trends

  13. SEPARATION OF PLUTONIUM VALUES FROM URANIUM AND FISSION PRODUCT VALUES

    DOEpatents

    Maddock, A.G.; Booth, A.H.

    1960-09-13

    Separation of plutonium present in small amounts from neutron irradiated uranium by making use of the phenomenon of chemisorption is described. Plutonium in the tetravalent state is chemically absorbed on a fluoride in solid form. The steps for the separation comprise dissolving the irradiated uranium in nitric acid, oxidizing the plutonium in the resulting solution to the hexavalent state, adding to the solution a soluble calcium salt which by the common ion effect inhibits dissolution of the fluoride by the solution, passing the solution through a bed or column of subdivided calcium fluoride which has been sintered to about 8OO deg C to remove the chemisorbable fission products, reducing the plutonium in the solution thus obtained to the tetravalent state, and again passing the solution through a similar bed or column of calcium fluoride to selectively absorb the plutonium, which may then be recovered by treating the calcium fluoride with a solution of ammonium oxalate.

  14. SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS FROM NEUTRON- BOMBARDED URANIUM

    DOEpatents

    Martin, A.E.; Johnson, I.; Burris, L. Jr.; Winsch, I.O.; Feder, H.M.

    1962-11-13

    A process is given for removing plutonium and/or fission products from uranium fuel. The fuel is dissolved in molten zinc--magnesium (10 to 18% Mg) alloy, more magnesium is added to obtain eutectic composition whereby uranium precipitates, and the uranium are separated from the Plutoniumand fission-product- containing eutectic. (AEC)

  15. Target and method for the production of fission product molybdenum-99

    DOEpatents

    Vandegrift, George F.; Vissers, Donald R.; Marshall, Simon L.; Varma, Ravi

    1989-01-01

    A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm.sup.2 of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99.

  16. Target and method for the production of fission product molybdenum-99

    DOEpatents

    Vandegrift, G.F.; Vissers, D.R.; Marshall, S.L.; Varma, R.

    1987-10-26

    A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm/sup 2/ of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99. 2 figs.

  17. Assessment of fission product yields data needs in nuclear reactor applications

    SciTech Connect

    Kern, K.; Becker, M.; Broeders, C.

    2012-07-01

    Studies on the build-up of fission products in fast reactors have been performed, with particular emphasis on the effects related to the physics of the nuclear fission process. Fission product yields, which are required for burn-up calculations, depend on the proton and neutron number of the target nucleus as well as on the incident neutron energy. Evaluated nuclear data on fission product yields are available for all relevant target nuclides in reactor applications. However, the description of their energy dependence in evaluated data is still rather rudimentary, which is due to the lack of experimental fast fission data and reliable physical models. Additionally, physics studies of evaluated JEFF-3.1.1 fission yields data have shown potential improvements, especially for various fast fission data sets of this evaluation. In recent years, important progress in the understanding of the fission process has been made, and advanced model codes are currently being developed. This paper deals with the semi-empirical approach to the description of the fission process, which is used in the GEF code being developed by K.-H. Schmidt and B. Jurado on behalf of the OECD Nuclear Energy Agency, and with results from the corresponding author's diploma thesis. An extended version of the GEF code, supporting the calculation of spectrum weighted fission product yields, has been developed. It has been applied to the calculation of fission product yields in the fission rate spectra of a MOX fuelled sodium-cooled fast reactor. Important results are compared to JEFF-3.1.1 data and discussed in this paper. (authors)

  18. Neutron Cross Section Covariances for Structural Materials and Fission Products

    SciTech Connect

    Hoblit, S.; Hoblit,S.; Cho,Y.-S.; Herman,M.; Mattoon,C.M.; Mughabghab,S.F.; Oblozinsky,P.; Pigni,M.T.; Sonzogni,A.A.

    2011-12-01

    We describe neutron cross section covariances for 78 structural materials and fission products produced for the new US evaluated nuclear reaction library ENDF/B-VII.1. Neutron incident energies cover full range from 10{sup -5} eV to 20 MeV and covariances are primarily provided for capture, elastic and inelastic scattering as well as (n,2n). The list of materials follows priorities defined by the Advanced Fuel Cycle Initiative, the major application being data adjustment for advanced fast reactor systems. Thus, in addition to 28 structural materials and 49 fission products, the list includes also {sup 23}Na which is important fast reactor coolant. Due to extensive amount of materials, we adopted a variety of methodologies depending on the priority of a specific material. In the resolved resonance region we primarily used resonance parameter uncertainties given in Atlas of Neutron Resonances and either applied the kernel approximation to propagate these uncertainties into cross section uncertainties or resorted to simplified estimates based on integral quantities. For several priority materials we adopted MF32 covariances produced by SAMMY at ORNL, modified by us by adding MF33 covariances to account for systematic uncertainties. In the fast neutron region we resorted to three methods. The most sophisticated was EMPIRE-KALMAN method which combines experimental data from EXFOR library with nuclear reaction modeling and least-squares fitting. The two other methods used simplified estimates, either based on the propagation of nuclear reaction model parameter uncertainties or on a dispersion analysis of central cross section values in recent evaluated data files. All covariances were subject to quality assurance procedures adopted recently by CSEWG. In addition, tools were developed to allow inspection of processed covariances and computed integral quantities, and for comparing these values to data from the Atlas and the astrophysics database KADoNiS.

  19. First-principles study of the stability of fission products in uranium monocarbide

    NASA Astrophysics Data System (ADS)

    Bévillon, Émile; Ducher, Roland; Barrachin, Marc; Dubourg, Roland

    2012-07-01

    The incorporation and stability of fission products in uranium monocarbide are studied by means of Density Functional Theory using the generalized gradient approximation and projector-augmented waves method. The computations are performed considering incorporation sites of UC, such as the U, C and interstitial sites, and Schottky defects. The computed incorporation energies are discussed on the basis of the atomic size of the fission products, their chemical environment and the electronic structure. These energies show that all the studied fission products would preferentially occupy the U site. However, incorporation energies do not provide any further information on the fission product location in the case of unavailability of the sites which is why the concept of solution energies is also used. The solution energies obtained confirm that all the fission products are expected to be more stable on a U site of a single uranium vacancy or within a non-bound Schottky defect in equilibrium conditions.

  20. Plutonium and surrogate fission products in a composite ceramic waste form.

    SciTech Connect

    Esh, D. W.; Frank, S. M.; Goff, K. M.; Johnson, S. G.; Moschetti, T. L.; O'Holleran, T.

    1999-05-19

    Argonne National Laboratory is developing a ceramic waste form to immobilize salt containing fission products and transuranic elements. Preliminary results have been presented for ceramic waste forms containing surrogate fission products such as cesium and the lanthanides. In this work results from scanning electron microscopy/energy dispersive spectroscopy and x-ray diffraction are presented in greater detail for ceramic waste forms containing surrogate fission products. Additionally, results for waste forms containing plutonium and surrogate fission products are presented. Most of the surrogate fission products appear to be silicates or aluminosilicates whereas the plutonium is usually found in an oxide form. There is also evidence for the presence of plutonium within the sodalite phase although the chemical speciation of the plutonium is not known.

  1. Derivation of effective fission gas diffusivities in UO2 from lower length scale simulations and implementation of fission gas diffusion models in BISON

    SciTech Connect

    Andersson, Anders David Ragnar; Pastore, Giovanni; Liu, Xiang-Yang; Perriot, Romain Thibault; Tonks, Michael; Stanek, Christopher Richard

    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 UO2 were derived for both intrinsic conditions and under irradiation. The importance of the large XeU3O cluster (a Xe atom in a uranium + oxygen vacancy trap site with two bound uranium vacancies) is emphasized, which is a consequence 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.

  2. Radiation re-solution of fission gas in non-oxide nuclear fuel

    SciTech Connect

    Matthews, Christopher; Schwen, Daniel; Klein, Andrew C.

    2015-02-01

    Renewed interest in fast nuclear reactors is creating a need for better understanding of fission gas bubble behavior in non-oxide fuels to support very long fuel lifetimes. Collisions between fission fragments and their subsequent cascades can knock fission gas atoms out of bubbles and back into the fuel lattice. We showed that these collisions can be treated as using the so-called ‘‘homogenous’’ atom-by-atom re-solution theory and calculated using the Binary Collision Approximation code 3DOT. The calculations showed that there is a decrease in the re-solution parameter as bubble radius increases until about 50 nm, at which the re-solution parameter stays nearly constant. Furthermore, our model shows ion cascades created in the fuel result in many more implanted fission gas atoms than collisions directly with fission fragments. This calculated re-solution parameter can be used to find a re-solution rate for future bubble simulations.

  3. Measurement of fission products yields in the quasi-mono-energetic neutron-induced fission of 232Th

    NASA Astrophysics Data System (ADS)

    Naik, H.; Mukherji, Sadhana; Suryanarayana, S. V.; Jagadeesan, K. C.; Thakare, S. V.; Sharma, S. C.

    2016-08-01

    The cumulative yields of various fission products in the 232Th(n, f) reaction at average neutron energies of 5.42, 7.75, 9.35 and 12.53 MeV have been determined by using an off-line γ-ray spectrometric technique. The neutron beam was produced from the 7Li(p, n) reaction by using the proton energies of 7.8, 12, 16 and 20 MeV. The mass chain yields were obtained from the cumulative fission yields by using the charge distribution correction of medium energy fission. The fine structure in the mass yield distribution was interpreted from the point of nuclear structure effect. On the other hand, the higher yield around mass number 133-134 and 143-144 as well as their complementary products were explained based on the standard I and standard II asymmetric mode of fission. From the mass yield data, the average value of light mass (), heavy mass (), the average number of neutrons (< ν >) and the peak-to-valley (P / V) ratios at different neutron energies of present work and literature data were obtained in the 232Th(n, f) reaction. The different parameters of the mass yield distribution in the 232Th(n, f) reaction were compared with the similar data in the 232Th(γ, f) reaction at comparable excitation energy and a surprising difference was observed.

  4. Early fission-gas behavior in oxide fuel: escape vs trapping

    SciTech Connect

    Cordoliani, V.; Olander, D.

    2007-07-01

    A Monte Carlo code has been developed to investigate the behavior of fission-gas on the grain boundaries of UO{sub 2} fuel in the early stages of irradiation. Gas atoms arriving at the grain boundary from the grain interior undergo a random walk until they meet one of four fates: 1. nucleate a bubble by jumping next to another atom; 2. trapped by an existing bubble; 3. escape at the termination of the grain boundary at a free surface; 4. re-solved by collision with a fission fragment intersecting the grain boundary. At issue is the relative importance of escape (release) and trapping by intergranular bubbles. The switch of the dominant mechanism occurs at a critical burnup when the rate of trapping equals the arrival rate of gas atoms on the grain boundary. The critical burnup depends on fission rate and temperature, but is always less than 0.003 MWd/kgU. A very low rate of fission-gas release persists beyond the critical burnup; gas atoms from the grain interior that arrive at the grain boundary immediately adjacent to the escape surface have a small but finite probability of avoiding trapping and instead are released. The code calculates the intergranular bubble-size distribution with attention paid to the effect of re-solution. (authors)

  5. HTR 2014 Paper - Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

    SciTech Connect

    Blaise P. Collin

    2001-10-01

    Safety tests were conducted on fourteen 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 the safety tests, and the predicted values were compared with experimental results. Preliminary comparisons between PARFUME predictions and post-irradiation examination (PIE) results of the safety tests show an overall over-prediction of the fractional release of these fission products, which is largely attributed to an over-estimation of the diffusivities used in the modeling of fission product transport in TRISO-coated particles. Correction factors to these diffusivities were assessed for silver and cesium in order to enable a better match between the modeling predictions and the safety testing results. 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. In the case of krypton, all the coating layers are partly retentive and the available data did not allow to determine their respective retention powers, hence preventing to derive any correction factors.

  6. Instabilities in fissioning plasmas as applied to the gas-core nuclear rocket-engine

    NASA Technical Reports Server (NTRS)

    1973-01-01

    The compressional wave spectrum excited in a fissioning uranium plasma confined in a cavity such as a gas cored nuclear reactor, is studied. Computer results are presented that solve the fluid equations for this problem including the effects of spatial gradients, nonlinearities, and neutron density gradients in the reactor. Typically the asymptotic fluctuation level for the plasma pressure is of order 1 percent.

  7. Implementation of a Thermodynamic Solver within a Computer Program for Calculating Fission-Product Release Fractions

    NASA Astrophysics Data System (ADS)

    Barber, Duncan Henry

    During some postulated accidents at nuclear power stations, fuel cooling may be impaired. In such cases, the fuel heats up and the subsequent increased fission-gas release from the fuel to the gap may result in fuel sheath failure. After fuel sheath failure, the barrier between the coolant and the fuel pellets is lost or impaired, gases and vapours from the fuel-to-sheath gap and other open voids in the fuel pellets can be vented. Gases and steam from the coolant can enter the broken fuel sheath and interact with the fuel pellet surfaces and the fission-product inclusion on the fuel surface (including material at the surface of the fuel matrix). The chemistry of this interaction is an important mechanism to model in order to assess fission-product releases from fuel. Starting in 1995, the computer program SOURCE 2.0 was developed by the Canadian nuclear industry to model fission-product release from fuel during such accidents. SOURCE 2.0 has employed an early thermochemical model of irradiated uranium dioxide fuel developed at the Royal Military College of Canada. To overcome the limitations of computers of that time, the implementation of the RMC model employed lookup tables to pre-calculated equilibrium conditions. In the intervening years, the RMC model has been improved, the power of computers has increased significantly, and thermodynamic subroutine libraries have become available. This thesis is the result of extensive work based on these three factors. A prototype computer program (referred to as SC11) has been developed that uses a thermodynamic subroutine library to calculate thermodynamic equilibria using Gibbs energy minimization. The Gibbs energy minimization requires the system temperature (T) and pressure (P), and the inventory of chemical elements (n) in the system. In order to calculate the inventory of chemical elements in the fuel, the list of nuclides and nuclear isomers modelled in SC11 had to be expanded from the list used by SOURCE 2.0. A

  8. High-Resolution Correlated Fission Product Measurements of 235U (nth , f) with SPIDER

    NASA Astrophysics Data System (ADS)

    Shields, Dan; Spider Team

    2015-10-01

    The SPIDER detector (SPectrometer for Ion DEtermination in fission Research) has obtained high-resolution, moderate-efficiency, correlated fission product data needed for many applications including the modeling of next generation nuclear reactors, stockpile stewardship, and the fundamental understanding of the fission process. SPIDER simultaneously measures velocity and energy of both fission products to calculate fission product yields (FPYs), neutron multiplicity (ν), and total kinetic energy (TKE). These data will be some of the first of their kind available to nuclear data evaluations. An overview of the SPIDER detector, analytical method, and preliminary results for 235U (nth , f) will be presented. LA-UR-15-20130 This work benefited from the use of the LANSCE accelerator facility and was performed under the auspices of the US Department of Energy by Los Alamos Security, LLC under Contract DE-AC52-06NA25396.

  9. Identifying and quantifying short-lived fission products from thermal fission of HEU using portable HPGe detectors

    SciTech Connect

    Pierson, Bruce D.; Finn, Erin C.; Friese, Judah I.; Greenwood, Lawrence R.; Kephart, Jeremy D.; Kephart, Rosara F.; Metz, Lori A.

    2013-03-01

    Due to the emerging potential for trafficking of special nuclear material, research programs are investigating current capabilities of commercially available portable gamma ray detection systems. Presented in this paper are the results of three different portable high-purity germanium (HPGe) detectors used to identify short-lived fission products generated from thermal neutron interrogation of small samples of highly enriched uranium. Samples were irradiated at the Washington State University (WSU) Nuclear Radiation Center’s 1MW TRIGA reactor. The three portable, HPGe detectors used were the ORTEC MicroDetective, the ORTEC Detective, and the Canberra Falcon. Canberra’s GENIE-2000 software was used to analyze the spectral data collected from each detector. Ultimately, these three portable detectors were able to identify a large range of fission products showing potential for material discrimination.

  10. Radiation Damage and Fission Product Release in Zirconium Nitride

    SciTech Connect

    Egeland, Gerald W.

    2005-08-29

    Zirconium nitride is a material of interest to the AFCI program due to some of its particular properties, such as its high melting point, strength and thermal conductivity. It is to be used as an inert matrix or diluent with a nuclear fuel based on transuranics. As such, it must sustain not only high temperatures, but also continuous irradiation from fission and decay products. This study addresses the issues of irradiation damage and fission product retention in zirconium nitride through an assessment of defects that are produced, how they react, and how predictions can be made as to the overall lifespan of the complete nuclear fuel package. Ion irradiation experiments are a standard method for producing radiation damage to a surface for observation. Cryogenic irradiations are performed to produce the maximum accumulation of defects, while elevated temperature irradiations may be used to allow defects to migrate and react to form clusters and loops. Cross-sectional transmission electron microscopy and grazing-incidence x-ray diffractometry were used in evaluating the effects that irradiation has on the crystal structure and microstructure of the material. Other techniques were employed to evaluate physical effects, such as nanoindentation and helium release measurements. Results of the irradiations showed that, at cryogenic temperatures, ZrN withstood over 200 displacements per atom without amorphization. No significant change to the lattice or microstructure was observed. At elevated temperatures, the large amount of damage showed mobility, but did not anneal significantly. Defect clustering was possibly observed, yet the size was too small to evaluate, and bubble formation was not observed. Defects, specifically nitrogen vacancies, affect the mechanical behavior of ZrN dramatically. Current and previous work on dislocations shows a distinct change in slip plane, which is evidence of the bonding characteristics. The stacking-fault energy changes dramatically with

  11. Critical temperature for the nuclear liquid-gas phase transition (from multifragmentation and fission)

    SciTech Connect

    Karnaukhov, V. A.; Oeschler, H.; Budzanowski, A.; Avdeyev, S. P.; Botvina, A. S.; Cherepanov, E. A.; Karcz, W.; Kirakosyan, V. V.; Rukoyatkin, P. A.; Skwirczynska, I.; Norbeck, E.

    2008-12-15

    Critical temperature T{sub c} for the nuclear liquid-gas phase transition is estimated from both the multifragmentation and fission data. In the first case, the critical temperature is obtained by analysis of the intermediate-mass-fragment yields in p(8.1 GeV) + Au collisions within the statistical model of multifragmentation. In the second case, the experimental fission probability for excited {sup 188}Os is compared with the calculated one with T{sub c} as a free parameter. It is concluded for both cases that the critical temperature is higher than 15 MeV.

  12. Transport of symmetric mass region fission products at the Oklo natural reactors

    NASA Astrophysics Data System (ADS)

    Loss, R. D.; Rosman, K. J. R.; de Laeter, J. R.

    1984-05-01

    The isotopic composition of Pd, Ag, Cd and Te has been measured by solid source mass spectrometry for four samples from reactor zones 2, 3-4, 5-6 and 7, and from four host rock samples external to the reactor zones from the Oklo mine site. The concentrations of these elements have also been determined in the eight samples using the stable isotope dilution technique. Cumulative fission yields have been derived from the reactor zone samples after correcting where necessary for the terrestrial component of the element concerned. It has been shown that fission-produced Pd and Te are retained almost in their entirety in the uraninite reactor zone samples, whereas a significant fraction of fission-produced Ag and Cd have migrated from the reactor zones. Fission product Cd is observed in the host rock samples, whereas no strong evidence of fission-produced Ag could be found. Thus the fission-produced Ag which has migrated from the reactor zones has not been retained in the four host rock samples analysed, although the presence of fission product Ag may be masked by the presence of natural Ag. It is possible that the fission product Ag has been retained in the Oklo mine-site, and further host rock samples will be studied to evaluate this possibility. The implications of these results to the storage of radioactive wastes in natural geological repositories is discussed.

  13. Background and Derivation of ANS-5.4 Standard Fission Product Release Model

    SciTech Connect

    Beyer, Carl E.; Turnbull, Andrew J.

    2010-01-29

    This background report describes the technical basis for the newly proposed American Nuclear Society (ANS) 5.4 standard, Methods for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuels. The proposed ANS 5.4 standard provides a methodology for determining the radioactive fission product releases from the fuel for use in assessing radiological consequences of postulated accidents that do not involve abrupt power transients. When coupled with isotopic yields, this method establishes the 'gap activity,' which is the inventory of volatile fission products that are released from the fuel rod if the cladding are breached.

  14. SEPARATION OF PLUTONIUM FROM FISSION PRODUCTS BY A COLLOID REMOVAL PROCESS

    DOEpatents

    Schubert, J.

    1960-05-24

    A method is given for separating plutonium from uranium fission products. An acidic aqueous solution containing plutonium and uranium fission products is subjected to a process for separating ionic values from colloidal matter suspended therein while the pH of the solution is maintained between 0 and 4. Certain of the fission products, and in particular, zirconium, niobium, lanthanum, and barium are in a colloidal state within this pH range, while plutonium remains in an ionic form, Dialysis, ultracontrifugation, and ultrafiltration are suitable methods of separating plutonium ions from the colloids.

  15. Deep Atomic Binding (DAB) Approach in Interpretation of Fission Products Behavior in Terrestrial and Water Ecosystems

    SciTech Connect

    Ajlouni, Abdul-Wali M.S.

    2006-07-01

    A large number of studies and models were established to explain the fission products (FP) behavior within terrestrial and water ecosystems, but a number of behaviors were non understandable, which always attributed to unknown reasons. According to DAB hypothesis, almost all fission products behaviors in terrestrial and water ecosystems could be interpreted in a wide coincidence. The gab between former models predictions, and field behavior of fission products after accidents like Chernobyl have been explained. DAB represents a tool to reduce radio-phobia as well as radiation protection expenses. (author)

  16. Development of a Gas Filled Magnet spectrometer coupled with the Lohengrin spectrometer for fission study

    NASA Astrophysics Data System (ADS)

    Kessedjian, G.; Chebboubi, A.; Faust, H.; Köster, U.; Materna, T.; Sage, C.; Serot, O.

    2013-03-01

    The accurate knowledge of the fission of actinides is necessary for studies of innovative nuclear reactor concepts. The fission yields have a direct influence on the evaluation of the fuel inventory or the reactor residual power after shutdown. A collaboration between the ILL, LPSC and CEA has developed a measurement program on fission fragment distributions at ILL in order to measure the isotopic and isomeric yields. The method is illustrated using the 233U(n,f)98Y reaction. However, the extracted beam from the Lohengrin spectrometer is not isobaric ions which limits the low yield measurements. Presently, the coupling of the Lohengrin spectrometer with a Gas Filled Magnet (GFM) is studied at the ILL in order to define and validate the enhanced purification of the extracted beam. This work will present the results of the spectrometer characterisation, along with a comparison with a dedicated Monte Carlo simulation especially developed for this purpose.

  17. Data summary report for fission product release Test VI-7

    SciTech Connect

    Osborne, M.F.; Lorentz, R.A.; Travis, J.R.; Collins, J.L.; Webster, C.S.

    1995-05-01

    Test VI-7 was the final 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 Monticello boiling water reactor (BWR). The fuel had experienced a burnup of {approximately}-40 Mwd/kg U. It was heated in an induction furnace for successive 20-min periods at 2000 and 2300 K in a moist air-helium atmosphere. Integral releases were 69% for {sup 85}Kr, 52% for {sup 125}Sb, 71% for both {sup 134}Cs and {sup 137}Cs, and 0.04% for {sup 154}Eu. For the non-gamma-emitting species, release values for 42% for I, 4.1% for Ba, 5.3% for Mo, and 1.2% for Sr were determined. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.89 g, with 37% being collected on the thermal gradient tubes and 63% downstream on filters. Posttest examination of the fuel specimen indicated that most of the cladding was completely oxidized to ZrO{sub 2}, but that oxidation was not quite complete at the upper end. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL-Booth Model.

  18. Fission products from the damaged Fukushima reactor observed in Hungary.

    PubMed

    Bihari, Árpád; Dezső, Zoltán; Bujtás, Tibor; Manga, László; Lencsés, András; Dombóvári, Péter; Csige, István; Ranga, Tibor; Mogyorósi, Magdolna; Veres, Mihály

    2014-01-01

    Fission products, especially (131)I, (134)Cs and (137)Cs, from the damaged Fukushima Dai-ichi nuclear power plant (NPP) were detected in many places worldwide shortly after the accident caused by natural disaster. To observe the spatial and temporal variation of these isotopes in Hungary, aerosol samples were collected at five locations from late March to early May 2011: Institute of Nuclear Research, Hungarian Academy of Sciences (ATOMKI, Debrecen, East Hungary), Paks NPP (Paks, South-Central Hungary) as well as at the vicinity of Aggtelek (Northeast Hungary), Tapolca (West Hungary) and Bátaapáti (Southwest Hungary) settlements. In addition to the aerosol samples, dry/wet fallout samples were collected at ATOMKI, and airborne elemental iodine and organic iodide samples were collected at Paks NPP. The peak in the activity concentration of airborne (131)I was observed around 30 March (1-3 mBq m(-3) both in aerosol samples and gaseous iodine traps) with a slow decline afterwards. Aerosol samples of several hundred cubic metres of air showed (134)Cs and (137)Cs in detectable amounts along with (131)I. The decay-corrected inventory of (131)I fallout at ATOMKI was 2.1±0.1 Bq m(-2) at maximum in the observation period. Dose-rate contribution calculations show that the radiological impact of this event at Hungarian locations was of no considerable concern.

  19. Prompt γ-ray production in neutron-induced fission of 239Pu

    NASA Astrophysics Data System (ADS)

    Ullmann, J. L.; Bond, E. M.; Bredeweg, T. A.; Couture, A.; Haight, R. C.; Jandel, M.; Kawano, T.; Lee, H. Y.; O'Donnell, J. M.; Hayes, A. C.; Stetcu, I.; Taddeucci, T. N.; Talou, P.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Chyzh, A.; Gostic, J.; Henderson, R.; Kwan, E.; Wu, C. Y.

    2013-04-01

    Background: The prompt gamma-ray spectrum from fission is important for understanding the physics of nuclear fission, and also in applications involving fission. Relatively few measurements of the prompt gamma spectrum from 239Pu(n,f) have been published.Purpose: This experiment measured the multiplicity, individual gamma energy spectrum, and total gamma energy spectrum of prompt fission gamma rays from 239Pu(n,f) in the neutron energy range from thermal to 30 keV, to test models of fission and to provide information for applications.Method: Gamma rays from neutron-induced fission of 239Pu were measured using the DANCE gamma-ray calorimeter. Fission events were tagged by detecting fission products in a parallel-plate avalanche counter in the center of DANCE. The measurements were corrected for detector response using a geant4 model of DANCE. A detailed analysis for the gamma rays from the 1+ resonance complex at 10.93 eV is presented.Results: A six-parameter analytical parametrization of the fission gamma-ray spectrum was obtained. A Monte Carlo Hauser-Feshbach calculation provided good general agreement with the data, but some differences remain to be resolved.Conclusions: An analytic parametrization can be made of the gamma-ray multiplicity, energy distribution, and total-energy distribution for the prompt gamma rays following neutron-induced fission of 239Pu. This parametrization may be useful for applications. Modern Monte Carlo Hauser-Feshbach calculations can do a good job of calculating the fission gamma-ray emission spectrum, although some details remain to be understood.

  20. Determination of {sup 140}La fission product interference factor for INAA

    SciTech Connect

    Ribeiro Jr, Iberê S.; Genezini, Frederico A.; Saiki, Mitiko; Zahn, Guilherme S.

    2014-11-11

    Instrumental Neutron Activation Analysis (INAA) is a technique widely used to determine the concentration of several elements in several kinds of matrices. However if the sample of interest has higher relative uranium concentration the obtained results can be interfered by the uranium fission products. One of these cases that is affected by interference due to U fission is the {sup 140}La, because this radioisotope used in INAA for the determination of concentration the La is also produced by the {sup −}β of {sup 140}Ba, an uranium fission product. The {sup 140}La interference factor was studied in this work and a factor to describe its time dependence was obtained.

  1. Investigation of inconsistent ENDF/B-VII.1 independent and cumulative fission product yields with proposed revisions

    SciTech Connect

    Pigni, Marco T; Francis, Matthew W; Gauld, Ian C

    2015-01-01

    A recent implementation of ENDF/B-VII. independent fission product yields and nuclear decay data identified inconsistencies in the data caused by the use of updated nuclear scheme in the decay sub-library that is not reflected in legacy fission product yield data. Recent changes in the decay data sub-library, particularly the delayed neutron branching fractions, result in calculated fission product concentrations that are incompatible with the cumulative fission yields in the library, and also with experimental measurements. A comprehensive set of independent fission product yields was generated for thermal and fission spectrum neutron induced fission for 235,238U and 239,241Pu in order to provide a preliminary assessment of the updated fission product yield data consistency. These updated independent fission product yields were utilized in the ORIGEN code to evaluate the calculated fission product inventories with experimentally measured inventories, with particular attention given to the noble gases. An important outcome of this work is the development of fission product yield covariance data necessary for fission product uncertainty quantification. The evaluation methodology combines a sequential Bayesian method to guarantee consistency between independent and cumulative yields along with the physical constraints on the independent yields. This work was motivated to improve the performance of the ENDF/B-VII.1 library in the case of stable and long-lived cumulative yields due to the inconsistency of ENDF/B-VII.1 fission p;roduct yield and decay data sub-libraries. The revised fission product yields and the new covariance data are proposed as a revision to the fission yield data currently in ENDF/B-VII.1.

  2. Investigation of Inconsistent ENDF/B-VII.1 Independent and Cumulative Fission Product Yields with Proposed Revisions

    SciTech Connect

    Pigni, M.T. Francis, M.W.; Gauld, I.C.

    2015-01-15

    A recent implementation of ENDF/B-VII.1 independent fission product yields and nuclear decay data identified inconsistencies in the data caused by the use of updated nuclear schemes in the decay sub-library that are not reflected in legacy fission product yield data. Recent changes in the decay data sub-library, particularly the delayed neutron branching fractions, result in calculated fission product concentrations that do not agree with the cumulative fission yields in the library as well as with experimental measurements. To address these issues, a comprehensive set of independent fission product yields was generated for thermal and fission spectrum neutron-induced fission for {sup 235,238}U and {sup 239,241}Pu in order to provide a preliminary assessment of the updated fission product yield data consistency. These updated independent fission product yields were utilized in the ORIGEN code to compare the calculated fission product inventories with experimentally measured inventories, with particular attention given to the noble gases. Another important outcome of this work is the development of fission product yield covariance data necessary for fission product uncertainty quantification. The evaluation methodology combines a sequential Bayesian method to guarantee consistency between independent and cumulative yields along with the physical constraints on the independent yields. This work was motivated to improve the performance of the ENDF/B-VII.1 library for stable and long-lived fission products. The revised fission product yields and the new covariance data are proposed as a revision to the fission yield data currently in ENDF/B-VII.1.

  3. New Fission-Product Waste Forms: Development and Characterization

    SciTech Connect

    Alexandra Navrotsky

    2010-07-30

    Research performed on the program “New Fission Product Waste Forms: Development and Characterization,” in the last three years has fulfilled the objectives of the proposal which were to 1) establish ceramic waste forms for disposing of Cs, Sr and minor actinides, 2) fully characterize the phase relationships, structures and thermodynamic and kinetic stabilities of promising waste forms, 3) establish a sound technical basis for understanding key waste form properties, such as melting temperatures and aqueous durability, based on an in-depth understanding of waste form structures and thermochemistry, and 4) establish synthesis, testing, scaleup and commercialization routes for wasteform implementation through out in-kind collaborations. In addition, since Cs and Sr form new elements by radioactive decay, the behavior and thermodynamics of waste forms containing different proportions of Cs, Sr and their decay products were discovered using non-radioactive analogues. Collaborations among researchers from three institutions, UC Davis, Sandia National Laboratories, and Shott Inc., were formed to perform the primary work on the program. The unique expertise of each of the members in the areas of waste form development, structure/property relationships, hydrothermal and high temperature synthesis, crystal/glass production, and thermochemistry was critical to program success. In addition, collaborations with the Brigham Young Univeristy, Ben Gurion University, and Los Alamos National Laboratory, were established for standard entropies of ceramic waste forms, sol-gel synthesis, and high temperature synthesis. This work has had a significant impact in a number of areas. First, the studies of the thermodynamic stability of the mineral analogues provided an important technical foundation for assessment the viability of multicomponent oxide phases for Cs and Sr removal. Moreover, the thermodynamic data discovered in this program established information on the reaction

  4. Angular distribution of products of ternary nuclear fission induced by cold polarized neutrons

    SciTech Connect

    Bunakov, V. E. Kadmensky, S. G. Kadmensky, S. S.

    2008-11-15

    Within quantum fission theory, angular distributions of products originating from the ternary fission of nuclei that is induced by polarized cold and thermal neutrons are investigated on the basis of a non-evaporative mechanism of third-particle emission and a consistent description of fission-channel coupling. It is shown that the inclusion of Coriolis interaction both in the region of the discrete and in the region of the continuous spectrum of states of the system undergoing fission leads to T-odd correlations in the aforementioned angular distributions. The properties of the TRI and ROT effects discovered recently, which are due to the interference between the fission amplitudes of neutron resonances, are explored. The results obtained here are compared with their counterparts from classic calculations based on the trajectory method.

  5. ARSENATE CARRIER PRECIPITATION METHOD OF SEPARATING PLUTONIUM FROM NEUTRON IRRADIATED URANIUM AND RADIOACTIVE FISSION PRODUCTS

    DOEpatents

    Thompson, S.G.; Miller, D.R.; James, R.A.

    1961-06-20

    A process is described for precipitating Pu from an aqueous solution as the arsenate, either per se or on a bismuth arsenate carrier, whereby a separation from uranium and fission products, if present in solution, is accomplished.

  6. The LANL C-NR counting room and fission product yields

    SciTech Connect

    Jackman, Kevin Richard

    2015-09-21

    This PowerPoint presentation focused on the following areas: LANL C-NR counting room; Fission product yields; Los Alamos Neutron wheel experiments; Recent experiments ad NCERC; and Post-detonation nuclear forensics

  7. RARE-EARTH METAL FISSION PRODUCTS FROM LIQUID U-Bi

    DOEpatents

    Wiswall, R.H.

    1960-05-10

    Fission product metals can be removed from solution in liquid bismuth without removal of an appreciable quantity of uranium by contacting the liquid metal solution with fused halides, as for example, the halides of sodium, potassium, and lithium and by adding to the contacted phases a quantity of a halide which is unstable relative to the halides of the fission products, a specific unstable halide being MgCl/sub 3/.

  8. Towards a multiscale approach for assessing fission product behaviour in UN

    NASA Astrophysics Data System (ADS)

    Klipfel, M.; Di Marcello, V.; Schubert, A.; van de Laar, J.; Van Uffelen, P.

    2013-11-01

    Ab initio modelling of fission products (i.e. Nb, Y, Gd, Nd, Zr, Sm, Eu, Ce, Ba, Mo, Sr, Rh, Pd, and Ru) in uranium nitride is carried out by assessing the incorporation, along with their contributions to local swelling of the fuel matrix. Fission products (FP's) in UN have shown to be preferably accommodated at U vacancies in bound [1 0 0]-Schottky defects, nevertheless, similar incorporation energies were found at a single U vacancy. From the investigated incorporation and migration mechanism, we found that FP's in UN predominately migrate along U-U vacancies, since the incorporation energies for all FP are lowest at single U vacancy or at the U vacancy in a Schottky defect. The energy required to induce a migration of a volatile FP from an N vacancy to U vacancy is about 4-5.5 eV. The local volume changes caused by the fission-product substitution have been assessed by means of DFT and combined with the fission-product concentrations obtained by means of neutron calculations (SCALE) to predict fission product swelling in UN. The linear swelling of nitride fuel resulting from these calculations, and the assumption that fission products do not interact and form secondary phases, leads to a reasonable estimation for the swelling rate as a function of burn-up (or time) when compared with empirical correlations in the open literature.

  9. Gas production apparatus

    DOEpatents

    Winsche, Warren E.; Miles, Francis T.; Powell, James R.

    1976-01-01

    This invention relates generally to the production of gases, and more particularly to the production of tritium gas in a reliable long operating lifetime systems that employs solid lithium to overcome the heretofore known problems of material compatibility and corrosion, etc., with liquid metals. The solid lithium is irradiated by neutrons inside low activity means containing a positive (+) pressure gas stream for removing and separating the tritium from the solid lithium, and these means are contained in a low activity shell containing a thermal insulator and a neutron moderator.

  10. Inherent safety phenomenon of fission-gas induced axial extrusion in oxide and metal fueled LMFBRs

    SciTech Connect

    Miles, K.J.; Kalimullah

    1985-01-01

    The current emphasis in LMFBR design is to develop reactor systems that contain as many features as possible to limit the severity of hypothetical accidents and provide the maximum time before corrective action is required while maintaining low capital costs. One feature is the possibility of fission-gas induced axial extrusion of the fuel within the intact cladding. The potential exists for this phenomenon to enable the reactor to withstand most accidents of the TOP variety, or at least provide an extended time for corrective action to be taken. Under transient conditions which produce a heating of the fuel above its nominal operating temperature, thermal expansion of the material axially produces a negative reactivity effect. This effect is presently considered in most accident analysis codes. The phenomenon of fission-gas induced axial extrusion has received renewed interest because of the consideration of metal alloys of uranium and plutonium for the fuel in some current reactor designs.

  11. 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.

  12. MOX and MOX with 237Np/241Am Inert Fission Gas Generation Comparison in ATR

    SciTech Connect

    G. S. Chang; M. Robel; W. J. Carmack; D. J. Utterbeck

    2006-06-01

    The treatment of spent fuel produced in nuclear power generation is one of the most important issues to both the nuclear community and the general public. One of the viable options to long-term geological disposal of spent fuel is to extract plutonium, minor actinides (MA), and potentially long-lived fission products from the spent fuel and transmute them into short-lived or stable radionuclides in currently operating light-water reactors (LWR), thus reducing the radiological toxicity of the nuclear waste stream. One of the challenges is to demonstrate that the burnup-dependent characteristic differences between Reactor-Grade Mixed Oxide (RG-MOX) fuel and RG-MOX fuel with MA Np-237 and Am 241 are minimal, particularly, the inert gas generation rate, such that the commercial MOX fuel experience base is applicable. Under the Advanced Fuel Cycle Initiative (AFCI), developmental fuel specimens in experimental assembly LWR-2 are being tested in the northwest (NW) I-24 irradiation position of the Advanced Test Reactor (ATR). The experiment uses MOX fuel test hardware, and contains capsules with MOX fuel consisting of mixed oxide manufactured fuel using reactor grade plutonium (RG-Pu) and mixed oxide manufactured fuel using RG-Pu with added Np/Am. This study will compare the fuel neutronics depletion characteristics of Case-1 RG-MOX and Case-2 RG-MOX with Np/Am.

  13. Detecting special nuclear materials in containers using high-energy gamma rays emitted by fission products

    DOEpatents

    Norman, Eric B.; Prussin, Stanley G.

    2007-10-02

    A method and a system for detecting the presence of special nuclear materials in a container. The system and its method include irradiating the container with an energetic beam, so as to induce a fission in the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

  14. Artificial neural network modeling for fission gas release in LWR UO 2 fuel under RIA conditions

    NASA Astrophysics Data System (ADS)

    Koo, Yang-Hyun; Oh, Jae-Yong; Lee, Byung-Ho; Tahk, Young-Wook; Song, Kun-Woo

    2010-10-01

    A fission gas release (FGR) model was developed by using an artificial neural network method to predict fission gas release in UO 2 fuel under reactivity initiated accident (RIA) conditions. Based on the test data obtained in the CABRI test reactor and nuclear safety research reactor, the model takes into account the effect of the five parameters: pellet average burnup, peak fuel enthalpy, the ratio of peak fuel enthalpy to pulse width, fission gas release during base-irradiation, and grain size of a fuel pellet. The parametric study of the model, producing a physically reasonable trend of FGR for each parameter, shows that the pellet average burnup and the ratio of peak fuel enthalpy to pulse width are two of the most important parameters. Depending on the combination of input values for the five parameters, the application of the model to a fuel rod under typical RIA conditions of light water reactor produces 1.7-14.0% of FGR for the pellet average burnup ranging from 20 to 70 MW d/kg U.

  15. Fission signal detection using helium-4 gas fast neutron scintillation detectors

    SciTech Connect

    Lewis, J. M. Kelley, R. P.; Jordan, K. A.; Murer, D.

    2014-07-07

    We demonstrate the unambiguous detection of the fission neutron signal produced in natural uranium during active neutron interrogation using a deuterium-deuterium fusion neutron generator and a high pressure {sup 4}He gas fast neutron scintillation detector. The energy deposition by individual neutrons is quantified, and energy discrimination is used to differentiate the induced fission neutrons from the mono-energetic interrogation neutrons. The detector can discriminate between different incident neutron energies using pulse height discrimination of the slow scintillation component of the elastic scattering interaction between a neutron and the {sup 4}He atom. Energy histograms resulting from this data show the buildup of a detected fission neutron signal at higher energies. The detector is shown here to detect a unique fission neutron signal from a natural uranium sample during active interrogation with a (d, d) neutron generator. This signal path has a direct application to the detection of shielded nuclear material in cargo and air containers. It allows for continuous interrogation and detection while greatly minimizing the potential for false alarms.

  16. Simulating γ-γ coincidences of β-delayed γ-rays from fission product nuclei

    NASA Astrophysics Data System (ADS)

    Padgett, Stephen; Wang, Tzu-Fang

    2015-01-01

    Analyzing radiation from material that has undergone neutron induced fission is important for fields such as nuclear forensics, reactor physics, and nonproliferation monitoring. The γ-ray spectroscopy of fission products is a major part of the characterization of a material's fissile inventory and the energy of incident neutrons inducing fission. Cumulative yields and γ-ray intensities from nuclear databases are inputs into a GEANT4 simulation to create expected γ-ray spectra from irradiated 235U. The simulations include not only isotropically emitted γ-rays but also γ-γ cascades from certain fission products, emitted with their appropriate angular correlations. Here γ singles spectra as well as γ-γ coincidence spectra are simulated in detectors at both 90° and 180° pairings. The ability of these GEANT4 Monte Carlo simulations to duplicate experimental data is explored in this work. These simulations demonstrate potential in exploiting angular correlations of γ-γ cascades in fission product decays to determine isotopic content. Analyzing experimental and simulated γ-γ coincidence spectra as opposed to singles spectra should improve the ability to identify fission product nuclei since such spectra are cleaner and contain more resolved peaks when compared to γ singles spectra.

  17. Mass-resolved angular distribution of fission products in the 20Ne+232Th reaction

    NASA Astrophysics Data System (ADS)

    Tripathi, R.; Sodaye, S.; Sudarshan, K.; Guin, R.

    2013-08-01

    Mass-resolved angular distributions of fission product were measured in the 20Ne + 232Th reaction at Elab = 125.6 and 142.5 MeV using the recoil catcher technique followed by offline γ-ray spectrometry. Angular anisotropy was found to decrease with increasing asymmetry of mass division. Angular anisotropies of the fission products in the symmetric region were significantly higher compared to those calculated using the statistical saddle-point model. Experimental anisotropies could be explained after considering the contribution from pre-equilibrium fission. Use of barrier energies corresponding to different mass asymmetry values in the calculations could reasonably reproduce the mass dependence of angular anisotropies. The role of barrier energies in governing the angular anisotropy indicates that the mass dependence of anisotropy may possibly be a distinguishing feature of pre-equilibrium fission from quasifission, in which the composite system escapes into the exit channel without being captured inside the saddle point.

  18. Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations

    SciTech Connect

    Mueller, Don; Bowen, Douglas G; Marshall, William BJ J

    2015-01-01

    The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members accept the use of either 1.5 or 3% of the FP&MA worth—in addition to bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF—to conservatively account for the bias and bias uncertainty associated with the specified unvalidated FP&MAs. The ISG recommends (1) use of 1.5% of the FP&MA worth if a modern version of SCALE and its nuclear data are used and (2) 3% of the FP&MA worth for well qualified, industry standard code systems other than SCALE with the Evaluated Nuclear Data Files, Part B (ENDF/B),-V, ENDF/B-VI, or ENDF/B-VII cross sections libraries. The work presented in this paper provides a basis for extending the use of the 1.5% of the FP&MA worth bias to BUC criticality calculations performed using the Monte Carlo N-Particle (MCNP) code. The extended use of the 1.5% FP&MA worth bias is shown to be acceptable by comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII–based nuclear data. The comparison supports use of the 1.5% FP&MA worth bias when the MCNP code is used for criticality calculations, provided that the cask design is similar to the hypothetical generic BUC-32 cask model and that the credited FP&MA worth is no more than 0.1 Δkeff (ISG-8, Rev. 3, Recommendation 4).

  19. Relative fission product yield determination in the USGS TRIGA Mark I reactor

    NASA Astrophysics Data System (ADS)

    Koehl, Michael A.

    Fission product yield data sets are one of the most important and fundamental compilations of basic information in the nuclear industry. This data has a wide range of applications which include nuclear fuel burnup and nonproliferation safeguards. Relative fission yields constitute a major fraction of the reported yield data and reduce the number of required absolute measurements. Radiochemical separations of fission products reduce interferences, facilitate the measurement of low level radionuclides, and are instrumental in the analysis of low-yielding symmetrical fission products. It is especially useful in the measurement of the valley nuclides and those on the extreme wings of the mass yield curve, including lanthanides, where absolute yields have high errors. This overall project was conducted in three stages: characterization of the neutron flux in irradiation positions within the U.S. Geological Survey TRIGA Mark I Reactor (GSTR), determining the mass attenuation coefficients of precipitates used in radiochemical separations, and measuring the relative fission products in the GSTR. Using the Westcott convention, the Westcott flux, modified spectral index, neutron temperature, and gold-based cadmium ratios were determined for various sampling positions in the USGS TRIGA Mark I reactor. The differential neutron energy spectrum measurement was obtained using the computer iterative code SAND-II-SNL. The mass attenuation coefficients for molecular precipitates were determined through experiment and compared to results using the EGS5 Monte Carlo computer code. Difficulties associated with sufficient production of fission product isotopes in research reactors limits the ability to complete a direct, experimental assessment of mass attenuation coefficients for these isotopes. Experimental attenuation coefficients of radioisotopes produced through neutron activation agree well with the EGS5 calculated results. This suggests mass attenuation coefficients of molecular

  20. 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.; Maki, John T.

    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

  1. Analysis of fission gas release in LWR fuel using the BISON code

    SciTech Connect

    G. Pastore; J.D. Hales; S.R. Novascone; D.M. Perez; B.W. Spencer; R.L. Williamson

    2013-09-01

    Recent advances in the development of the finite-element based, multidimensional fuel performance code BISON of Idaho National Laboratory are presented. Specifically, the development, implementation and testing of a new model for the analysis of fission gas behavior in LWR-UO2 fuel during irradiation are summarized. While retaining a physics-based description of the relevant mechanisms, the model is characterized by a level of complexity suitable for application to engineering-scale nuclear fuel analysis and consistent with the uncertainties pertaining to some parameters. The treatment includes the fundamental features of fission gas behavior, among which are gas diffusion and precipitation in fuel grains, growth and coalescence of gas bubbles at grain faces, grain growth and grain boundary sweeping effects, thermal, athermal, and transient gas release. The BISON code incorporating the new model is applied to the simulation of irradiation experiments from the OECD/NEA International Fuel Performance Experiments database, also included in the IAEA coordinated research projects FUMEX-II and FUMEX-III. The comparison of the results with the available experimental data at moderate burn-up is presented, pointing out an encouraging predictive accuracy, without any fitting applied to the model parameters.

  2. Measuring and Predicting Fission Product Noble Metals in SRS HLW Sludges

    SciTech Connect

    Bibler, N

    2005-04-05

    The noble metals Ru, Rh, Pd, and Ag were produced in the Savannah River Site (SRS) reactors as products of the fission of U-235. Consequently they are in the High Level Waste (HLW) sludges that are currently being immobilized into a borosilicate glass in the Defense Waste Processing Facility (DWPF). The noble metals are a concern in the DWPF because they catalyze the decomposition of formic acid used in the process to produce the flammable gas hydrogen. As the concentration of these noble metals in the sludge increases, more hydrogen will be produced when this sludge is processed. In the SRS Tank Farm it takes approximately two years to prepare a sludge batch for processing in the DWPF. This length of time is necessary to mix the appropriate sludges, blend them to form a sludge batch and then wash it to enable processing in the DWPF. This means that the exact composition of a sludge batch is not known for {approx}two years. During this time, studies with simulated nonradioactive sludges must be performed to determine the desired DWPF processing parameters for the new sludge batch. Consequently, prediction of the noble metal concentrations is desirable to prepare appropriate simulated sludges for studies of the DWPF process for that sludge batch. These studies give a measure of the amount of hydrogen that will be produced when that sludge batch is processed. This report describes in detail the measurement of these noble metal concentrations in sludges and a way to predict their concentrations from an estimate of the lanthanum concentration in the sludge. Results for two sludges are presented in this report. These are Sludge Batch 3 (SB3) currently being processed by the DWPF and a sample of unwashed sludge from Tank 11 that will be part of Sludge Batch 4. The concentrations of the noble metals in HLW sludges are measured by using mass spectroscopy to determine concentrations of the isotopes that comprise each noble metal. For example, the noble metal Ru is comprised

  3. Continuous fission-product monitor system at Oyster Creek. Final report

    SciTech Connect

    Collins, L.L.; Chulick, E.T.

    1980-10-01

    A continuous on-line fission product monitor has been installed at the Oyster Creek Nuclear Generating Station, Forked River, New Jersey. The on-line monitor is a minicomputer-controlled high-resolution gamma-ray spectrometer system. An intrinsic Ge detector scans a collimated sample line of coolant from one of the plant's recirculation loops. The minicomputer is a Nuclear Data 6620 system. Data were accumulated for the period from April 1979 through January 1980, the end of cycle 8 for the Oyster Creek plant. Accumulated spectra, an average of three a day, were stored on magnetic disk and subsequently analyzed for fisson products, Because of difficulties in measuring absolute detector efficiency, quantitative fission product concentrations in the coolant could not be determined. Data for iodine fission products are reported as a function of time. The data indicate the existence of fuel defects in the Oyster Creek core during cycle 8.

  4. Migration of fission products at the Nevada Test Site: Detection with an isotopic tracer

    SciTech Connect

    Thompton, J.L.; Gilmore, J.S. )

    1989-01-01

    Researchers at Los Alamos National Laboratory are studying the migration of fission products away from explosion cavities formed by underground nuclear tests at the Nevada Test Site. In some cases, the isotopic composition of the fission products or activation products associated with a particular test are distinctive and we may identify them many years after the event. In this paper we describe a case in which we used rhodium isotopes to identify the source of radioactive material that had moved some 350 m from the explosion site. 4 refs., 2 figs., 2 tabs.

  5. On the Radiochemical Separations of the Beta-emitting Fission Products

    NASA Astrophysics Data System (ADS)

    Chang, Zheng; Sudowe, Ralf

    2013-04-01

    This research aims at developing fast and effective radiochemical procedures for separation of the beta-emitting fission products that are difficult to analyze by gamma-spectrometry. Post-detonation analysis, as one of the major tasks of nuclear forensics, can provide crucial information for identification of the explosion levels, fuel sources, and industrial processes of a nuclear device. However, a dozen of radionuclides with high fission yields such as Zr-93, Tc-99, Sr-90 are either pure beta-emitters or only emitting gamma-rays that are difficult to analyze. Although the analysis of these radionuclides was thoroughly studied, samples from unknown nuclear detonations can be complicated by the number of fission products, radioactivity levels, sample matrices, and time limits for analysis. The challenge facing the forensic analysis should not be underestimated. A sequential separation procedure is designed to analyze the major beta-emitting fission products. Radiochemical techniques such as solvent extraction, precipitation, and column chromatography are utilized. The procedure will be tested and improved by experiments. The final procedure should be capable of analyzing the fission products under various sample conditions effectively and rapidly.

  6. Investigation of the diffusion of atomic fission products in UC by density functional calculations

    NASA Astrophysics Data System (ADS)

    Bévillon, Émile; Ducher, Roland; Barrachin, Marc; Dubourg, Roland

    2013-03-01

    Activation energies of U and C atoms self-diffusion in UC, as well as activation energies of hetero-diffusion of fission products (FPs) are investigated by first-principles calculations. According to a previous study which showed a likely U site occupation was favoured for all the FPs, their diffusion is restricted to the uranium sublattice of UC in the present study. In this framework, long-range displacements are only possible through a concerted mechanism with a surrounding uranium vacancy. Using the apparent formation energies of the uranium vacancy defect calculated in our previous study and the classical approach used in UO2 by Andersson et al., the activation energies of the main fission products in the various stoichiometric domains have been calculated. The results are compared to those obtained with the five frequency model applied to two representative fission products, Xe and Zr. Interestingly, despite strong differences of formalism, both models provided similar activation energies.

  7. Results of fission products β decay properties measurement performed with a total absorption spectrometer

    NASA Astrophysics Data System (ADS)

    Zakari-Issoufou, A.-A.; Porta, A.; Fallot, M.; Algora, A.; Tain, J. L.; Valencia, E.; Rice, S.; Agramunt, J.; Äystö, J.; Bowry, M.; Bui, V. M.; Caballero-Folch, R.; Cano-Ott, D.; Eloma, V.; Estévez, E.; Farrelly, G. F.; Garcia, A.; Gelletly, W.; Gomez-Hornillos, M. B.; Gorlychev, V.; Hakala, J.; Jokinen, A.; Jordan, M. D.; Kankainen, A.; Kondev, F. G.; Martinez, T.; Mendoza, E.; Molina, F.; Moore, I.; Perez, A.; Podolyak, Zs.; Penttilä, H.; Regan, P. H.; Rissanen, J.; Rubio, B.; Weber, C.

    2014-03-01

    β-decay properties of fission products are very important for applied reactor physics, for instance to estimate the decay heat released immediately after the reactor shutdown and to estimate the bar ν flux emitted. An accurate estimation of the decay heat and the bar ν emitted flux from reactors, are necessary for purposes such as reactors operation safety and non-proliferation. In order to improve the precision in the prediction for these quantities, the bias due to the Pandemonium effect affecting some important fission product data has to be corrected. New measurements of fission products β-decay, not sensitive to this effect, have been performed with a Total Absorption Spectrometer (TAS) at the JYFL facility of Jyväskylä. An overview of the TAS technique and first results from the 2009 campaign will be presented.

  8. Trapping and diffusion of fission products in ThO2 and CeO2

    SciTech Connect

    Xiao, Haiyan; Zhang, Yanwen; Weber, William J

    2011-01-01

    The trapping and diffusion of Br, Rb, Cs and Xe in ThO2 and CeO{sub 2} have been studied using an Ab Initio total energy method in the local-density approximation of density functional theory. Fission products incorporated in cation mono-vacancy, cation-anion di-vacancy and Schottky defect sites are found to be stable, with the cation mono-vacancy being the preferred site in most cases. In both oxides, Rb and Cs are the most likely to be trapped, and Xe is more difficult to incorporate than other fission products. The energy barriers for migration of each species in ThO{sub 2} and CeO{sub 2} are also calculated. Alkali metals are relatively more mobile than other fission products, and bromine is the least mobile.

  9. Immobilization of fission products arising from pyrometallurgical reprocessing in chloride media

    NASA Astrophysics Data System (ADS)

    Leturcq, G.; Grandjean, A.; Rigaud, D.; Perouty, P.; Charlot, M.

    2005-12-01

    Spent nuclear fuel reprocessing to recover energy-producing elements such as uranium or plutonium can be performed by a pyrochemical process. In such method, the actinides and fission products are extracted by electrodeposition in a molten chloride medium. These processes generate chlorinated alkali salt flows contaminated by fission products, mainly Cs, Ba, Sr and rare earth elements constituting high-level waste. Two possible alternatives are investigated for managing this wasteform; a protocol is described for dechlorinating the fission products to allow vitrification, and mineral phases capable of immobilizing chlorides are listed to allow specification of a dedicated ceramic matrix suitable for containment of these chlorinated waste streams. The results of tests to synthesize chlorosilicate phases are also discussed.

  10. PROCESS USING BISMUTH PHOSPHATE AS A CARRIER PRECIPITATE FOR FISSION PRODUCTS AND PLUTONIUM VALUES

    DOEpatents

    Finzel, T.G.

    1959-03-10

    A process is described for separating plutonium from fission products carried therewith when plutonium in the reduced oxidation state is removed from a nitric acid solution of irradiated uranium by means of bismuth phosphate as a carrier precipitate. The bismuth phosphate carrier precipitate is dissolved by treatment with nitric acid and the plutonium therein is oxidized to the hexavalent oxidation state by means of potassium dichromate. Separation of the plutonium from the fission products is accomplished by again precipitating bismuth phosphate and removing the precipitate which now carries the fission products and a small percentage of the plutonium present. The amount of plutonium carried in this last step may be minimized by addition of sodium fluoride, so as to make the solution 0.03N in NaF, prior to the oxidation and prccipitation step.

  11. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 3: Fission-Product Transport and Dose PIRTs

    SciTech Connect

    Morris, Robert Noel

    2008-03-01

    This Fission Product Transport (FPT) Phenomena Identification and Ranking Technique (PIRT) report briefly reviews the high-temperature gas-cooled reactor (HTGR) FPT mechanisms and then documents the step-by-step PIRT process for FPT. The panel examined three FPT modes of operation: (1) Normal operation which, for the purposes of the FPT PIRT, established the fission product circuit loading and distribution for the accident phase. (2) Anticipated transients which were of less importance to the panel because a break in the pressure circuit boundary is generally necessary for the release of fission products. The transients can change the fission product distribution within the circuit, however, because temperature changes, flow perturbations, and mechanical vibrations or shocks can result in fission product movement. (3) Postulated accidents drew the majority of the panel's time because a breach in the pressure boundary is necessary to release fission products to the confinement. The accidents of interest involved a vessel or pipe break, a safety valve opening with or without sticking, or leak of some kind. Two generic scenarios were selected as postulated accidents: (1) the pressurized loss-of-forced circulation (P-LOFC) accident, and (2) the depressurized loss-of-forced circulation (D-LOFC) accidents. FPT is not an accident driver; it is the result of an accident, and the PIRT was broken down into a two-part task. First, normal operation was seen as the initial starting point for the analysis. Fission products will be released by the fuel and distributed throughout the reactor circuit in some fashion. Second, a primary circuit breach can then lead to their release. It is the magnitude of the release into and out of the confinement that is of interest. Depending on the design of a confinement or containment, the impact of a pressure boundary breach can be minimized if a modest, but not excessively large, fission product attenuation factor can be introduced into the

  12. Diffusion of fission products and radiation damage in SiC

    NASA Astrophysics Data System (ADS)

    Malherbe, Johan B.

    2013-11-01

    A major problem with most of the present nuclear reactors is their safety in terms of the release of radioactivity into the environment during accidents. In some of the future nuclear reactor designs, i.e. Generation IV reactors, the fuel is in the form of coated spherical particles, i.e. TRISO (acronym for triple coated isotropic) particles. The main function of these coating layers is to act as diffusion barriers for radioactive fission products, thereby keeping these fission products within the fuel particles, even under accident conditions. The most important coating layer is composed of polycrystalline 3C-SiC. This paper reviews the diffusion of the important fission products (silver, caesium, iodine and strontium) in SiC. Because radiation damage can induce and enhance diffusion, the paper also briefly reviews damage created by energetic neutrons and ions at elevated temperatures, i.e. the temperatures at which the modern reactors will operate, and the annealing of the damage. The interaction between SiC and some fission products (such as Pd and I) is also briefly discussed. As shown, one of the key advantages of SiC is its radiation hardness at elevated temperatures, i.e. SiC is not amorphized by neutrons or bombardment at substrate temperatures above 350 °C. Based on the diffusion coefficients of the fission products considered, the review shows that at the normal operating temperatures of these new reactors (i.e. less than 950 °C) the SiC coating layer is a good diffusion barrier for these fission products. However, at higher temperatures the design of the coated particles needs to be adapted, possibly by adding a thin layer of ZrC.

  13. A physical description of fission product behavior fuels for advanced power reactors.

    SciTech Connect

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

  14. Design and operation of gamma scan and fission gas sampling systems for characterization of irradiated commercial nuclear fuel

    SciTech Connect

    Knox, C.A.; Thornhill, R.E.; Mellinger, G.B.

    1989-09-01

    One of the primary objectives of the Materials Characterization Center (MCC) is to acquire and characterize spent fuels used in waste form testing related to nuclear waste disposal. The initial steps in the characterization of a fuel rod consist of gamma scanning the rod and sampling the gas contained in the fuel rod (referred to as fission gas sampling). The gamma scan and fission gas sampling systems used by the MCC are adaptable to a wide range of fuel types and have been successfully used to characterize both boiling water reactor (BWR) and pressurized water reactor (PWR) fuel rods. This report describes the design and operation of systems used to gamma scan and fission gas sample full-length PWR and BWR fuel rods. 1 ref., 10 figs., 1 tab.

  15. Modeling the influence of bubble pressure on grain boundary separation and fission gas release

    SciTech Connect

    Pritam Chakraborty; Michael R. Tonks; Giovanni Pastore

    2014-09-01

    Grain boundary (GB) separation as a mechanism for fission gas release (FGR), complementary to gas bubble interlinkage, has been experimentally observed in irradiated light water reactor fuel. However there has been limited effort to develop physics-based models incorporating this mechanism for the analysis of FGR. In this work, a computational study is carried out to investigate GB separation in UO2 fuel under the effect of gas bubble pressure and hydrostatic stress. A non-dimensional stress intensity factor formula is obtained through 2D axisymmetric analyses considering lenticular bubbles and Mode-I crack growth. The obtained functional form can be used in higher length-scale models to estimate the contribution of GB separation to FGR.

  16. Chemical forms of solid fission products in the irradiated uranium—plutonium mixed nitride fuel

    NASA Astrophysics Data System (ADS)

    Arai, Yasuo; Maeda, Atsushi; Shiozawa, Ken-ichi; Ohmichi, Toshihiko

    1994-06-01

    Chemical forms of solid fission products in the irradiated (U, Pu)N fuel were estimated by both thermodynamic equilibrium calculation and electron microprobe analysis on burnup simulated samples prepared by carbothermic reduction. Besides the MX type matrix phase dissolving zirconium, niobium, yttrium and rare earth elements, the existence of two kinds of inclusion was recognized. One is URu 3 type intermetallic compound constituted by uranium and platinum group elements. The other is an alloy containing molybdenum as a principal constituent. Furthermore, the swelling rate due to solid fission products precipitation was evaluated to be about 0.5% per %FIMA.

  17. A separate effect study of the influence of metallic fission products on CsI radioactive release from nuclear fuel

    NASA Astrophysics Data System (ADS)

    Di Lemma, F. G.; Colle, J. Y.; Beneš, O.; Konings, R. J. M.

    2015-10-01

    The chemistry of cesium and iodine is of main importance to quantify the radioactive release in case of a nuclear reactor accident, or sabotage involving irradiated nuclear materials. We studied the interaction of CsI with different metallic fission products such as Mo and Ru. These elements can be released from nuclear fuel when exposed to oxidising conditions, as in the case of contact of overheated nuclear fuel with air (e.g. in a spent fuel cask sabotage, uncovering of a spent fuel pond, or air ingress accidents). Experiments were performed by vaporizing mixtures of the compounds in air, and analysing the produced aerosols in view of a possible gas-gas and gas-aerosol reactions between the compounds. These results were compared with the gaseous species predicted by thermochemical equilibrium calculations and experimental equilibrium vaporization tests using Knudsen Effusion Mass Spectrometry.

  18. Integrated separation scheme for measuring a suite of fission and activation products from a fresh mixed fission and activation product sample

    SciTech Connect

    Morley, Shannon M.; Seiner, Brienne N.; Finn, Erin C.; Greenwood, Lawrence R.; Smith, Steven C.; Gregory, Stephanie J.; Haney, Morgan M.; Lucas, Dawn D.; Arrigo, Leah M.; Beacham, Tere A.; Swearingen, Kevin J.; Friese, Judah I.; Douglas, Matthew; Metz, Lori A.

    2015-05-01

    Mixed fission and activation materials resulting from various nuclear processes and events contain a wide range of isotopes for analysis spanning almost the entire periodic table. In some applications such as environmental monitoring, nuclear waste management, and national security a very limited amount of material is available for analysis and characterization so an integrated analysis scheme is needed to measure multiple radionuclides from one sample. This work describes the production of a complex synthetic sample containing fission products, activation products, and irradiated soil and determines the percent recovery of select isotopes through the integrated chemical separation scheme. Results were determined using gamma energy analysis of separated fractions and demonstrate high yields of Ag (76 ± 6%), Au (94 ± 7%), Cd (59 ± 2%), Co (93 ± 5%), Cs (88 ± 3%), Fe (62 ± 1%), Mn (70 ± 7%), Np (65 ± 5%), Sr (73 ± 2%) and Zn (72 ± 3%). Lower yields (< 25%) were measured for Ga, Ir, Sc, and W. Based on the results of this experiment, a complex synthetic sample can be prepared with low atom/fission ratios and isotopes of interest accurately and precisely measured following an integrated chemical separation method.

  19. Nuclear Power from Fission Reactors. An Introduction.

    ERIC Educational Resources Information Center

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  20. Pyrene degradation by a Mycobacterium sp. : Identification of ring oxidation and ring fission products

    SciTech Connect

    Heitkamp, M.A.; Freeman, J.P.; Miller, D.W.; Cerniglia, C.E. )

    1988-10-01

    The degradation of pyrene, a polycyclic aromatic hydrocarbon containing four aromatic rings, by pure cultures of a Mycobacterium sp. was studied. Over 60% of ({sup 14}C)pyrene was mineralized to CO{sub 2} after 96 h of incubation at 24{degree}C. High-pressure liquid chromatography analyses showed the presence of one major and at least six other metabolites that accounted for 95% of the total organic-extractable {sup 14}C-labeled residues. Analyses by UV, infrared, mass, and nuclear magnetic resonance spectrometry and gas chromatography identified both pyrene cis- and trans-4,5-dihydrodiols and pyrenol as initial microbial ring-oxidation products of pyrene. The major metabolite, 4-phenanthroic acid, and 4-hydroxyperinaphthenone and cinnamic and phthalic acids were identified as ring fission products. {sup 18}O{sub 2} studies showed that the formation of cis- and trans-4,5-dihydrodiols were catalyzed by dioxygenase and monooxygenase enzymes, respectively. This is the first report of the chemical pathway for the microbial catabolism of pyrene.

  1. Natural gas production from Arctic gas hydrates

    SciTech Connect

    Collett, T.S. )

    1993-01-01

    The natural gas hydrates of the Messoyakha field in the West Siberian basin of Russia and those of the Prudhoe Bay-Kuparuk River area on the North Slope of Alaska occur within a similar series of interbedded Cretaceous and Tertiary sandstone and siltstone reservoirs. Geochemical analyses of gaseous well-cuttings and production gases suggest that these two hydrate accumulations contain a mixture of thermogenic methane migrated from a deep source and shallow, microbial methane that was either directly converted to gas hydrate or was first concentrated in existing traps and later converted to gas hydrate. Studies of well logs and seismic data have documented a large free-gas accumulation trapped stratigraphically downdip of the gas hydrates in the Prudhoe Bay-Kuparuk River area. The presence of a gas-hydrate/free-gas contact in the Prudhoe Bay-Kuparuk River area is analogous to that in the Messoyakha gas-hydrate/free-gas accumulation, from which approximately 5.17x10[sup 9] cubic meters (183 billion cubic feet) of gas have been produced from the hydrates alone. The apparent geologic similarities between these two accumulations suggest that the gas-hydrated-depressurization production method used in the Messoyakha field may have direct application in northern Alaska. 30 refs., 15 figs., 3 tabs.

  2. Concentration-triggered fission product release from zirconia: consequences for nuclear safety

    NASA Astrophysics Data System (ADS)

    Gentils, A.; Thomé, L.; Jagielski, J.; Garrido, F.

    2002-02-01

    Crystalline oxide ceramics, more particularly zirconia and spinel, are promising matrices for plutonium and minor actinide transmutation. An important issue concerning these materials is the investigation of their ability to confine radiotoxic elements resulting from the fission of actinides. This letter reports the study of the release, upon annealing or irradiation at high temperature, of one of the most toxic fission product (Cs) in zirconia. The foreign species are introduced by ion implantation and the release is studied by Rutherford backscattering experiments. The results emphasize the decisive influence of the fission product concentration on the release properties. The Cs mobility in zirconia is strongly increased when the impurity concentration exceeds a threshold of the order of a few atomic per cent. Irradiation with medium-energy heavy ions is shown to enhance Cs outdiffusion with respect to annealing at the same temperature.

  3. HTR-2014 Paper Comparison of fission product release predictions using PARFUME with results from the AGR-1 irradiation experiment

    SciTech Connect

    Blaise Collin

    2001-10-01

    The PARFUME (PARticle FUel ModEl) code was used to predict fission product release from tristructural isotropic (TRISO) coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of fission products silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination (PIE) measurements provided data on release of fission products from fuel compacts and fuel particles, and retention of fission products in the compacts outside of the SiC layer. PARFUME-predicted fractional release of these fission products was determined and compared to the PIE measurements. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed silicon carbide (SiC) layers, the over-prediction is by a factor of about two, corresponding to an over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of about 100. For intact particles, whose release is much lower, the over-prediction is by an average of about an order of magnitude, which could additionally be attributed to an over-estimated diffusivity in SiC by about 30%. The release of strontium from intact particles is also over-estimated by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release from intact particles varied considerably from compact to compact, making it difficult to assess the effective over-estimation of the diffusivities. Furthermore, the release of strontium from particles with failed SiC is difficult to observe experimentally due to the release from intact particles, preventing any conclusions to be made on the accuracy or validity of the

  4. Method of fission product beta spectra measurements for predicting reactor anti-neutrino emission

    SciTech Connect

    Asner, David M.; Burns, Kimberly A.; Campbell, Luke W.; Greenfield, Bryce A.; Kos, Marek S.; Orrell, John L.; Schram, Malachi; VanDevender, Brent A.; Wood, Lynn S.; Wootan, David W.

    2015-03-01

    The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron-rich fission products that subsequently beta decay and emit electron antineutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to today's precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurements performed in the 1970s and 1980s. The steadily improving reactor neutrino measurement techniques and recent reconsiderations of the agreement between the predicted and observed reactor neutrino flux motivates revisiting the underlying beta spectra measurements. A method is proposed to use an accelerator proton beam delivered to an engineered target to yield a neutron field tailored to reproduce the neutron energy spectrum present in the core of an operating nuclear reactor. Foils of the primary reactor fissionable isotopes placed in this tailored neutron flux will ultimately emit beta particles from the resultant fission products. Measurement of these beta particles in a time projection chamber with a perpendicular magnetic field provides a distinctive set of systematic considerations for comparison to the original seminal beta spectra measurements. Ancillary measurements such as gamma-ray emission and post-irradiation radiochemical analysis will further constrain the absolute normalization of beta emissions per fission. The requirements for unfolding the beta spectra measured with this method into a predicted reactor neutrino spectrum are explored.

  5. Fission product release and microstructure changes of irradiated MOX fuel at high temperatures

    NASA Astrophysics Data System (ADS)

    Colle, J.-Y.; Hiernaut, J.-P.; Wiss, T.; Beneš, O.; Thiele, H.; Papaioannou, D.; Rondinella, V. V.; Sasahara, A.; Sonoda, T.; Konings, R. J. M.

    2013-11-01

    burnups correspond reasonably well with measurement of Walker et al. [11]. All those data are shown Fig. 2.Fragments of 2-8 mg were chosen for the experiments. Since these specimens are small compared to the drilled sample size and were taken randomly, the precise radial position could not be determined, in particular the specimens of sample type, A and B could be from close radial locations.Specimens from each drilled sample type were annealed up to complete vaporisation (˜2600 K) at a speed of about 10 K min-1 in a Knudsen effusion mass spectrometer (KEMS) described previously [13,14]. In addition to helium and to the FGs all the species present in the vapour between 83 and 300 a.m.u. were measured during the heating. Additionally, the 85Kr isotope was analysed in a cold trap by β and γ counting. The long-lived fission gas isotopes correspond to masses 131, 132, 134 and 136 for Xe and 83, 84, 85 and 86 for Kr. The absolute quantities of gas released from specimens of sample types A and B were also determined using the in-house built Q-GAMES (Quantitative gas measurement system), described in detail in [15].For each of the samples, fragments were also annealed and measured in the KEMS up to specific temperatures corresponding to different stages of the FGs or He release. These fragments were subsequently analysed by Scanning Electron Microscopy (SEM, Philips XL40) [16] in order to investigate the relationship between structural changes, burn-up, irradiation temperature and fission products release. SEM observations were also done on the samples before the KEMS experiments and the fracture surface appearance of the samples is shown in Fig. 3, revealing the presence of the high burnup structure (HBS) in the Pu-rich agglomerates.A summary of the 12 samples analysed by KEMS, SEM and Q-GAMES is given in Table 1. At 1300 K no clear change potentially related to gas release appears in the UM and PA. At 1450 K a beginning of grain boundaries opening can be observed as well as

  6. PERCOLATION ON GRAIN BOUNDARY NETWORKS: APPLICATION TO FISSION GAS RELEASE IN NUCLEAR FUELS

    SciTech Connect

    Paul C. Millett

    2012-02-01

    The percolation behavior of grain boundary networks is characterized in two- and three-dimensional lattices with circular macroscale cross-sections that correspond to nuclear fuel elements. The percolation of gas bubbles on grain boundaries, and the subsequent percolation of grain boundary networks is the primary mechanism of fission gas release from nuclear fuels. Both radial cracks and radial gradients in grain boundary property distributions are correlated with the fraction of grain boundaries vented to the free surfaces. Our results show that cracks surprisingly do not significantly increase the percolation of uniform grain boundary networks. However, for networks with radial gradients in boundary properties, the cracks can considerably raise the vented grain boundary content.

  7. Report on the Behavior of Fission Products in the Co-decontamination Process

    SciTech Connect

    Martin, Leigh Robert; Riddle, Catherine Lynn

    2015-09-30

    This document was prepared to meet FCT level 3 milestone M3FT-15IN0302042, “Generate Zr, Ru, Mo and Tc data for the Co-decontamination Process.” This work was carried out under the auspices of the Lab-Scale Testing of Reference Processes FCT work package. This document reports preliminary work in identifying the behavior of important fission products in a Co-decontamination flowsheet. Current results show that Tc, in the presence of Zr alone, does not behave as the Argonne Model for Universal Solvent Extraction (AMUSE) code would predict. The Tc distribution is reproducibly lower than predicted, with Zr distributions remaining close to the AMUSE code prediction. In addition, it appears there may be an intricate relationship between multiple fission product metals, in different combinations, that will have a direct impact on U, Tc and other important fission products such as Zr, Mo, and Rh. More extensive testing is required to adequately predict flowsheet behavior for these variances within the fission products.

  8. Fission product behavior in the Peach Bottom and Fort St. Vrain HTGRs

    SciTech Connect

    Hanson, D.L.; Baldwin, N.L.; Strong, D.E.

    1980-11-01

    Actual operating data from Peach Bottom and Fort St. Vrain were compared with code predictions to assess the validity of the methods used to predict the behavior of fission products in the primary coolant circuit. For both reactors the measured circuit activities were significantly below design values, and the observations generally verify the codes used for large HTGR design.

  9. SEPARATION OF FISSION PRODUCT VALUES FROM THE HEXAVALENT PLUTONIUM BY CARRIER PRECIPITATION

    DOEpatents

    Davies, T.H.

    1959-12-15

    An improved precipitation of fission products on bismuth phosphate from an aqueous mineral acid solution also containing hexavalent plutonium by incorporating, prior to bismuth phosphate precipitation, from 0.05 to 2.5 grams/ liter of zirconium phosphate, niobium oxide. and/or lanthanum fluoride is described. The plutonium remains in solution.

  10. An innovative acoustic sensor for first in-pile fission gas release determination - REMORA 3 experiment

    SciTech Connect

    Rosenkrantz, E.; Ferrandis, J. Y.; Augereau, F.; Lambert, T.; Fourmentel, D.; Tiratay, X.

    2011-07-01

    A fuel rod has been instrumented with a new design of an acoustic resonator used to measure in a non destructive way the internal rod plenum gas mixture composition. This ultrasonic sensor has demonstrated its ability to operate in pile during REMORA 3 irradiation experiment carried out in the OSIRIS Material Testing Reactor (CEA Saclay, France). Due to very severe experimental conditions such as temperature rising up to 150 deg.C and especially, high thermal fluence level up to 3.5 10{sup 19} n.cm{sup 2}, the initial sensor gas speed of sound efficiency measurement was strongly reduced due to the irradiation effects on the piezo-ceramic properties. Nevertheless, by adding a differential signal processing method to the initial data analysis procedure validated before irradiation, the gas resonance peaks were successfully extracted from the output signal. From these data, the molar fractions variations of helium and fission gas were measured from an adapted Virial state equation. Thus, with this sensor, the kinetics of gas release inside fuel rods could be deduced from the in-pile measurements and specific calculations. These data will also give information about nuclear reaction effect on piezo-ceramics sensor under high neutron and gamma flux. (authors)

  11. Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

    SciTech Connect

    Blaise Collin

    2014-09-01

    Safety tests were conducted on fourteen 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 the safety tests, and the predicted values were compared with experimental results. Preliminary comparisons between PARFUME predictions and post-irradiation examination (PIE) results of the safety tests show different trends in the prediction of the fractional release depending on the species, and it leads to different conclusions regarding the diffusivities used in the modeling of fission product transport in TRISO-coated particles: • For silver, the diffusivity in silicon carbide (SiC) might be over-estimated by a factor of at least 102 to 103 at 1600°C and 1700°C, and at least 10 to 102 at 1800°C. The diffusivity of silver in uranium oxy-carbide (UCO) might also be over-estimated, but the available data are insufficient to allow definitive conclusions to be drawn. • For cesium, the diffusivity in UCO might be over-estimated by a factor of at least 102 to 103 at 1600°C, 105 at 1700°C, and 103 at 1800°C. The diffusivity of cesium in SiC might also over-estimated, by a factor of 10 at 1600°C and 103 at 1700°C, based upon the comparisons between calculated and measured release fractions from intact particles. There is no available estimate at 1800°C since all the compacts heated up at 1800°C contain particles with failed SiC layers whose release dominates the release from intact particles. • For strontium, the diffusivity in SiC might be over-estimated by a factor of 10 to 102 at 1600 and 1700°C, and 102 to 103 at 1800°C. These

  12. STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment

    NASA Astrophysics Data System (ADS)

    Leng, B.; van Rooyen, I. J.; Wu, Y. Q.; Szlufarska, I.; Sridharan, K.

    2016-07-01

    Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd2Si2U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously.

  13. Application of adjusted data in calculating fission-product decay energies and spectra

    NASA Astrophysics Data System (ADS)

    George, D. C.; Labauve, R. J.; England, T. R.

    1982-06-01

    The code ADENA, which approximately calculates fussion-product beta and gamma decay energies and spectra in 19 or fewer energy groups from a mixture of U235 and Pu239 fuels, is described. The calculation uses aggregate, adjusted data derived from a combination of several experiments and summation results based on the ENDF/B-V fission product file. The method used to obtain these adjusted data and the method used by ADENA to calculate fission-product decay energy with an absorption correction are described, and an estimate of the uncertainty of the ADENA results is given. Comparisons of this approximate method are made to experimental measurements, to the ANSI/ANS 5.1-1979 standard, and to other calculational methods. A listing of the complete computer code (ADENA) is contained in an appendix. Included in the listing are data statements containing the adjusted data in the form of parameters to be used in simple analytic functions.

  14. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    SciTech Connect

    McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel; Wirth, Brian; Kennedy, Rory

    2014-04-07

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  15. FITPULS: a code for obtaining analytic fits to aggregate fission-product decay-energy spectra. [In FORTRAN

    SciTech Connect

    LaBauve, R.J.; George, D.C.; England, T.R.

    1980-03-01

    The operation and input to the FITPULS code, recently updated to utilize interactive graphics, are described. The code is designed to retrieve data from a library containing aggregate fine-group spectra (150 energy groups) from fission products, collapse the data to few groups (up to 25), and fit the resulting spectra along the cooling time axis with a linear combination of exponential functions. Also given in this report are useful results for aggregate gamma and beta spectra from the decay of fission products released from /sup 235/U irradiated with a pulse (10/sup -4/ s irradiation time) of thermal neutrons. These fits are given in 22 energy groups that are the first 22 groups of the LASL 25-group decay-energy group structure, and the data are expressed both as MeV per fission second and particles per fission second; these pulse functions are readily folded into finite fission histories. 65 figures, 11 tables.

  16. Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

    SciTech Connect

    Behafarid, F.; Shaver, D. R.; Bolotnov, I. A.; Jansen, K. E.; Antal, S. P.; Podowski, M. Z.

    2012-07-01

    The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)

  17. Exploratory study of fission product yields of neutron-induced fission of 235U , 238U , and 239Pu at 8.9 MeV

    NASA Astrophysics Data System (ADS)

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

    2015-06-01

    Using dual-fission chambers each loaded with a thick (200 -400 -mg /c m2) actinide target of 235 ,238U or 239Pu and two thin (˜10 -100 -μ g /c m2) reference foils of the same actinide, the cumulative yields of fission products ranging from 92Sr to 147Nd have been measured at En= 8.9 MeV . The 2H(d ,n ) 3He reaction provided the quasimonoenergetic neutron beam. The experimental setup and methods used to determine the fission product yield (FPY) are described, and results for typically eight high-yield fission products are presented. Our FPYs for 235U(n ,f ) , 238U(n ,f ) , and 239Pu(n ,f ) at 8.9 MeV are compared with the existing data below 8 MeV from Glendenin et al. [Phys. Rev. C 24, 2600 (1981), 10.1103/PhysRevC.24.2600], Nagy et al. [Phys. Rev. C 17, 163 (1978), 10.1103/PhysRevC.17.163], Gindler et al. [Phys. Rev. C 27, 2058 (1983), 10.1103/PhysRevC.27.2058], and those of Mac Innes et al. [Nucl. Data Sheets 112, 3135 (2011), 10.1016/j.nds.2011.11.009] and Laurec et al. [Nucl. Data Sheets 111, 2965 (2010), 10.1016/j.nds.2010.11.004] at 14.5 and 14.7 MeV, respectively. This comparison indicates a negative slope for the energy dependence of most fission product yields obtained from 235U and 239Pu , whereas for 238U the slope issue remains unsettled.

  18. Fission product transport and behavior during two postulated loss of flow transients in the air

    SciTech Connect

    Adams, J.P.; Carboneau, M.L.

    1991-12-31

    This document discusses fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradation, respectively) in the Advanced Test Reactor (ATR). These transients are designated ATR Transient LCPI5 (high-pressure) and LPP9 (low-pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be analyzed. A probabilistic risk assessment indicated that the probability of occurrence for these two transients is of the order of 10{sup {minus}5 }and 10{sup {minus}7} per reactor year for LCP15 and LPP9, respectively.

  19. Fission product transport and behavior during two postulated loss of flow transients in the air

    SciTech Connect

    Adams, J.P.; Carboneau, M.L.

    1991-01-01

    This document discusses fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradation, respectively) in the Advanced Test Reactor (ATR). These transients are designated ATR Transient LCPI5 (high-pressure) and LPP9 (low-pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be analyzed. A probabilistic risk assessment indicated that the probability of occurrence for these two transients is of the order of 10{sup {minus}5 }and 10{sup {minus}7} per reactor year for LCP15 and LPP9, respectively.

  20. High-power proton linac for transmuting the long-lived fission products in nuclear waste

    SciTech Connect

    Lawrence, G.P.

    1991-01-01

    High power proton linacs are being considered at Los Alamos as drivers for high-flux spallation neutron sources that can be used to transmute the troublesome long-lived fission products in defense nuclear waste. The transmutation scheme being studied provides a high flux (> 10{sup 16}/cm{sup 2}{minus}s) of thermal neutrons, which efficiently converts fission products to stable or short-lived isotopes. A medium-energy proton linac with an average beam power of about 110 MW can burn the accumulated Tc99 and I129 inventory at the DOE's Hanford Site within 30 years. Preliminary concepts for this machine are described. 3 refs., 5 figs., 2 tabs.

  1. Formation and characterization of fission-product aerosols under postulated HTGR accident conditions

    SciTech Connect

    Tang, I.N.; Munkelwitz, H.R.

    1982-07-01

    The paper presents the results of an experimental investigation on the formation mechanism and physical characterization of simulated nuclear aerosols that could likely be released during an HTGR core heat-up accident. Experiments were carried out in a high-temperature flow system consisting essentially of an inductively heated release source, a vapor deposition tube, and a filter assembly for collecting particulate matter. Simulated fission products Sr and Ba as oxides are separately impregnated in H451 graphite wafers and released at elevated temperatures into a dry helium flow. In the presence of graphite, the oxides are quantitatively reduced to metals, which subsequently vaporize at temperatures much lower than required for the oxides alone to vaporize in the absence of graphite. A substantial fraction of the released material is associated with particulate matter, which is collected on filters located downstream at ambient temperature. The release and transport of simulated fission product Ag as metal are also investigated.

  2. Fission products behaviour in UO2 submitted to nuclear severe accident conditions

    NASA Astrophysics Data System (ADS)

    Geiger, E.; Bès, R.; Martin, P.; Pontillon, Y.; Solari, P. L.; Salome, M.

    2016-05-01

    The objective of this work was to study the molybdenum chemistry in UO2 based materials, known as SIMFUELS. These materials could be used as an alternative to irradiated nuclear fuels in the study of fission products behaviour during a nuclear severe accident. UO2 samples doped with 12 stable isotopes of fission products were submitted to annealing tests in conditions representative to intermediate steps of severe accidents. Samples were characterized by SEM-EDS and XAS. It was found that Mo chemistry seems to be more complex than what is normally estimated by thermodynamic calculations: XAS spectra indicate the presence of Mo species such as metallic Mo, MoO2, MoO3 and Cs2MoO4.

  3. Fission product Pd-SiC interaction in irradiated coated particle fuels

    SciTech Connect

    Tiegs, T.N.

    1980-04-01

    Silicon carbide is the main barrier to fission product release from coated particle fuels. Consequently, degradation of the SiC must be minimized. Electron microprobe analysis has identified that palladium causes corrosion of the SiC in irradiated coated particles. Further ceramographic and electron microprobe examinations on irradiated particles with kernels ranging in composition from UO/sub 2/ to UC/sub 2/, including PuO/sub 2 -x/ and mixed (Th, Pu) oxides, and in enrichment from 0.7 to 93.0% /sup 235/U revealed that temperature is the major factor affecting the penetration rate of SiC by Pd. The effects of kernel composition, Pd concentration, other fission products, and SiC properties are secondary.

  4. SELECTIVE SEPARATION OF URANIUM FROM THORIUM, PROTACTINIUM AND FISSION PRODUCTS BY PEROXIDE DISSOLUTION METHOD

    DOEpatents

    Seaborg, G.T.; Gofman, J.W.; Stoughton, R.W.

    1959-08-18

    A method is described for separating U/sup 233/ from thorium and fission products. The separation is effected by forming a thorium-nitric acid solution of about 3 pH, adding hydrogen peroxide to precipitate uranium and thorium peroxide, treating the peroxides with sodium hydroxide to selectively precipitate the uranium peroxide, and reacting the separated solution with nitric acid to re- precipitate the uranium peroxide.

  5. METHOD OF SEPARATING URANIUM, PLUTONIUM AND FISSION PRODUCTS BY BROMINATION AND DISTILLATION

    DOEpatents

    Jaffey, A.H.; Seaborg, G.T.

    1958-12-23

    The method for separation of plutonium from uranium and radioactive fission products obtained by neutron irradiation of uranlum consists of reacting the lrradiated material with either bromine, hydrogen bromide, alumlnum bromide, or sulfur and bromine at an elevated temperature to form the bromides of all the elements, then recovering substantlally pure plutonium bromide by dlstillatlon in combinatlon with selective condensatlon at prescribed temperature and pressure.

  6. PROCESS FOR SEGREGATING URANIUM FROM PLUTONIUM AND FISSION-PRODUCT CONTAMINATION

    DOEpatents

    Ellison, C.V.; Runion, T.C.

    1961-06-27

    An aqueous nitric acid solution containing uranium, plutonium, and fission product values is contacted with an organic extractant comprised of a trialkyl phosphate and an organic diluent. The relative amounts of trialkyl phosphate and uranium values are controlled to achieve a concentration of uranium values in the organic extractant of at least 0.35 moles uranium per mole of trialkyl phosphate, thereby preferentially extracting uranium values into the organic extractant.

  7. Enhanced Generic Phase-field Model of Irradiation Materials: Fission Gas Bubble Growth Kinetics in Polycrystalline UO2

    SciTech Connect

    Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert O.; Gao, Fei; Sun, Xin

    2012-05-30

    Experiments show that inter-granular and intra-granular gas bubbles have different growth kinetics which results in heterogeneous gas bubble microstructures in irradiated nuclear fuels. A science-based model predicting the heterogeneous microstructure evolution kinetics is desired, which enables one to study the effect of thermodynamic and kinetic properties of the system on gas bubble microstructure evolution kinetics and morphology, improve the understanding of the formation mechanisms of heterogeneous gas bubble microstructure, and provide the microstructure to macroscale approaches to study their impact on thermo-mechanical properties such as thermo-conductivity, gas release, volume swelling, and cracking. In our previous report 'Mesoscale Benchmark Demonstration, Problem 1: Mesoscale Simulations of Intra-granular Fission Gas Bubbles in UO2 under Post-irradiation Thermal Annealing', we developed a phase-field model to simulate the intra-granular gas bubble evolution in a single crystal during post-irradiation thermal annealing. In this work, we enhanced the model by incorporating thermodynamic and kinetic properties at grain boundaries, which can be obtained from atomistic simulations, to simulate fission gas bubble growth kinetics in polycrystalline UO2 fuels. The model takes into account of gas atom and vacancy diffusion, vacancy trapping and emission at defects, gas atom absorption and resolution at gas bubbles, internal pressure in gas bubbles, elastic interaction between defects and gas bubbles, and the difference of thermodynamic and kinetic properties in matrix and grain boundaries. We applied the model to simulate gas atom segregation at grain boundaries and the effect of interfacial energy and gas mobility on gas bubble morphology and growth kinetics in a bi-crystal UO2 during post-irradiation thermal annealing. The preliminary results demonstrate that the model can produce the equilibrium thermodynamic properties and the morphology of gas bubbles at

  8. Mechanistic determination of fission product releases for a Mark III BWR plant

    SciTech Connect

    Ludewig, H.; Yu, W.S.; Jaung, R.; Pratt, W.T.

    1986-01-01

    During the review of the GESSAR-II PRA by the Nuclear Regulatory Commission (NRC) and their contractors at Brookhaven National Laboratory (BNL) it was necessary to reanalyze potential fission product releases to the environment for several core meltdown accident sequences. The reanalysis was performed at BNL in two stages. The first stage was carried out prior to detailed mechanistic models were available at BNL and consisted of a sensitivity analysis using the MARCH and CORRAL computer codes. The effects of uncertainties in primary system retention, suppression pool scrubbing and core/concrete interactions on fission product release were handled by varying inputs to the MARCH/CORRAL codes. In this paper we outline the second stage of fission product release calculations, which was based on a system of codes developed under sponsorship of the Accident Source Term Program Office (ASTPO), USNRC. A comparison will be made between the range of source terms calculated by the first approach and the point estimate calculations provided by the more mechanistic codes. 8 refs., 2 tabs.

  9. Linear Free Energy Correlations for Fission Product Release from the Fukushima-Daiichi Nuclear Accident

    SciTech Connect

    Abrecht, David G.; Schwantes, Jon M.

    2015-03-03

    This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes, et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the source of the radionuclides to be from active reactors rather than the spent fuel pool. Linear correlations of the form ln χ = -α (ΔGrxn°(TC))/(RTC)+β were obtained between the deposited concentration and the reduction potential of the fission product oxide species using multiple reduction schemes to calculate ΔG°rxn(TC). These models allowed an estimate of the upper bound for the reactor temperatures of TC between 2130 K and 2220 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, 151Sm through atmospheric venting and releases during the first month following the accident were performed, and indicate large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores.

  10. Analysis of fission product revaporization in a BWR reactor cooling system during a station blackout accident

    SciTech Connect

    Yang, J.W.; Schmidt, E.; Cazzoli, E.; Khatib-Rahbar, M.

    1988-01-01

    This report presents a preliminary analysis of fission product revaporization in the Reactor Cooling System (RCS) after the vessel failure. The station blackout transient for BWR Mark I Power Plant is considered. The TRAPMELT3 models of evaporization, chemisorption, and the decay heating of RCS structures and gases are adopted in the analysis. The RCS flow models based on the density-difference between the RCS and containment pedestal region are developed to estimate the RCS outflow which carries the revaporized fission product to the containment. A computer code called REVAP is developed for the analysis. The REVAP is incorporated with the MARCH, TRAPMELT3 and NAUA codes of the Source Term Code Pack Package (STCP). The NAUA code is used to estimate the impact of revaporization on environmental release. The results show that the thermal-hydraulic conditions between the RCS and the pedestal region are important factors determining the magnitude of revaporization and subsequent release of the volatile fission product. 8 figs., 1 tab.

  11. Linear free energy correlations for fission product release from the Fukushima-Daiichi nuclear accident.

    PubMed

    Abrecht, David G; Schwantes, Jon M

    2015-03-01

    This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the initial source of the radionuclides to the environment to be from active reactors rather than the spent fuel pool. Linear correlations of the form In χ = −α ((ΔGrxn°(TC))/(RTC)) + β were obtained between the deposited concentrations, and the reduction potentials of the fission product oxide species using multiple reduction schemes to calculate ΔG°rxn (TC). These models allowed an estimate of the upper bound for the reactor temperatures of TC between 2015 and 2060 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, and 151Sm through atmospheric venting during the first month following the accident were obtained, indicating that large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores.

  12. Fission Product Transport in TRISO Particle Layers under Operating and Off-Normal Conditions

    SciTech Connect

    Van der Ven, Anton; Was, Gary; Wang, Lumin; Taheri, Mitra

    2014-04-26

    The objective of this project is to determine the diffusivity and chemical behavior of key fission products (ag, Cs, I. Te, Eu and Sr) through SiC and PyC both thermally, under irradiation, and under stress using FP introduction techniques that avoid the pitfalls of past experiments. The experimental approach is to create thin PyC-SiC couples containing the fission product to be studied embedded in the PyC layer. These samples will then be subjected to high temperature exposures in a vacuum and also to irradiation at high temperature, and last, to irradiation under stress at high temperature. The PyC serves as a host layer, providing a means of placing the fission product close to the SiC without damaging the SiC layer by its introduction or losing the FP during heating. Experimental measurements of grain boundary structure and distribution (EBSD, HRTEM, APT) will be used in the modeling effort to determine the qualitative dependence of FP diffusion coefficients on grain boundary orientation, temperature and stress.

  13. Analysis of fission product revaporization in a BWR Reactor Coolant System during a station blackout accident

    SciTech Connect

    Yang, J.W.; Schmidt, E.; Cazzoli, E.; Khatib-Rahbar, M.

    1988-01-01

    This paper presents an analysis of fission product revaporization from the Reactor Coolant System (RCS) following the Reactor Pressure Vessel (RPV) failure. The station blackout accident in a BWR Mark I Power Plant was considered. The TRAPMELT3 models for vaporization, chemisorption, and the decay heating of RCS structures and gases were used and extended beyond the RPV failure in the analysis. The RCS flow models based on the density-difference or pressure-difference between the RCS and containment pedestal region were developed to estimate the RCS outflow which carries the revaporized fission product to the containment. A computer code called REVAP was developed for the analysis. The REVAP code was incorporated with the MARCH, TRAPMELT3 and NAUA codes from the Source Term Code Package (STCP) to estimate the impact of revaporization on environmental release. The results show that the thermal-hydraulic conditions between the RCS and the pedestal region are important factors in determining the magnitude of revaporization and subsequent release of the volatile fission product into the environment. 6 refs., 8 figs.

  14. Insights on fission products behaviour in nuclear severe accident conditions by X-ray absorption spectroscopy

    NASA Astrophysics Data System (ADS)

    Geiger, E.; Bès, R.; Martin, Ph; Pontillon, Y.; Ducros, G.; Solari, P. L.

    2016-04-01

    Many research programs have been carried out aiming to understand the fission products behaviour during a Nuclear Severe Accident. Most of these programs used highly radioactive irradiated nuclear fuel, which requires complex instrumentation. Moreover, the radioactive character of samples hinders an accurate chemical characterisation. In order to overcome these difficulties, SIMFUEL stand out as an alternative to perform complementary tests. A sample made of UO2 doped with 11 fission products was submitted to an annealing test up to 1973 K in reducing atmosphere. The sample was characterized before and after the annealing test using SEM-EDS and XAS at the MARS beam-line, SOLEIL Synchrotron. It was found that the overall behaviour of several fission products (such as Mo, Ba, Pd and Ru) was similar to that observed experimentally in irradiated fuels and consistent with thermodynamic estimations. The experimental approach presented in this work has allowed obtaining information on chemical phases evolution under nuclear severe accident conditions, that are yet difficult to obtain using irradiated nuclear fuel samples.

  15. Comparisons of Neutron Cross Sections and Isotopic Composition Calculations for Fission-Product Evaluations

    NASA Astrophysics Data System (ADS)

    Kim, Do Heon; Gil, Choong-Sup; Chang, Jonghwa; Lee, Yong-Deok

    2005-05-01

    The neutron absorption cross sections for 18 fission products evaluated within the framework of the KAERI (Korea Atomic Energy Research Institute)-BNL (Brookhaven National Laboratory) international collaboration have been compared with ENDF/B-VI.7. Also, the influence of the new evaluations on the isotopic composition calculations of the fission products has been estimated through the OECD/NEA burnup credit criticality benchmarks (Phase 1B) and the LWR/Pu recycling benchmarks. These calculations were performed by WIMSD-5B with the 69-group libraries prepared from three evaluated nuclear data libraries: ENDF/B-VI.7, ENDF/B-VI.8 including the new evaluations in the resonance region covering the thermal region, and the expected ENDF/B-VII including those in the upper resonance region up to 20 MeV. For Xe-131, the composition calculated with ENDF/B-VI.8 shows a maximum difference of 5.02% compared to ENDF/B-VI.7. However, the isotopic compositions of all the fission products calculated with the expected ENDF/B-VII show no differences when compared to ENDF/B-VI.7 for the thermal reactor benchmark cases.

  16. Fission Product Separation from Pyrochemical Electrolyte by Cold Finger Melt Crystallization

    SciTech Connect

    Versey, Joshua R.

    2013-08-01

    This work contributes to the development of pyroprocessing technology as an economically viable means of separating used nuclear fuel from fission products and cladding materials. Electrolytic oxide reduction is used as a head-end step before electrorefining to reduce oxide fuel to metallic form. The electrolytic medium used in this technique is molten LiCl-Li2O. Groups I and II fission products, such as cesium (Cs) and strontium (Sr), have been shown to partition from the fuel into the molten LiCl-Li2O. Various approaches of separating these fission products from the salt have been investigated by different research groups. One promising approach is based on a layer crystallization method studied at the Korea Atomic Energy Research Institute (KAERI). Despite successful demonstration of this basic approach, there are questions that remain, especially concerning the development of economical and scalable operating parameters based on a comprehensive understanding of heat and mass transfer. This research explores these parameters through a series of experiments in which LiCl is purified, by concentrating CsCl in a liquid phase as purified LiCl is crystallized and removed via an argon-cooled cold finger.

  17. Linear free energy correlations for fission product release from the Fukushima-Daiichi nuclear accident.

    PubMed

    Abrecht, David G; Schwantes, Jon M

    2015-03-01

    This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the initial source of the radionuclides to the environment to be from active reactors rather than the spent fuel pool. Linear correlations of the form In χ = −α ((ΔGrxn°(TC))/(RTC)) + β were obtained between the deposited concentrations, and the reduction potentials of the fission product oxide species using multiple reduction schemes to calculate ΔG°rxn (TC). These models allowed an estimate of the upper bound for the reactor temperatures of TC between 2015 and 2060 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, and 151Sm through atmospheric venting during the first month following the accident were obtained, indicating that large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores. PMID:25675358

  18. Disposition of plutonium-239 via production of fission molybdenum-99.

    PubMed

    Mushtaq, A

    2011-04-01

    A heritage of physical consequences of the U.S.-Soviet arms race has accumulated, the weapons-grade plutonium (WPu), which will become excess as a result of the dismantlement of the nuclear weapons under the arms reduction agreements. Disposition of Pu has been proposed by mixing WPu with high-level radioactive waste with subsequent vitrification into large, highly radioactive glass logs or fabrication into mixed oxide fuel with subsequent irradiation in existing light water reactors. A potential option may be the production of medical isotope molybdenum-99 by using Pu-239 targets.

  19. Determination of fission gas release of spent nuclear fuel in puncturing test and in leaching experiments under anoxic conditions

    NASA Astrophysics Data System (ADS)

    González-Robles, E.; Metz, V.; Wegen, D. H.; Herm, M.; Papaioannou, D.; Bohnert, E.; Gretter, R.; Müller, N.; Nasyrow, R.; de Weerd, W.; Wiss, T.; Kienzler, B.

    2016-10-01

    During reactor operation the fission gases Kr and Xe are formed within the UO2 matrix of nuclear fuel. Their quantification is important to evaluate their impact on critical parameters regarding the fuel behaviour during irradiation and (long-term) interim storage, such as internal pressure of the fuel rod and fuel swelling. Moreover the content of Kr and Xe in the plenum of a fuel rod and their content in the UO2 fuel itself are widely used as indicators for the release properties of 129I, 137Cs, and other safety relevant radionuclides with respect to final disposal of spent nuclear fuel. The present study deals with the fission gas release from spent nuclear fuel exposed to simulated groundwater in comparison with the fission gas previously released to the fuel rod plenum during irradiation in reactor. In a unique approach we determined both the Kr and Xe inventories in the plenum by means of a puncturing test and in leaching experiments with a cladded fuel pellet and fuel fragments in bicarbonate water under 3.2 bar H2 overpressure. The fractional inventory of the fission gases released during irradiation into the plenum was (8.3 ± 0.9) %. The fraction of inventory of fission gases released during the leaching experiments was (17 ± 2) % after 333 days of leaching of the cladded pellet and (25 ± 2) % after 447 days of leaching of the fuel fragments, respectively. The relatively high release of fission gases in the experiment with fuel fragments was caused by the increased accessibility of water to the Kr and Xe occluded in the fuel.

  20. Fission product release and survivability of UN-kernel LWR TRISO fuel

    NASA Astrophysics Data System (ADS)

    Besmann, T. M.; Ferber, M. K.; Lin, H.-T.; Collin, B. P.

    2014-05-01

    A thermomechanical assessment of the LWR application of TRISO fuel with UN kernels was performed. Fission product release under operational and transient temperature conditions was determined by extrapolation from fission product recoil calculations and limited data from irradiated UN pellets. Both fission recoil and diffusive release were considered and internal particle pressures computed for both 650 and 800 μm diameter kernels as a function of buffer layer thickness. These pressures were used in conjunction with a finite element program to compute the radial and tangential stresses generated within a TRISO particle undergoing burnup. Creep and swelling of the inner and outer pyrolytic carbon layers were included in the analyses. A measure of reliability of the TRISO particle was obtained by computing the probability of survival of the SiC barrier layer and the maximum tensile stress generated in the pyrolytic carbon layers from internal pressure and thermomechanics of the layers. These reliability estimates were obtained as functions of the kernel diameter, buffer layer thickness, and pyrolytic carbon layer thickness. The value of the probability of survival at the end of irradiation was inversely proportional to the maximum pressure.

  1. Fission product release and survivability of UN-kernel LWR TRISO fuel

    SciTech Connect

    T. M. Besmann; M. K. Ferber; H.-T. Lin; B. P. Collin

    2014-05-01

    A thermomechanical assessment of the LWR application of TRISO fuel with UN kernels was performed. Fission product release under operational and transient temperature conditions was determined by extrapolation from fission product recoil calculations and limited data from irradiated UN pellets. Both fission recoil and diffusive release were considered and internal particle pressures computed for both 650 and 800 um diameter kernels as a function of buffer layer thickness. These pressures were used in conjunction with a finite element program to compute the radial and tangential stresses generated within a TRISO particle undergoing burnup. Creep and swelling of the inner and outer pyrolytic carbon layers were included in the analyses. A measure of reliability of the TRISO particle was obtained by computing the probability of survival of the SiC barrier layer and the maximum tensile stress generated in the pyrolytic carbon layers from internal pressure and thermomechanics of the layers. These reliability estimates were obtained as functions of the kernel diameter, buffer layer thickness, and pyrolytic carbon layer thickness. The value of the probability of survival at the end of irradiation was inversely proportional to the maximum pressure.

  2. Fundamental Studies of Irradiation-Induced Defect Formation and Fission Product Dynamics in Oxide Fuels

    SciTech Connect

    Stubbins, James

    2012-12-19

    The objective of this research program is to address major nuclear fuels performance issues for the design and use of oxide-type fuels in the current and advanced nuclear reactor applications. Fuel performance is a major issue for extending fuel burn-up which has the added advantage of reducing the used fuel waste stream. It will also be a significant issue with respect to developing advanced fuel cycle processes where it may be possible to incorporate minor actinides in various fuel forms so that they can be 'burned' rather than join the used fuel waste stream. The potential to fission or transmute minor actinides and certain long-lived fission product isotopes would transform the high level waste storage strategy by removing the need to consider fuel storage on the millennium time scale.

  3. Effects of microstructural constraints on the transport of fission products in uranium dioxide at low burnups

    NASA Astrophysics Data System (ADS)

    Lim, Harn Chyi; Rudman, Karin; Krishnan, Kapil; McDonald, Robert; Dickerson, Patricia; Gong, Bowen; Peralta, Pedro

    2016-08-01

    Diffusion of fission gases in UO2 is studied at low burnups, before bubble growth and coalescence along grain boundaries (GBs) become dominant, using a 3-D finite element model that incorporates actual UO2 microstructures. Grain boundary diffusivities are assigned based on crystallography with lattice and GB diffusion coupled with temperature to account for temperature gradients. Heterogeneity of GB properties and connectivity can induce regions where concentration is locally higher than without GB diffusion. These regions are produced by "bottlenecks" in the GB network because of lack of connectivity among high diffusivity GBs due to crystallographic constraints, and they can lead to localized swelling. Effective diffusivities were calculated assuming a uniform distribution of high diffusivity among GBs. Results indicate an increase over the bulk diffusivity with a clear grain size effect and that connectivity and properties of different GBs become important factors on the variability of fission product concentration at the microscale.

  4. General Point-Depletion and Fission Product Code System and Four-Group Fission Product Neutron Absorption Chain Data Library Generated from ENDF/B-IV for Thermal Reactors

    1981-12-01

    EPRI-CINDER calculates, for any specified initial fuel (actinide) description and flux or power history, the fuel and fission-product nuclide concentrations and associated properties. Other nuclide chains can also be computed with user-supplied libraries. The EPRI-CINDER Data Library (incorporating ENDF/B-IV fission-product processed 4-group cross sections, decay constants, absorption and decay branching fractions, and effective fission yields) is used in each constant-flux time step calculation and in time step summaries of nuclide decay rates and macroscopic absorptionmore » and barns-per-fission (b/f) absorption cross sections (by neutron group). User-supplied nuclide decay energy and multigroup-spectra data libraries may be attached to permit decay heating and decay-spectra calculations. An additional 12-chain library, explicitly including 27 major fission-product neutron absorbers and 4 fictitious nuclides, may be used to accurately calculate the aggregate macroscopic absorption buildup in fission products.« less

  5. Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

    DOEpatents

    Norman, Eric B [Oakland, CA; Prussin, Stanley G [Kensington, CA

    2009-05-05

    A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

  6. Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

    DOEpatents

    Norman, Eric B.; Prussin, Stanley G.

    2009-01-27

    A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

  7. Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

    DOEpatents

    Norman, Eric B.; Prussin, Stanley G.

    2009-01-06

    A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

  8. Isomer production ratios and the angular momentum distribution of fission fragments

    NASA Astrophysics Data System (ADS)

    Stetcu, I.; Talou, P.; Kawano, T.; Jandel, M.

    2013-10-01

    Latest generation fission experiments provide an excellent testing ground for theoretical models. In this contribution we compare the measurements for 235U(nth,f), obtained with the Detector for Advanced Neutron Capture Experiments (DANCE) calorimeter at Los Alamos Neutron Science Center (LANSCE), with our full-scale simulation of the primary fragment de-excitation, using the recently developed cgmf code, based on a Monte Carlo implementation of the Hauser-Feshbach theoretical model. We compute the isomer ratios as a function of the initial angular momentum of the fission fragments, for which no direct information exists. Comparison with the available experimental data allows us to determine the initial spin distribution. We also study the dependence of the isomer ratio on the knowledge of the low-lying discrete spectrum input for nuclear fission reactions, finding a high degree of sensitivity. Finally, in the same Hauser-Feshbach approach, we calculate the isomer production ratio for thermal neutron capture on stable isotopes, where the initial conditions (spin, excitation energy, etc.) are well understood. We find that with the current parameters involved in Hauser-Feshbach calculations, we obtain up to a factor of 2 deviation from the measured isomer ratios.

  9. Seminar on Fission VI

    NASA Astrophysics Data System (ADS)

    Wagemans, Cyriel; Wagemans, Jan; D'Hondt, Pierre

    2008-04-01

    Topical reviews. Angular momentum in fission / F. Gönnenwein ... [et al.]. The processes of fusion-fission and quasi-fission of heavy and super-heavy nuclei / M. G. Itkis ... [et al.] -- Fission cross sections and fragment properties. Minor-actinides fission cross sections and fission fragment mass yields via the surrogate reaction technique / B. Jurado ... [et al.]. Proton-induced fission on actinide nuclei at medium energy / S. Isaev ... [et al.]. Fission cross sections of minor actinides and application in transmutation studies / A. Letourneau ... [et al.]. Systematics on even-odd effects in fission fragments yields: comparison between symmetric and asymmetric splits / F. Rejmund, M Caamano. Measurement of kinetic energy distributions, mass and isotopic yields in the heavy fission products region at Lohengrin / A. Bail ... [et al.] -- Ternary fission. On the Ternary [symbol] spectrum in [symbol]Cf(sf) / M. Mutterer ... [et al.]. Energy degrader technique for light-charged particle spectroscopy at LOHENGRIN / A. Oberstedt, S. Oberstedt, D. Rochman. Ternary fission of Cf isotopes / S. Vermote ... [et al.]. Systematics of the triton and alpha particle emission in ternary fission / C. Wagemans, S. Vermote, O. Serot -- Neutron emission in fission. Scission neutron emission in fission / F.-J. Hambsch ... [et al.]. At and beyond the Scission point: what can we learn from Scission and prompt neutrons? / P. Talou. Fission prompt neutron and gamma multiplicity by statistical decay of fragments / S. Perez-Martin, S. Hilaire, E. Bauge -- Fission theory. Structure and fission properties of actinides with the Gogny force / H. Goutte ... [et al.]. Fission fragment properties from a microscopic approach / N. Dubray, H. Goutte, J.-P. Delaroche. Smoker and non-smoker neutron-induced fission rates / I. Korneev ... [et al.] -- Facilities and detectors. A novel 2v2E spectrometer in Manchester: new development in identification of fission fragments / I. Tsekhanovich ... [et al

  10. MELCOR 1.8.5 modeling aspects of fission product release, transport and deposition an assessment with recommendations.

    SciTech Connect

    Gauntt, Randall O.

    2010-04-01

    needed and should be included as part of a fuel oxidation model since fuel stoichiometry is a fundamental parameter in determining the vapor pressure of ruthenium oxides over the fuel. There is also a need to expand the MELCOR architecture for tracking fission product classes to allow for more speciation of fission products. An example is the formation of CsI and Cs{sub 2}MoO{sub 4} and potentially CsOH if all Mo is combined with Cs such that excess Cs exists in the fuel. Presently, MELCOR can track only one class combination (CsI) accurately, where excess Cs is assumed to be CsOH. Our recommended interim modifications map the CsOH (MELCOR Radionuclide Class 2) and Mo (Class 7) vapor pressure properties to Cs{sub 2}MoO{sub 4}, which approximates the desired formal class combination of Cs and Mo. Other extensions to handle properly iodine speciation from pool/gas chemistry are also needed.

  11. Deposition of fission and activation products after the Fukushima Dai-ichi nuclear power plant accident.

    PubMed

    Shozugawa, Katsumi; Nogawa, Norio; Matsuo, Motoyuki

    2012-04-01

    The Great Eastern Japan Earthquake on March 11, 2011, damaged reactor cooling systems at Fukushima Dai-ichi nuclear power plant. The subsequent venting operation and hydrogen explosion resulted in a large radioactive nuclide emission from reactor containers into the environment. Here, we collected environmental samples such as soil, plant species, and water on April 10, 2011, in front of the power plant main gate as well as 35 km away in Iitate village, and observed gamma-rays with a Ge(Li) semiconductor detector. We observed activation products ((239)Np and (59)Fe) and fission products ((131)I, (134)Cs ((133)Cs), (137)Cs, (110m)Ag ((109)Ag), (132)Te, (132)I, (140)Ba, (140)La, (91)Sr, (91)Y, (95)Zr, and (95)Nb). (239)Np is the parent nuclide of (239)Pu; (59)Fe are presumably activation products of (58)Fe obtained by corrosion of cooling pipes. The results show that these activation and fission products, diffused within a month of the accident.

  12. Deposition of fission and activation products after the Fukushima Dai-ichi nuclear power plant accident.

    PubMed

    Shozugawa, Katsumi; Nogawa, Norio; Matsuo, Motoyuki

    2012-04-01

    The Great Eastern Japan Earthquake on March 11, 2011, damaged reactor cooling systems at Fukushima Dai-ichi nuclear power plant. The subsequent venting operation and hydrogen explosion resulted in a large radioactive nuclide emission from reactor containers into the environment. Here, we collected environmental samples such as soil, plant species, and water on April 10, 2011, in front of the power plant main gate as well as 35 km away in Iitate village, and observed gamma-rays with a Ge(Li) semiconductor detector. We observed activation products ((239)Np and (59)Fe) and fission products ((131)I, (134)Cs ((133)Cs), (137)Cs, (110m)Ag ((109)Ag), (132)Te, (132)I, (140)Ba, (140)La, (91)Sr, (91)Y, (95)Zr, and (95)Nb). (239)Np is the parent nuclide of (239)Pu; (59)Fe are presumably activation products of (58)Fe obtained by corrosion of cooling pipes. The results show that these activation and fission products, diffused within a month of the accident. PMID:22266366

  13. Delayed-neutron branching ratios of precursors in the fission product region

    SciTech Connect

    Rudstam, G.; Aleklett, K.; Sihver, L. )

    1993-01-01

    Delayed-neutron branching ratios in the fission product region have been tabulated, and average values have been determined. In order to provide data complementary to published values an experiment covering the mass range 79-150 has been carried out at the OSIRIS isotope-separator on-line facility at Studsvik. This experiment has resulted in branching ratios for some precursors ([sup 84]Ge, [sup 133]Sn, and [sup 150]La) for which such data have not been reported before. In several other cases the new results are accurate than older determinations. 19 refs., 2 figs., 1 tab.

  14. Fission-product data analysis from actinide samples exposed in the Dounreay Prototype Fast Reactor

    SciTech Connect

    Murphy, B.D.; Dickens, J.K.; Walker, R.L.; Newton, T.D.

    1994-12-31

    Since 1979 a cooperative agreement has been in effect between the United States and the United Kingdom to investigate the irradiation of various actinide species placed in the core of the Dounreay Prototype Fast Reactor (PFR). The irradiated species were isotopes of thorium, protactinium, uranium, neptunium, plutonium, americium, and curium. A set of actinide samples (mg quantities) was exposed to about 490 effective full power days (EFPD) of reactor operations. The fission-product results are reported here. The actinide results will be report elsewhere.

  15. Design and Expected Performance of the AGR-1 Fission Product Monitoring System (FPMS)

    SciTech Connect

    John K. Hartwell; Dawn M. Scates

    2005-09-01

    The effluent from each test capsule of the AGR-1 experiment will be monitored by a detector system consisting of a gamma-ray spectrometer and a gross radiation detector. This collection of radiation measurement systems will be known as the AGR-1 Fission Product Monitoring System (FPMS). Proper design and functioning of the FPMS is critical to the success of the AGR-1 fuel test experiment.This document describes the AGR-1 FPMS and presents calculations indicating that this design will meet the pertinent test requirements.

  16. Ion exchange in the atomic energy industry with particular reference to actinide and fission product separation

    SciTech Connect

    Jenkins, I.L.

    1984-01-01

    Reviewed are some of the uses of ion exchange processes used by the nuclear industry for the period April, 1978 to April, 1983. The topics dealt with are: thorium, protactinium, uranium, neptunium, plutonium, americium, cesium and actinide-lanthanide separations; the higher actinides - Cm, Bk, Cf, Es and Fm; fission products; ion exchange in the geological disposal of radioactive waste. Consideration is given to safety in the use of ion exchangers and in safe methods of disposal of such materials. Full scale and pilot plant process descriptions are included as well as summaries of laboratory studies. 130 references.

  17. Fission product release phenomena during core melt accidents in metal fueled heavy water reactors

    SciTech Connect

    Ellison, P G; Hyder, M L; Monson, P R; Randolph, H W; Hagrman, D L; McClure, P R; Leonard, M T

    1990-01-01

    The phenomena that determine fission product release rates from a core melting accident in a metal-fueled, heavy water reactor are described in this paper. This information is obtained from the analysis of the current metal fuel experimental data base and from the results of analytical calculations. Experimental programs in place at the Savannah River Site are described that will provide information to resolve uncertainties in the data base. The results of the experiments will be incorporated into new severe accident computer codes recently developed for this reactor design. 47 refs., 4 figs.

  18. Progress towards the production of the 236gNp standard sources and competing fission fragment production

    NASA Astrophysics Data System (ADS)

    Larijani, C.; Pickford, O. L.; Collins, S. M.; Ivanov, P.; Jerome, S. M.; Keightley, J. D.; Pearce, A. K.; Regan, P. H.

    2015-11-01

    The isobaric distribution of fission residues produced following the bombardment of a natural uranium target with a beam of 25 MeV protons has been evaluated. Decay analysis of thirteen isobarically distinct fission residues were carried out using high-resolution γ-spectrometry at the UK National Physical Laboratory. Stoichiometric abundances were calculated via the determination of absolute activity concentrations associated with the longest-lived members of each isobaric chain. This technique was validated by computational modelling of likely sequential decay processes through an isobaric decay chain. The results were largely in agreement with previously published values for neutron bombardments on 238U at energies of 14 MeV. Higher yields of products with mass numbers A~110-130 were found, consistent with the increasing yield of these radionuclides as the bombarding energy is increased.

  19. Fission-product behaviour in irradiated TRISO-coated particles: Results of the HFR-EU1bis experiment and their interpretation

    NASA Astrophysics Data System (ADS)

    Barrachin, M.; Dubourg, R.; de Groot, S.; Kissane, M. P.; Bakker, K.

    2011-08-01

    It is important to understand fission-product (FP) and kernel micro-structure evolution in TRISO-coated fuel particles. FP behaviour, while central to severe-accident evaluation, impacts: evolution of the kernel oxygen potential governing in turn carbon oxidation (amoeba effect and pressurization); particle pressurization through fission-gas release from the kernel; and coating mechanical resistance via reaction with some FPs (Pd, Cs, Sr). The HFR-Eu1bis experiment irradiated five HTR fuel pebbles containing TRISO-coated UO 2 particles and went beyond current HTR specifications (e.g., central temperature of 1523 K). This study presents ceramographic and EPMA examinations of irradiated urania kernels and coatings. Significant evolutions of the kernel (grain structure, porosity, metallic-inclusion size, intergranular bubbles) as a function of temperature are shown. Results concerning FP migration are presented, e.g., significant xenon, caesium and palladium release from the kernel, molybdenum and ruthenium mainly present in metallic precipitates. The observed FP and micro-structural evolutions are interpreted and explanations proposed. The effect of high flux rate and high temperature on fission-gas behaviour, grain-size evolution and kernel swelling is discussed. Furthermore, Cs, Mo and Zr behaviour is interpreted in connection with oxygen-potential. This paper shows that combining state-of-the-art post-irradiation examination and state-of-the-art modelling fundamentally improves understanding of HTR fuel behaviour.

  20. Immobilization of fission products in low-temperature ceramic waste forms

    SciTech Connect

    Singh, D.; Wagh, A.S.; Tlustochowicz, M.; Mandalika, V.

    1997-01-01

    Over the last few years, Argonne National Laboratory has been developing room-temperature-setting chemically bonded phosphate ceramics (CBPCs) for use in solidifying and stabilizing low-level mixed wastes. The focus of this work is development of CBPCs for use with fission-product wastes generated from high-level waste (HLW) tank cleaning or other decontamination and decommissioning activities. The volatile fission products such as Tc, Cs, and Sr removed from HLW need to be disposed of in a low-temperature immobilization system. Specifically, this paper reports on the solidification and stabilization of separated {sup 99}Tc from Los Alamos National Laboratory`s complexation-elution process. Using rhenium as a surrogate form technetium, we fabricated CBPC waste forms by acid-base reactions. Dense and hard ceramic waste forms are produced in this process. The principal advantage of this technology is that the contaminants are immobilized by both chemical stabilization and subsequent microencapsulation of the reaction products. This paper reports the results of durability studies conducted on waste forms made with 35 wt.% waste loading. Standard leaching tests such as ANS 16.1 and PCT were conducted on the final waste forms. In addition, stability of the waste forms in aqueous environments was evaluated by long-term water-immersion tests.

  1. Monte Carlo Models for the Production of beta-delayed Gamma Rays Following Fission of Special Nuclear Materials

    SciTech Connect

    Pruet, J; Prussin, S; Descalle, M; Hall, J

    2004-02-03

    A Monte Carlo method for the estimation of {beta}-delayed {gamma}-ray spectra following fission is described that can accommodate an arbitrary time-dependent fission rate and photon collection history. The method invokes direct sampling of the independent fission yield distributions of the fissioning system, the branching ratios for decay of individual fission products and the spectral distributions for photon emission for each decay mode. Though computationally intensive, the method can provide a detailed estimate of the spectrum that would be recorded by an arbitrary spectrometer, and can prove useful in assessing the quality of evaluated data libraries, for identifying gaps in these libraries, etc. The method is illustrated by a first comparison of calculated and experimental spectra from decay of short-lived fission products following the reactions {sup 235}U(n{sub th}, f) and {sup 239}Pu(n{sub th}, f). For general purpose transport calculations, where detailed consideration of the large number of individual {gamma}-ray transitions in a spectrum may be unnecessary, it is shown that an accurate and simple parameterization of a {gamma}-ray source function can be obtained. These parametrizations should provide high-quality average spectral distributions that should prove useful in calculations describing photons escaping from thick attenuating media.

  2. Chemical thermodynamics of Cs and Te fission product interactions in irradiated LMFBR mixed-oxide fuel pins

    NASA Astrophysics Data System (ADS)

    Adamson, M. G.; Aitken, E. A.; Lindemer, T. B.

    1985-02-01

    A combination of fuel chemistry modelling and equilibrium thermodynamic calculations has been used to predict the atom ratios of Cs and Te fission products (Cs:Te) that find their way into the fuel-cladding interface region of irradiated stainless steel-clad mixed-oxide fast breeder reactor fuel pins. It has been concluded that the ratio of condensed, chemically-associated Cs and Te in the interface region,Čs:Te, which in turn determines the Te activity, is controlled by an equilibrium reaction between Cs 2Te and the oxide fuel, and that the value of Čs:Te is, depending on fuel 0:M, either equal to or slightly less than 2:1. Since Cs and Te fission products are both implicated as causative agents in FCCI (fission product-assisted inner surface attack of stainless steel cladding) and in FPLME (fission product-assisted liquid metal embrittlement of AISI-Type 316), the observed out-of-pile Cs:Te thresholds for FCCI (4˜:1) and FPLME (2˜:1) have been rationalized in terms of Cs:Te thermochemistry and phase equilibria. Also described in the paper is an updated chemical evolution model for reactive/volatile fission product behavior in irradiated oxide pins.

  3. Towards the inclusion of open fabrication porosity in a fission gas release model

    NASA Astrophysics Data System (ADS)

    Claisse, Antoine; Van Uffelen, Paul

    2015-11-01

    A model is proposed for fission product release in oxide fuels that takes into account the open porosity in a mechanistic manner. Its mathematical framework, assumptions and limitations are presented. It is based on the model for open porosity in the sintering process of crystalline solids. More precisely, a grain is represented by a tetrakaidecahedron and the open porosity is represented by a continuous cylinder along the grain edges. It has been integrated in the TRANSURANUS fuel performance code and applied to the first case of the first FUMEX project as well as to neptunium and americium containing pins irradiated during the SUPERFACT experiment and in the JOYO reactor. The results for LWR and FBR fuels are consistent with the experimental data and the predictions of previous empirical models when the thermal mechanisms are the main drivers of the release, even without using a fitting parameter. They also show a different but somewhat expected behaviour when very high porosity fuels are irradiated at a very low burn-up and at low temperature.

  4. EIA's Natural Gas Production Data

    EIA Publications

    2009-01-01

    This special report examines the stages of natural gas processing from the wellhead to the pipeline network through which the raw product becomes ready for transportation and eventual consumption, and how this sequence is reflected in the data published by the Energy Information Administration (EIA).

  5. Krypton and xenon in Apollo 14 samples - Fission and neutron capture effects in gas-rich samples

    NASA Technical Reports Server (NTRS)

    Drozd, R.; Hohenberg, C.; Morgan, C.

    1975-01-01

    Gas-rich Apollo 14 breccias and trench soil are examined for fission xenon from the decay of the extinct isotopes Pu-244 and I-129, and some samples have been found to have an excess fission component which apparently was incorporated after decay elsewhere and was not produced by in situ decay. Two samples have excess Xe-129 resulting from the decay of I-129. The excess is correlated at low temperatures with excess Xe-128 resulting from neutron capture on I-127. This neutron capture effect is accompanied by related low-temperature excesses of Kr-80 and Kr-82 from neutron capture on the bromine isotopes. Surface correlated concentrations of iodine and bromine are calculated from the neutron capture excesses.

  6. Diffusion modeling of fission product release during depressurized core conduction cooldown conditions

    SciTech Connect

    Martin, R.C.

    1990-01-01

    A simple model for diffusion through the silicon carbide layer of TRISO particles is applied to the data for accident condition testing of fuel spheres for the High-Temperature Reactor program of the Federal Republic of Germany (FRG). Categorization of sphere release of {sup 137}Cs based on fast neutron fluence permits predictions of release with an accuracy comparable to that of the US/FRG accident condition fuel performance model. Calculations are also performed for {sup 85}Kr, {sup 90}Sr, and {sup 110m}Ag. Diffusion of cesium through SiC suggests that models of fuel failure should consider fuel performance during repeated accident condition thermal cycling. Microstructural considerations in models in fission product release are discussed. The neutron-induced segregation of silicon within the SiC structure is postulated as a mechanism for enhanced fission product release during accident conditions. An oxygen-enhanced SiC decomposition mechanism is also discussed. 12 refs., 11 figs., 2 tabs.

  7. Influence of SiC grain boundary character on fission product transport in irradiated TRISO fuel

    NASA Astrophysics Data System (ADS)

    Lillo, T. M.; van Rooyen, I. J.

    2016-05-01

    In this study, the fission product precipitates at silicon carbide grain boundaries from an irradiated TRISO particle were identified and correlated with the associated grain boundary characteristics. Precession electron diffraction in the transmission electron microscope provided the crystallographic information needed to identify grain boundary misorientation and boundary type (i.e., low angle, random high angle or coincident site lattice (CSL)-related). The silicon carbide layer was found to be composed mainly of twin boundaries and small fractions of random high angle and low angle grain boundaries. Most fission products were found at random, high-angle grain boundaries, with small fractions at low-angle and CSL-related grain boundaries. Palladium (Pd) was found at all types of grain boundaries while Pd-uranium and Pd-silver precipitates were only associated with CSL-related and random, high-angle grain boundaries. Precipitates containing only Ag were found only at random, high-angle grain boundaries, but not at low angle or CSL-related grain boundaries.

  8. Partitioning of selected fission products from irradiated oxide fuel induced by thermal treatment

    NASA Astrophysics Data System (ADS)

    Shcherbina, Natalia; Kivel, Niko; Günther-Leopold, Ines

    2013-06-01

    The release of fission products (FPs) from spent nuclear fuel (SNF) has been studied as a function of the temperature and redox conditions. The present paper concerns essentially the high temperature separation of Cs and Sr from irradiated pressurized (PWR) and boiling water reactor (BWR) fuel of different burn-up levels with use of an in-house designed system for inductive vaporization (InVap). Using thermodynamic calculations with the Module of Fission Product Release (MFPR) code along with annealing experiments on SNF in the InVap it was shown that the speciation of Cs and Sr, hence their release behavior at high temperature, is sensitive to the redox conditions during thermal treatment. It was demonstrated that annealing conditions in the InVap can be adjusted in the way to promote the release of selected FPs without significant loss of the fuel matrix or actinides: complete release of Cs and I was achieved during treatment of irradiated fuel at 1800 °C under reducing atmosphere (0.7% H2/Ar mixture). The developed partitioning procedure can be used for the SNF pretreatment as an advanced head-end step in the hydrometallurgical or pyrochemical reprocessing technology.

  9. High-level waste glass field burial test: leaching and migration of fission products

    SciTech Connect

    Melnyk, T.W.; Johnson, L.H.; Walton, F.B.

    1984-01-01

    In June 1960, 25 nepheline syenite-based glass hemispheres containing the fission products /sup 137/Cs, /sup 90/Sr, /sup 144/Ce and /sup 106/Ru were buried below the water table in a sandy-soil aquifer at the Chalk River Nuclear Laboratories of Atomic Energy of Canada Limited. Measurements of soil and groundwater concentrations of /sup 90/Sr and /sup 137/Cs have been interpreted using non-equilibrium migration models to deduce the leaching history of the glass for these burial conditions. The leaching history derived from the field data has been compared to laboratory leaching of samples taken from a glass hemisphere retrieved in 1978, and also to pre-burial laboratory leaching of identical hemispheres. The time dependence of the leach rates observed for the buried specimens suggests that leaching is inhibited by the formation of a protective surface layer. The effect of the kinetic limitations of the fission-product/sandy-soil interactions is discussed with respect to the migration of /sup 90/Sr and /sup 137/Cs over a 20 year time scale. It is concluded that kinetically limited sorption by oxyhdroxides, rather than equilibrium ion exchange, controls the long-term migration of /sup 90/Sr. Cesium is initially rapidly bound to the micaceous fraction of the sand, but slow remobilization of /sup 137/Cs in particulate form is observed and is believed to be related to bacterial action.

  10. Alloy waste forms for metal fission products and actinides isolated by spent nuclear fuel treatment

    SciTech Connect

    McDeavitt, S.M.; Abraham, D.P.; Keiser, D.D. Jr.; Park, J.Y.

    1996-10-01

    Waste form alloys are being developed at Argonne National Laboratory for the disposal of remnant metallic wastes from an electrometallurgical process developed to treat spent nuclear fuel. This metal waste form consists of the fuel cladding (stainless steel or Zircaloy), noble metal fission products (e.g., Ru, Pd, Mo and Tc), and other metallic wastes. The main constituents of the metal waste stream are the cladding hulls (85 to 90 wt%); using the hulls as the dominant alloying component minimizes the overall waste volume as compared to vitrification or metal encapsulation. Two nominal compositions for the waste form are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels and (2) zirconium-8 wt% stainless steel for Zircaloy-clad fuels. The noble metal fission products are the primary source of radiation in the metal waste form. However, inclusion of actinides in the metal waste form is being investigated as an option for interim or ultimate storage. Simulated waste form alloys were prepared and analyzed to determine the baseline alloy microstructures and the microstructural distribution of noble metals and actinides. Corrosion tests of the metal waste form alloys indicate that they are highly resistant to corrosion.

  11. Investigation of Fission Product Transport into Zeolite-A for Pyroprocessing Waste Minimization

    SciTech Connect

    James R. Allensworth; Michael F. Simpson; Man-Sung Yim; Supathorn Phongikaroon

    2013-02-01

    Methods to improve fission product salt sorption into zeolite-A have been investigated in an effort to reduce waste associated with the electrochemical treatment of spent nuclear fuel. It was demonstrated that individual fission product chloride salts were absorbed by zeolite-A in a solid-state process. As a result, recycling of LiCl-KCl appears feasible via adding a zone-freezing technique to the current treatment process. Ternary salt molten-state experiments showed the limiting kinetics of CsCl and SrCl2 sorption into the zeolite. CsCl sorption occurred rapidly relative to SrCl2 with no observed dependence on zeolite particle size, while SrCl2 sorption was highly dependent on particle size. The application of experimental data to a developed reaction-diffusion-based sorption model yielded diffusivities of 8.04 × 10-6 and 4.04 × 10-7 cm2 /s for CsCl and SrCl2, respectively. Additionally, the chemical reaction term in the developed model was found to be insignificant compared to the diffusion term.

  12. METHOD FOR SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS BY SOLVENT EXTRACTION

    DOEpatents

    Seaborg, G.T.; Blaedel, W.J.; Walling, M.T. Jr.

    1960-08-23

    A process is given for separating from each other uranium, plutonium, and fission products in an aqueous nitric acid solution by the so-called Redox process. The plutonium is first oxidized to the hexavalent state, e.g., with a water-soluble dichromate or sodium bismuthate, preferably together with a holding oxidant such as potassium bromate. potassium permanganate, or an excess of the oxidizing agent. The solution is then contacted with a water-immiscible organic solvent, preferably hexone. whereby uranium and plutonium are extracted while the fission products remain in the aqueous solution. The separated organic phase is then contacted with an aqueous solution of a reducing agent, with or without a holding reductant (e.g., with a ferrous salt plus hydrazine or with ferrous sulfamate), whereby plutonium is reduced to the trivalent state and back- extracted into the aqueous solution. The uranium may finally be back-extracted from the organic solvent (e.g., with a 0.1 N nitric acid).

  13. Breast cancer: evidence for a relation to fission products in the diet.

    PubMed

    Sternglass, E J; Gould, J M

    1993-01-01

    To establish the possible relation between breast cancer mortality and low doses of radiation due to fission products in the environment, the mortality rates in the nine census regions of the United States for the years 1984-1988 were correlated with the cumulative airborne releases from all the nuclear plants in each region for the period 1970-1987. A high correlation coefficient of 0.91 was obtained for a logarithmic dependence on the total releases, consistent with an indirect action via free-radical oxygen at very low dose rates, in contrast to a direct action on DNA at high dose rates, explaining the wide differences in risk per unit dose obtained in earlier studies. The recent temporal changes of breast cancer rates in the New York metropolitan area including nearby Connecticut, Westchester, and Long Island were examined in relation to the releases from nearby nuclear plants and found to be consistent with a dominant role of short-lived fission products in drinking water and fresh milk. The results support a major role for nuclear plant releases in industrial countries in the recent rises of breast and other forms of cancers not related to smoking, especially among older persons, and strongly support the need to replace nuclear reactors with more benign ways to generate electricity.

  14. Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

    SciTech Connect

    Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

    1986-05-01

    In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant.

  15. Possible production of actinide spontaneous fission activities in damped collisions of /sup 209/Bi+/sup 56/Fe

    SciTech Connect

    Viola, V.E. Jr.; Mignerey, A.C.; Breuer, H.; Wolf, K.L.; Glagola, B.G.; Wilcke, W.W.; Schroeder, W.U.; Huizenga, J.R.; Hilscher, D.; Birkelund, J.R.

    1980-07-01

    Results of recent nuclear reaction studies on the /sup 209/Bi+/sup 56/Fe system are analyzed in order to investigate the cross sections for production of spontaneously fissioning species with 90 < or = Z < or = 100. This reaction is nearly identical to those used as the basis for identification of new elements with Z=104 --107, reported to be spontaneous fission emitters. Evidence is presented which indicates that actinide elements are formed with primary yields of the order of millibarns in strongly damped collisions of 464 MeV /sup 56/Fe with /sup 209/Bi. Although strongly depleted by fission competition, final yields of possible spontaneous fission activities /sup 240en-dash244/Am/sup f/, /sup 234en-dash238/Cf, and /sup 242en-dash246/Fm are calculated to be of the order of nanobarns. These results suggest that the spontaneous fission activities observed in similar reactions may be due at least in part to fission decay of actinide elements rather than elements with Z=104 --107.

  16. The chemical state of fission products in oxide fuels at different stages of the nuclear fuel cycle

    SciTech Connect

    Kleykamp, H.

    1988-03-01

    A survey of work at the Kernforschungszentrum Karlsruhe is presented on the chemical state of selected fission products that are relevant in the fuel cycle of light water reactor (LWR) and fast breeder reactor fuels. The influence of fuel type and irradiation progress on the composition of the Mo-Tc-Ru-Rh-Pd fission product alloys precipitated in the oxide matrix is examined using the respective multicomponent phase diagrams. The kinetics of dissolution of these phases in nitric acid at the reprocessing stage is discussed. Composition and structure of the residues, and the reprecipitation phenomena from highly active waste (HAW), are elucidated. A second metamorphosis of the fission products is recognized during the vitrification process. The formation of Ru(Rh) oxide and Pd(Rh, U, Te) alloys in simulated vitrified HAW concentrate and in HAW concentrate from the reprocessing of irradiated LWR fuels in interpreted on the basis of heterogeneous equilibria.

  17. Review of ENDF/B-VI Fission-Product Cross Section

    SciTech Connect

    Wright, R.Q.

    1999-01-01

    the uncertainty in calculated results and provide a better interpretation of criticality safety margins. Thus, the thrust of the Nuclear Data Task is to obtain high-resolution data in the intermediate energy region and provide fits to the data that utilize the modern RM formalism and covariance information for subsequent use in criticality predictability applications. As a subtask of the Nuclear Data Task, this review of the fission-product cross sections has several objectives. The first objective is a general data status review at various levels for the some 200 fission products. The second objective is a more detailed investigation of the top 20 fission products with regard to thermal- and intermediate-energy capture and scatter cross sections. The third objective is to demonstrate the revision of ENDF/B evaluations utilizing new data and evaluation techniques for 13 fission products. The fourth objective is to make recommendations for improvements, both specific and general in nature.

  18. 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

    SciTech Connect

    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 in 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.

  19. Electron Microscopic Evaluation and Fission Product Identification of Irradiated TRISO Coated Particles from the AGR-1 Experiment: A Preliminary Review

    SciTech Connect

    IJ van Rooyen; DE Janney; BD Miller; PA DEmkowicz; J Riesterer

    2014-05-01

    Post-irradiation examination of coated particle fuel from the AGR-1 experiment is in progress at Idaho National Laboratory and Oak Ridge National Laboratory. In this paper a brief summary of results from characterization of microstructures in the coating layers of selected irradiated fuel particles with burnup of 11.3% and 19.3% FIMA will be given. The main objectives of the characterization were to study irradiation effects, fuel kernel porosity, layer debonding, layer degradation or corrosion, fission-product precipitation, grain sizes, and transport of fission products from the kernels across the TRISO layers. Characterization techniques such as scanning electron microscopy, transmission electron microscopy, energy dispersive spectroscopy, and wavelength dispersive spectroscopy were used. A new approach to microscopic quantification of fission-product precipitates is also briefly demonstrated. Microstructural characterization focused on fission-product precipitates in the SiC-IPyC interface, the SiC layer and the fuel-buffer interlayer. The results provide significant new insights into mechanisms of fission-product transport. Although Pd-rich precipitates were identified at the SiC-IPyC interlayer, no significant SiC-layer thinning was observed for the particles investigated. Characterization of these precipitates highlighted the difficulty of measuring low concentrations of Ag in precipitates with significantly higher concentrations of Pd and U. Different approaches to resolving this problem are discussed. An initial hypothesis is provided to explain fission-product precipitate compositions and locations. No SiC phase transformations were observed and no debonding of the SiC-IPyC interlayer as a result of irradiation was observed for the samples investigated. Lessons learned from the post-irradiation examination are described and future actions are recommended.

  20. Electron microscopic evaluation and fission product identification of irradiated TRISO coated particles from the AGR-1 experiment: A preliminary Study

    SciTech Connect

    I J van Rooyen; D E Janney; B D Miller; J L Riesterer; P A Demkowicz

    2012-10-01

    ABSTRACT Post-irradiation examination of coated particle fuel from the AGR-1 experiment is in progress at Idaho National Laboratory and Oak Ridge National Laboratory. In this presentation a brief summary of results from characterization of microstructures in the coating layers of selected irradiated fuel particles with burnup of 11.3% and 19.3% FIMA will be given. The main objective of the characterization were to study irradiation effects, fuel kernel porosity, layer debonding, layer degradation or corrosion, fission-product precipitation, grain sizes, and transport of fission products from the kernels across the TRISO layers. Characterization techniques such as scanning electron microscopy, transmission electron microscopy, energy dispersive spectroscopy, and wavelength dispersive spectroscopy were used. A new approach to microscopic quantification of fission-product precipitates is also briefly demonstrated. The characterization emphasized fission-product precipitates in the SiC-IPyC interface, SiC layer and the fuel-buffer interlayer, and provided significant new insights into mechanisms of fission-product transport. Although Pd-rich precipitates were identified at the SiC-IPyC interlayer, no significant SiC-layer thinning was observed for the particles investigated. Characterization of these precipitates highlighted the difficulty of measuring low concentration Ag in precipitates with significantly higher concentrations of contain Pd and U. Different approaches to resolving this problem are discussed. Possible microstructural differences between particles with high and low releases of Ag particles are also briefly discussed, and an initial hypothesis is provided to explain fission-product precipitate compositions and locations. No SiC phase transformations or debonding of the SiC-IPyC interlayer as a result of irradiation were observed. Lessons learned from the post-irradiation examination are described and future actions are recommended.

  1. New antineutrino energy spectra predictions from the summation of beta decay branches of the fission products.

    PubMed

    Fallot, M; Cormon, S; Estienne, M; Algora, A; Bui, V M; Cucoanes, A; Elnimr, M; Giot, L; Jordan, D; Martino, J; Onillon, A; Porta, A; Pronost, G; Remoto, A; Taín, J L; Yermia, F; Zakari-Issoufou, A-A

    2012-11-16

    In this Letter, we study the impact of the inclusion of the recently measured beta decay properties of the (102;104;105;106;107)Tc, (105)Mo, and (101)Nb nuclei in an updated calculation of the antineutrino energy spectra of the four fissible isotopes (235,238)U and (239,241)Pu. These actinides are the main contributors to the fission processes in pressurized water reactors. The beta feeding probabilities of the above-mentioned Tc, Mo, and Nb isotopes have been found to play a major role in the γ component of the decay heat of (239)Pu, solving a large part of the γ discrepancy in the 4-3000 s range. They have been measured by using the total absorption technique, insensitive to the pandemonium effect. The calculations are performed by using the information available nowadays in the nuclear databases, summing all the contributions of the beta decay branches of the fission products. Our results provide a new prediction of the antineutrino energy spectra of (235)U, (239,241)Pu, and, in particular, (238)U for which no measurement has been published yet. We conclude that new total absorption technique measurements are mandatory to improve the reliability of the predicted spectra. PMID:23215477

  2. New Antineutrino Energy Spectra Predictions from the Summation of Beta Decay Branches of the Fission Products

    NASA Astrophysics Data System (ADS)

    Fallot, M.; Cormon, S.; Estienne, M.; Algora, A.; Bui, V. M.; Cucoanes, A.; Elnimr, M.; Giot, L.; Jordan, D.; Martino, J.; Onillon, A.; Porta, A.; Pronost, G.; Remoto, A.; Taín, J. L.; Yermia, F.; Zakari-Issoufou, A.-A.

    2012-11-01

    In this Letter, we study the impact of the inclusion of the recently measured beta decay properties of the Tc102;104;105;106;107, Mo105, and Nb101 nuclei in an updated calculation of the antineutrino energy spectra of the four fissible isotopes U235,238 and Pu239,241. These actinides are the main contributors to the fission processes in pressurized water reactors. The beta feeding probabilities of the above-mentioned Tc, Mo, and Nb isotopes have been found to play a major role in the γ component of the decay heat of Pu239, solving a large part of the γ discrepancy in the 4-3000 s range. They have been measured by using the total absorption technique, insensitive to the pandemonium effect. The calculations are performed by using the information available nowadays in the nuclear databases, summing all the contributions of the beta decay branches of the fission products. Our results provide a new prediction of the antineutrino energy spectra of U235, Pu239,241, and, in particular, U238 for which no measurement has been published yet. We conclude that new total absorption technique measurements are mandatory to improve the reliability of the predicted spectra.

  3. New antineutrino energy spectra predictions from the summation of beta decay branches of the fission products.

    PubMed

    Fallot, M; Cormon, S; Estienne, M; Algora, A; Bui, V M; Cucoanes, A; Elnimr, M; Giot, L; Jordan, D; Martino, J; Onillon, A; Porta, A; Pronost, G; Remoto, A; Taín, J L; Yermia, F; Zakari-Issoufou, A-A

    2012-11-16

    In this Letter, we study the impact of the inclusion of the recently measured beta decay properties of the (102;104;105;106;107)Tc, (105)Mo, and (101)Nb nuclei in an updated calculation of the antineutrino energy spectra of the four fissible isotopes (235,238)U and (239,241)Pu. These actinides are the main contributors to the fission processes in pressurized water reactors. The beta feeding probabilities of the above-mentioned Tc, Mo, and Nb isotopes have been found to play a major role in the γ component of the decay heat of (239)Pu, solving a large part of the γ discrepancy in the 4-3000 s range. They have been measured by using the total absorption technique, insensitive to the pandemonium effect. The calculations are performed by using the information available nowadays in the nuclear databases, summing all the contributions of the beta decay branches of the fission products. Our results provide a new prediction of the antineutrino energy spectra of (235)U, (239,241)Pu, and, in particular, (238)U for which no measurement has been published yet. We conclude that new total absorption technique measurements are mandatory to improve the reliability of the predicted spectra.

  4. Fission product release and microstructure changes during laboratory annealing of a very high burn-up fuel specimen

    NASA Astrophysics Data System (ADS)

    Hiernaut, J.-P.; Wiss, T.; Colle, J.-Y.; Thiele, H.; Walker, C. T.; Goll, W.; Konings, R. J. M.

    2008-07-01

    A commercial PWR fuel sample with a local burn-up of about 240 MWd/kgHM was annealed in a Knudsen cell mass spectrometer system with a heating rate of 10 K/min up to 2750 K at which temperature the sample was completely vaporized. The release of fission gases and fission products was studied as a function of temperature. In one of the runs the heating was interrupted successively at 900, 1500 and 1860 K and at each step a small fragment of the sample was examined by SEM and analysed by energy dispersive electron probe microanalysis. The release behaviour of volatile, gaseous and other less volatile fission products is presented and analysed with the EFFUS program and related to the structural changes of the fuel.

  5. Radioactive Beams from 252Cf Fission Using a Gas Catcher and an ECR Charge Breeder at ATLAS

    SciTech Connect

    Savard, Guy; Pardo, Richard C.; Moore, E. Frank; Hecht, Adam A.; Baker, Sam

    2005-03-15

    A proposed upgrade to the radioactive beam capability of the ATLAS facility has been proposed using 252Cf fission fragments thermalized and collected into a low-energy particle beam using a helium gas catcher. In order to reaccelerate these beams the ATLAS ECR-I will be reconfigured as a charge breeder source. A 1Ci 252Cf source is expected to provide sufficient yield to deliver beams of up to {approx}103 far from stability ions per second on target. A brief facility description and the expected performance information are provided in this report.

  6. Desulfurized gas production from vertical kiln pyrolysis

    DOEpatents

    Harris, Harry A.; Jones, Jr., John B.

    1978-05-30

    A gas, formed as a product of a pyrolysis of oil shale, is passed through hot, retorted shale (containing at least partially decomposed calcium or magnesium carbonate) to essentially eliminate sulfur contaminants in the gas. Specifically, a single chambered pyrolysis vessel, having a pyrolysis zone and a retorted shale gas into the bottom of the retorted shale zone and cleaned product gas is withdrawn as hot product gas near the top of such zone.

  7. Laboratory-Scale Bismuth Phosphate Extraction Process Simulation To Track Fate of Fission Products

    SciTech Connect

    Serne, R. JEFFREY; Lindberg, Michael J.; Jones, Thomas E.; Schaef, Herbert T.; Krupka, Kenneth M.

    2007-02-28

    Recent field investigation that collected and characterized vadose zone sediments from beneath inactive liquid disposal facilities at the Hanford 200 Areas show lower than expected concentrations of a long-term risk driver, Tc-99. Therefore laboratory studies were performed to re-create one of the three processes that were used to separate the plutonium from spent fuel and that created most of the wastes disposed or currently stored in tanks at Hanford. The laboratory simulations were used to compare with current estimates based mainly on flow sheet estimates and spotty historical data. Three simulations of the bismuth phosphate precipitation process show that less that 1% of the Tc-99, Cs-135/137, Sr-90, I-129 carry down with the Pu product and thus these isotopes should have remained within the metals waste streams that after neutralization were sent to single shell tanks. Conversely, these isotopes should not be expected to be found in the first and subsequent cycle waste streams that went to cribs. Measurable quantities (~20 to 30%) of the lanthanides, yttrium, and trivalent actinides (Am and Cm) do precipitate with the Pu product, which is higher than the 10% estimate made for current inventory projections. Surprisingly, Se (added as selenate form) also shows about 10% association with the Pu/bismuth phosphate solids. We speculate that the incorporation of some Se into the bismuth phosphate precipitate is caused by selenate substitution into crystal lattice sites for the phosphate. The bulk of the U daughter product Th-234 and Np-237 daughter product Pa-233 also associate with the solids. We suspect that the Pa daughter products of U (Pa-234 and Pa-231) would also co-precipitate with the bismuth phosphate induced solids. No more than 1 % of the Sr-90 and Sb-125 should carry down with the Pu product that ultimately was purified. Thus the current scheme used to estimate where fission products end up being disposed overestimates by one order of magnitude the

  8. Method and device for fabricating dispersion fuel comprising fission product collection spaces

    DOEpatents

    Shaber, Eric L; Fielding, Randall S

    2015-05-05

    A method of fabricating a nuclear fuel comprising a fissile material, one or more hollow microballoons, a phenolic resin, and metal matrix. The fissile material, phenolic resin and the one or more hollow microballoons are combined. The combined fissile material, phenolic resin and the hollow microballoons are heated sufficiently to form at least some fissile material carbides creating a nuclear fuel particle. The resulting nuclear fuel particle comprises one or more fission product collection spaces. In a preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by forming the fissile material into microspheres. The fissile material microspheres are then overcoated with the phenolic resin and microballoon. In another preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by overcoating the microballoon with the fissile material, and phenolic resin.

  9. Fission product partitioning in aerosol release from simulated spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Di Lemma, F. G.; Colle, J. Y.; Rasmussen, G.; Konings, R. J. M.

    2015-10-01

    Aerosols created by the vaporization of simulated spent nuclear fuel (simfuel) were produced by laser heating techniques and characterised by a wide range of post-analyses. In particular attention has been focused on determining the fission product behaviour in the aerosols, in order to improve the evaluation of the source term and consequently the risk associated with release from spent fuel sabotage or accidents. Different simulated spent fuels were tested with burn-up up to 8 at. %. The results from the aerosol characterisation were compared with studies of the vaporization process by Knudsen Effusion Mass Spectrometry and thermochemical equilibrium calculations. These studies permit an understanding of the aerosol gaseous precursors and the gaseous reactions taking place during the aerosol formation process.

  10. Fission Product Release from Molten U/Al Alloy Fuel: A Vapor Transpiration Model

    SciTech Connect

    Whitkop, P.G.

    2001-06-26

    This report describes the application of a vapor transportation model to fission product release data obtained for uranium/aluminum alloy fuel during early Oak Ridge fuel melt experiments. The Oak Ridge data validates the vapor transpiration model and suggests that iodine and cesium are released from the molten fuel surface in elemental form while tellurium and ruthenium are released as oxides. Cesium iodide is postulated to form in the vapor phase outside of the fuel matrix. Kinetic data indicates that cesium iodide can form from Cs atoms and diatomic iodine in the vapor phase. Temperatures lower than those capable of melting fuel are necessary in order to maintain a sufficient I2 concentration. At temperatures near the fuel melting point, cesium can react with iodine atoms to form CsI only on solid surfaces such as aerosols.

  11. IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

    SciTech Connect

    Gilles Youinou; Andrea Alfonsi

    2012-03-01

    This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis, the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  12. Electrochemistry of actinides and fission products in molten salts-Data review

    NASA Astrophysics Data System (ADS)

    Zhang, Jinsuo

    2014-04-01

    The thermodynamic and electrochemical properties of actinides and fission products in the molten salt determine the pyroprocessing separation performance. Extensive measurements have been carried out to provide fundamental data for evaluating the separation efficiency and technology feasibility of pyroprocessing although the technology has been very well developed in laboratory. The state of the art of fundamental data for substance or materials involved in pyropocessing will be reviewed in the present article. The available data will be summarized and reanalyzed. New correlations, which extend the available data to a broad range of applications, will be developed based on available data from different measurements. Further research topics on providing fundamental data that is needed for scaling the current laboratory technology to industrial applications are identified.

  13. Nuclear structure and shapes from prompt gamma ray spectroscopy of fission products

    SciTech Connect

    Ahmad, I.; Morss, L.R.; Durell, J.L.

    1996-10-01

    Many nuclear shape phenomena are predicted to occur in neutron-rich nuclei. The best source for the production of these nuclides is the spontaneous fission which produces practically hundreds of nuclides with yields of greater than 0.1 % per decay. Measurements of coincident gamma rays with large Ge arrays have recently been made to obtain information on nuclear structures and shapes of these neutron- rich nuclei. Among the important results that have been obtained from such measurements are octupole correlations in Ba isotopes, triaxial shapes in Ru nuclei, two-phonon vibrations in {sup 106}Mo and level lifetimes and quadrupole moments in Nd isotopes and A=100 nuclei. These data have been used to test theoretical models.

  14. Cherenkov light detection as a velocity selector for uranium fission products at intermediate energies

    NASA Astrophysics Data System (ADS)

    Yamaguchi, T.; Enomoto, A.; Kouno, J.; Yamaki, S.; Matsunaga, S.; Suzaki, F.; Suzuki, T.; Abe, Y.; Nagae, D.; Okada, S.; Ozawa, A.; Saito, Y.; Sawahata, K.; Kitagawa, A.; Sato, S.

    2014-12-01

    The in-flight particle separation capability of intermediate-energy radioactive ion (RI) beams produced at a fragment separator can be improved with the Cherenkov light detection technique. The cone angle of Cherenkov light emission varies as a function of beam velocity. This can be exploited as a velocity selector for secondary beams. Using heavy ion beams available at the HIMAC synchrotron facility, the Cherenkov light angular distribution was measured for several thin radiators with high refractive indices (n = 1.9 ~ 2.1). A velocity resolution of ~10-3 was achieved for a 56Fe beam with an energy of 500 MeV/nucleon. Combined with the conventional rigidity selection technique coupled with energy-loss analysis, the present method will enable the efficient selection of an exotic species from huge amounts of various nuclides, such as uranium fission products at the BigRIPS fragment separator located at the RI Beam Factory.

  15. Fission product plateout and liftoff in the MHTGR primary system: A review

    SciTech Connect

    Wichner, R.P. )

    1991-04-01

    A review is presented of the technical basis for predicting radioactivity release resulting from depressurization of an MHTGR primary system. Consideration is restricted to so called dry events with no involvement of the steam system. The various types of deposition mechanisms effective for iodine, cesium, strontium, and silver are discussed in terms of their chemical characteristics and the nature of the materials in the primary system. Emphasis is given to iodine behavior, including means for estimating the quantity available for release, the types of plateout locations in the primary system, and the effect of dust on distribution and release. The behavior of fission products cesium, strontium, and silver in such accidents is presented qualitatively. A major part of the review deals with expected dust levels, types, and transport. Available information on the level and nature of dust in the HTGR primary system is reviewed. A summary is presented of dust deposition and liftoff mechanisms. It was concluded that recent approaches to dust liftoff modeling, based on turbulent burst concepts for removal from surfaces, probably offer advantages over the current shear ratio approach. This study concludes that iodine releases from dry depressurization events are likely to be extremely low, on the order of millicuries, due to a predictably low degree of chemical desorption, a low degree of dust liftoff, and a low involvement of iodine with dust. It was also concluded that deposition mechanisms controlling the distribution of fission product material in the primary system, and hence also controlling the degree of liftoff, depend strongly on the chemical nature of the individual elements. Therefore contrary to the current practice, both plateout and liftoff models should reflect those unique chemical and physical properties. 56 refs., 16 figs., 23 tabs.

  16. Preliminary investigations on the use of uranium silicide targets for fission Mo-99 production

    SciTech Connect

    Cols, H.; Cristini, P.; Marques, R.

    1997-08-01

    The National Atomic Energy Commission (CNEA) of Argentine Republic owns and operates an installation for production of molybdenum-99 from fission products since 1985, and, since 1991, covers the whole national demand of this nuclide, carrying out a program of weekly productions, achieving an average activity of 13 terabecquerel per week. At present they are finishing an enlargement of the production plant that will allow an increase in the volume of production to about one hundred of terabecquerel. Irradiation targets are uranium/aluminium alloy with 90% enriched uranium with aluminium cladding. In view of international trends held at present for replacing high enrichment uranium (HEU) for enrichment values lower than 20 % (LEU), since 1990 the authors are in contact with the RERTR program, beginning with tests to adapt their separation process to new irradiation target conditions. Uranium silicide (U{sub 3}Si{sub 2}) was chosen as the testing material, because it has an uranium mass per volume unit, so that it allows to reduce enrichment to a value of 20%. CNEA has the technology for manufacturing miniplates of uranium silicide for their purposes. In this way, equivalent amounts of Molybdenum-99 could be obtained with no substantial changes in target parameters and irradiation conditions established for the current process with Al/U alloy. This paper shows results achieved on the use of this new target.

  17. Investigation of the Distribution of Fission Products Silver, Palladium and Cadmium in Neutron Irradiated SIC using a Cs Corrected HRTEM

    SciTech Connect

    I. J. van Rooyen; E. Olivier; J. H Neethlin

    2014-10-01

    Electron microscopy examinations of selected coated particles from the first advanced gas reactor experiment (AGR-1) at Idaho National Laboratory (INL) provided important information on fission product distribution and chemical composition. Furthermore, recent research using STEM analysis led to the discovery of Ag at SiC grain boundaries and triple junctions. As these Ag precipitates were nano-sized, high resolution transmission electron microscopy (HRTEM) examination was used to provide more information at the atomic level. This paper describes some of the first HRTEM results obtained by examining a particle from Compact 4-1-1, which was irradiated to an average burnup of 19.26% fissions per initial metal atom (FIMA), a time average, volume-averaged temperature of 1072°C; a time average, peak temperature of 1182°C and an average fast fluence of 4.13 x 1021 n/cm2. Based on gamma analysis, it is estimated that this particle may have released as much as 10% of its available Ag-110m inventory during irradiation. The HRTEM investigation focused on Ag, Pd, Cd and U due to the interest in Ag transport mechanisms and possible correlation with Pd, Ag and U previously found. Additionally, Compact 4-1-1 contains fuel particles fabricated with a different fuel carrier gas composition and lower deposition temperatures for the SiC layer relative to the Baseline fabrication conditions, which are expected to reduce the concentration of SiC defects resulting from uranium dispersion. Pd, Ag, and Cd were found to co-exist in some of the SiC grain boundaries and triple junctions whilst U was found to be present in the micron-sized precipitates as well as separately in selected areas at grain boundaries. This study confirmed the presence of Pd both at inter- and intragranular positions; in the latter case specifically at stacking faults. Small Pd nodules were observed at a distance of about 6.5 micron from the inner PyC/SiC interface.

  18. On the role of energy separated in fission process, excitation energy and reaction channels effects in the isomeric ratios of fission product 135Xe in photofission of actinide elements

    NASA Astrophysics Data System (ADS)

    Thiep, Tran Duc; An, Truong Thi; Cuong, Phan Viet; Vinh, Nguyen The; Mishinski, G. V.; Zhemenik, V. I.

    2016-07-01

    In this work we present the isomeric ratio of fission product 135Xe in the photo-fission of actinide elements 232Th, 233U and 237Np induced by end-point bremsstrahlung energies of 13.5, 23.5 and 25.0 MeV which were determined by the method of inert gaseous flow. The data were analyzed, discussed and compared with the similar data from literature to examine the role of energy separated in fission process, excitation energy and reaction channels effects.

  19. Atomic scale mobility of the volatile fission products Xe, Kr and I in cubic SiC.

    PubMed

    Cooper, M W D; Kelly, S; Bertolus, M

    2016-06-22

    The migration barriers for the vacancy-assisted migration of fission products in 3C-SiC are reported and analysed in the context of the five frequency model, which enables one to calculate an effective diffusion coefficient from elementary mechanisms. Calculations were carried out using the nudged elastic band method (NEB) with interatomic forces determined from density functional theory (DFT). Justification for treating vacancy-assisted fission product migration as limited to the FCC carbon sublattice is based on the stability of carbon vacancies, unfavourable silicon vacancy formation and the accommodation of fission products on the carbon sublattice. Results show that for most Fermi levels within the band gap the activation energy for I exceeds that of Xe which exceeds that of Kr. Results also indicate that activation energies are higher near the conduction edge, thus, implying that enhanced fission product retention can be achieved through n-type doping of 3C-SiC, which limits the availability of the migration mediating carbon vacancies. PMID:27282287

  20. Substrate preference of citrus naringenin rhamnosyltransferases and their application to flavonoid glycoside production in fission yeast.

    PubMed

    Ohashi, Takao; Hasegawa, Yuka; Misaki, Ryo; Fujiyama, Kazuhito

    2016-01-01

    Flavonoids, which comprise a large family of secondary plant metabolites, have received increased attention in recent years due to their wide range of features beneficial to human health. One of the most abundant flavonoid skeletons in citrus species is the flavanone naringenin, which is accumulated as glycosides containing terminal rhamnose (Rha) after serial glycosylation steps. The linkage type of Rha residues is a determining factor in the bitterness of the citrus fruit. Such Rha residues are attached by either an α1,2- or an α1,6-rhamnosyltransferase (1,2RhaT or 1,6RhaT). Although the genes encoding these RhaTs from pummelo (Citrus maxima) and orange (Citrus sinensis) have been functionally characterized, the details of the biochemical characterization, including the substrate preference, remain elusive due to the lack of availability of the UDP-Rha required as substrate. In this study, an efficient UDP-Rha in vivo production system using the engineered fission yeast expressing Arabidopsis thaliana rhamnose synthase 2 (AtRHM2) gene was constructed. The in vitro RhaT assay using the constructed UDP-Rha revealed that recombinant RhaT proteins (Cm1,2RhaT; Cs1,6RhaT; or Cm1,6RhaT), which were heterologously produced in fission yeast, catalyzed the rhamnosyl transfer to naringenin-7-O-glucoside as an acceptor. The substrate preference analysis showed that Cm1,2RhaT had glycosyl transfer activity toward UDP-xylose as well as UDP-Rha. On the other hand, Cs1,6RhaT and Cm1,6RhaT showed rhamnosyltransfer activity toward quercetin-3-O-glucoside in addition to naringenin-7-O-glucoside, indicating weak specificity toward acceptor substrates. Finally, naringin and narirutin from naringenin-7-O-glucoside were produced using the engineered fission yeast expressing the AtRHM2 and the Cm1,2RhaT or the Cs1,6RhaT genes as a whole-cell-biocatalyst.

  1. Interaction of fission products and SiC in TRISO fuel particles: a limiting HTGR design parameter

    SciTech Connect

    Stansfield, O.M.; Homan, F.J.; Simon, W.A.; Turner, R.F.

    1983-09-01

    The fuel particle system for the steam cycle cogeneration HTGR being developed in the US consists of 20% enriched UC/sub 0/./sub 3/O/sub 1/./sub 7/ and ThO/sub 2/ kernels with TRISO coatings. The reaction of fission products with the SiC coating is the limiting thermochemical coating failure mechanism affecting performance. The attack of the SiC by palladium (Pd) is considered the controlling reaction with systems of either oxide or carbide fuels. The lanthanides, such as cerium, neodymium, and praseodymium, also attack SiC in carbide fuel particles. In reactor design, the time-temperature relationships at local points in the core are used to calculate the depth of SiC-Pd reaction. The depth of penetration into the SiC during service varies with core power density, power distribution, outlet gas temperature, and fuel residence time. These parameters are adjusted in specifying the core design to avoid SiC coating failure.

  2. Simulation of radiation driven fission gas diffusion in UO2, ThO2 and PuO2

    DOE PAGES

    Cooper, Michael William D.; Stanek, Christopher Richard; Turnbull, James Anthony; Uberuaga, Blas P.; Andersson, David Anders

    2016-09-16

    Below 1000 K it is thought that fission gas diffusion in nuclear fuel during irradiation occurs through atomic mixing due to radiation damage. Here we present a molecular dynamics (MD) study of Xe, Kr, Th, U, Pu and O diffusion due to irradiation. It is concluded that the ballistic phase does not sufficiently account for the experimentally observed diffusion. Thermal spike simulations are used to confirm that electronic stopping remedies the discrepancy with experiment and the predicted diffusivities lie within the scatter of the experimental data. Here, our results predict that the diffusion coefficients are ordered such that D*0 >more » D*Kr > D*Xe > D*U. For all species >98.5% of diffusivity is accounted for by electronic stopping. Fission gas diffusivity was not predicted to vary significantly between ThO2, UO2 and PuO2, indicating that this process would not change greatly for mixed oxide fuels.« less

  3. Partitioning of fission products from irradiated nitride fuel using inductive vaporization

    SciTech Connect

    Shcherbina, N.; Kulik, D.A.; Kivel, N.; Potthast, H.D.; Guenther-Leopold, I.

    2013-07-01

    Irradiated nitride fuel (Pu{sub 0.3}Zr{sub 0.7})N fabricated at PSI in frame of the CONFIRM project and having a burn-up of 10.4 % FIMA (Fission per Initial Metal Atom) has been investigated by means of inductive vaporization. The study of thermal stability and release behavior of Pu, Am, Zr and fission products (FPs) was performed in a wide temperature range (up to 2300 C. degrees) and on different redox conditions. On-line monitoring by ICP-MS detected low nitride stability and significant loss of Pu and Am at T>1900 C. degrees during annealing under inert atmosphere (Ar). The oxidative pre-treatment of nitride fuel on air at 1000 C. degrees resulted in strong retention of Pu and Am in the solid, as well as of most FPs. Thermodynamic modelling of elemental speciation using GEM-Selektor v.3 code (Gibbs Energy Minimization Selektor), supported by a comprehensive literature review on thermodynamics of actinides and FPs, revealed a number of binary compounds of Cs, Mo, Te, Sr and Ba to occur in the solid. Speciation of some FPs in the fuel is discussed and compared to earlier results of electron probe microanalysis (EPMA). Predominant vapor species predicted by GEM-Selektor calculations were Pu(g), Am(g) and N{sub 2}. Nitrogen can be completely released from the fuel after complete oxidation at 1000 C. degrees. With regard to the irradiated nitride reprocessing technology, this result can have an important practical application as an alternative way for {sup 15}N recovery. (authors)

  4. Determination of Fission Product Yields of 235U, 238U and 239Pu for Neutron Energies from 0.5 to 14.8 MeV

    NASA Astrophysics Data System (ADS)

    Gooden, Matthew; Arnold, Charles; Becker, John; Bhatia, Chitra; Bhike, Megha; Fallin, Brent; Fowler, Malcolm; Howell, Calvin; Kelley, John; Stoyer, Mark; Tonchev, Anton; Tornow, Werner; Viera, David; Wilhelmy, Jerry

    2013-10-01

    A joint TUNL-LANL-LLNL collaboration has been formed to study the issue of possible energy dependences for fission product isotopes. Work has been carried out at the TUNL 10 MV Tandem accelerator which produces nearly mono-energetic neutrons via either 2H(d,n)3He,3H(d,n)4He, or 3H(p,n)3He reactions. Three dual fission ionization chambers dedicated to 235U, 238U and 239Pu thick target foils and thin monitor foils respectively, were exposed to the neutron beams. After irradiation, thick target foils were gamma counted over a period of 1-2 months and characteristic gamma rays from fission products were recorded using HPGe detectors at TUNL's low background counting area. Using the dual fission chambers fission product yields relative to total number of fissions were determined at a high precision of 2-3% as well as absolute fission product yields at a lower precision of 5-6%. Results will be presented for a number of fission product isotopes at 1.38, 4.6 and 14.8 MeV as well as preliminary results at 9 MeV.

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

    SciTech Connect

    Tipnis, S.V.; Campbell, J.M.; Couchell, G.P.; Li, S.; Nguyen, H.V.; Pullen, D.J.; Schier, W.A.; Seabury, E.H.; England, T.R.

    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 the 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}

  6. Energy Dependence of Neutron-Induced Fission Product Yields of 235U, 238U and 239Pu Between 0.5 and 14.8 MeV

    NASA Astrophysics Data System (ADS)

    Gooden, Matthew; Tornow, Werner; Tonchev, Anton; Vieira, Dave; Wilhelmy, Jerry; Arnold, Charles; Fowler, Malcolm; Stoyer, Mark

    2014-09-01

    Under a joint collaboration between TUNL-LANL-LLNL, a set of absolute fission product yield measurements have been performed. The energy dependence of a number of cumulative fission products between 0.5 and 14.8 MeV have been measured using quasi-monoenergetic neutron beams for three actinide targets, 235U, 238U and 239Pu, between 0.5 and 14.8 MeV. The FPYs were measured by a combination of activation utilizing specially designed dual-fission chambers and γ-ray counting. The dual-fission chambers are back-to-back ionization chambers encasing a target with thin deposits of the same target isotope in each chamber. This method allows for the direct measurement of the fission rate in the activation target with no reference to the fission cross-section, reducing uncertainties. γ-ray counting was performed on well-shield HPGe detectors over a period of 2 months per activation to properly identify fission products. Reported are absolute cumulative fission product yields for incident neutron energies of 0.5, 1.37, 2.4, 4.6 and 14.8 MeV.

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

    NASA Astrophysics Data System (ADS)

    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 88Kr and compared to previously tabulated values. 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.

  8. Fuel and fission product behaviour in early phases of a severe accident. Part II: Interpretation of the experimental results of the PHEBUS FPT2 test

    NASA Astrophysics Data System (ADS)

    Dubourg, R.; Barrachin, M.; Ducher, R.; Gavillet, D.; De Bremaecker, A.

    2014-10-01

    One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO2 fuel bundle and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 mm and 900 mm) of the test section previously reported are interpreted in the present paper. Solid state interactions between fuel and cladding have been compared with the characteristics of interaction identified in the previous separate-effect tests. Corium resulting from the interaction between fuel and cladding was formed. The uranium concentration in the corium is compared to analytical tests and a scenario for the corium formation is proposed. The analysis showed that, despite the rather low fuel burn up, the conditions of temperature and oxygen potential reached during the starvation phase are able to give an early very significant release fraction of caesium. A significant part (but not all) of the molybdenum was segregated at grain boundaries and trapped in metallic inclusions from which they were totally removed in the final part of the experiment. During the steam starvation phase, the conditions of oxygen potential were favourable for the formation of simple Ba and BaO chemical forms but the temperature was too low to provoke their volatility. This is one important difference with out-of-pile experiments such as VERCORS for which only a combination of high temperature and low oxygen potential induced a significant barium release. Finally another significant difference with analytical out-of-pile experiments comes from the formation of foamy zones due to the fission gas presence in FPT2-type experiments which give an additional possibility for the formation of stable fission product compounds.

  9. Primary system fission product release and transport: A state-of-the-art report to the committee on the safety of nuclear installations

    SciTech Connect

    Wright, A.L.

    1994-06-01

    This report presents a summary of the status of research activities associated with fission product behavior (release and transport) under severe accident conditions within the primary systems of water-moderated and water-cooled nuclear reactors. For each of the areas of fission product release and fission product transport, the report summarizes relevant information on important phenomena, major experiments performed, relevant computer models and codes, comparisons of computer code calculations with experimental results, and general conclusions on the overall state of the art. Finally, the report provides an assessment of the overall importance and knowledge of primary system release and transport phenomena and presents major conclusions on the state of the art.

  10. Fission Product Gamma-Ray Line Pairs Sensitive to Fissile Material and Neutron Energy

    SciTech Connect

    Marrs, R E; Norman, E B; Burke, J T; Macri, R A; Shugart, H A; Browne, E; Smith, A R

    2007-11-15

    The beta-delayed gamma-ray spectra from the fission of {sup 235}U, {sup 238}U, and {sup 239}Pu by thermal and near-14-MeV neutrons have been measured for delay times ranging from 1 minute to 14 hours. Spectra at all delay times contain sets of prominent gamma-ray lines with intensity ratios that identify the fissile material and distinguish between fission induced by low-energy or high-energy neutrons.

  11. Separation of the rare-earth fission product poisons from spent nuclear fuel

    DOEpatents

    Christian, Jerry D.; Sterbentz, James W.

    2016-08-30

    A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO.sub.2 and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO.sub.2 in the nuclear fuel, preferably to at least to 850.degree. C. for Th and up to 600.degree. C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.

  12. Hot beta particles in the lung: Results from dogs exposed to fission product radionuclides

    SciTech Connect

    Hahn, F.F.; Griffith, W.C.; Hobbs, C.H.

    1995-12-01

    The Chernobyl nuclear reactor accident resulted in the release of uranium dioxide fuel and fission product radionuclides into the environment with the fallout of respirable, highly radioactive particles that have been termed {open_quotes}hot beta particles.{close_quotes} There is concern that these hot beta particles (containing an average of 150-20,000 Bq/particle), when inhaled and deposited in the lung, may present an extraordinary hazard for the induction of lung cancer. We reviewed data from a group of studies in dogs exposed to different quantities of beta-emitting radionuclides with varied physical half-lives to determine if those that inhaled hot beta particles were at unusual risk for lung cancer. This analysis indicates that the average dose to the lung is adequate to predict biologic effects of lung cancer for inhaled beta-emitting radionuclides in the range of 5-50 Gy to the lung and with particle activities in the range of 0.10-50 Bq/particle.

  13. Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

    SciTech Connect

    Radulescu, Georgeta; Gauld, Ian C.; Ilas, Germina; Wagner, John C.

    2014-11-01

    This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application of the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.

  14. Electrochemical separation of actinides and fission products in molten salt electrolyte

    SciTech Connect

    Gay, R. L.; Grantham, L. F.; Fusselman, S. P.; Grimmett, D. L.; Roy, J. J.

    1995-09-15

    Molten salt electrochemical separation may be applied to accelerator-based conversion (ABC) and transmutation systems by dissolving the fluoride transport salt in LiCl-KCl eutectic solvent. The resulting fluoride-chloride mixture will contain small concentrations of fission product rare earths (La, Nd, Gd, Pr, Ce, Eu, Sm, and Y) and actinides (U, Np, Pu, Am, and Cm). The Gibbs free energies of formation of the metal chlorides are grouped advantageously such that the actinides can be deposited on a solid cathode with the majority of the rare earths remaining in the electrolyte. Thus, the actinides are recycled for further transmutation. Rockwell and its partners have measured the thermodynamic properties of the metal chlorides of interest (rare earths and actinides) and demonstrated separation of actinides from rare earths in laboratory studies. A model is being developed to predict the performance of a commercial electrochemical cell for separations starting with PUREX compositions. This model predicts excellent separation of plutonium and other actinides from the rare earths in metal-salt systems.

  15. Study on collection efficiency of fission products by spray: Experimental device and modelling

    SciTech Connect

    Ducret, D.; Roblot, D.; Vendel, J.; Billarand, Y.

    1997-08-01

    Consequences of an hypothetical overheating reactor accident in nuclear power plants can be limited by spraying cold water drops into containment building. The spray reduces the pressure and the temperature levels by condensation of steam and leads to the washout of fission products (aerosols and gaseous iodine). The present study includes a large program devoted to the evaluation of realistic washout rates. An experimental device (named CARAIDAS) was designed and built in order to determine the collection efficiency of aerosols and iodine absorption by drops with representative conditions of post-accident atmosphere. This experimental device is presented in the paper and more particularly: (1) the experimental enclosure in which representative thermodynamic conditions can be achieved, (2) the monosized drops generator, the drops diameter measurement and the drops collector, (3) the cesium iodide aerosols generator and the aerosols measurements. Modelling of steam condensation on drops aerosols collection and iodine absorption are described. First experimental and code results on drops and aerosols behaviour are compared. 8 refs., 18 figs.

  16. Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

    DOE PAGES

    Radulescu, Georgeta; Gauld, Ian C.; Ilas, Germina; Wagner, John C.

    2014-11-01

    This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application ofmore » the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.« less

  17. High-Resolution Compton-Suppressed CZT Detector for Fission Products Identification

    SciTech Connect

    R. Aryaeinejd; J. K. Hartwell; Wade W. Scates

    2004-10-01

    Room temperature semiconductor CdZnTe (CZT) detectors are currently limited to total detector volumes of 1-2 cm3, which is dictated by the poor charge transport characteristics. Because of this size limitation one of the problems in accurately determining isotope identification is the enormous background from the Compton scattering events. Eliminating this background will not only increase the sensitivity and accuracy of measurements but also help us to resolve peaks buried under the background and peaks in close vicinity of others. We are currently developing a fission products detection system based on the Compton-suppressed CZT detector. In this application, the detection system is required to operate in high radiation fields. Therefore, a small 10x10x5 mm3 CZT detector is placed inside the center of a well-shielded 3" in diameter by 3" long Nal detector. So far we have been able to successfully reduce the Compton background by a factor of 5.4 for a 137Cs spectrum. This reduction of background will definitely enhance the quality of the gamma-ray spectrum in the information-rich energy range below 1 MeV, which consequently increases the detection sensitivity. In this work, we will discuss the performance of this detection system as well as its applications.

  18. Flowsheet Testing of the Fission Product Extraction Process as Part of Advanced Aqueous Reprocessing

    SciTech Connect

    Jack Law; Dean R. Peterman; catherine Riddle; David H. Meikrantz; Terry Todd

    2007-06-01

    As part of the Advanced Fuel Cycle Initiative (AFCI), the reduction in volume and heat generation of spent nuclear fuel requiring geologic disposal is currently being addressed. The goal is to optimize utilization of the nation’s first repository and potentially reduce or eliminate the need for additional repositories. This will be achieved through separating long-lived, highly toxic elements, reducing high-level waste volumes and the toxicity of spent nuclear fuel, and reducing the heat generation of spent nuclear fuel. The Idaho National Laboratory (INL) is working closely with a team of national laboratories and other organizations to support development of these separations processes. Key to the reduction of short-term heat load in a geological repository is the separation of 137Cs and 90Sr. Removal of these highly radioactive fission products reduces the short-term (~100 yr) heat generation source of the spent nuclear fuel. Once separated, the Cs and Sr would be placed in storage until the activity has decayed to LLW levels, at which time it could potentially be disposed of as a non-transuranic (TRU) low-level waste (LLW).

  19. EXTRACTION METHOD FOR SEPARATING URANIUM, PLUTONIUM, AND FISSION PRODUCTS FROM COMPOSITIONS CONTAINING SAME

    DOEpatents

    Seaborg, G.T.

    1957-10-29

    Methods for separating plutonium from the fission products present in masses of neutron irradiated uranium are reported. The neutron irradiated uranium is first dissolved in an aqueous solution of nitric acid. The plutonium in this solution is present as plutonous nitrate. The aqueous solution is then agitated with an organic solvent, which is not miscible with water, such as diethyl ether. The ether extracts 90% of the uraryl nitrate leaving, substantially all of the plutonium in the aqueous phase. The aqueous solution of plutonous nitrate is then oxidized to the hexavalent state, and agitated with diethyl ether again. In the ether phase there is then obtained 90% of plutonium as a solution of plutonyl nitrate. The ether solution of plutonyl nitrate is then agitated with water containing a reducing agent such as sulfur dioxide, and the plutonium dissolves in the water and is reduced to the plutonous state. The uranyl nitrate remains in the ether. The plutonous nitrate in the water may be recovered by precipitation.

  20. Fission product iodine release and retention in nuclear reactor accidents— experimental programme at PSI

    NASA Astrophysics Data System (ADS)

    Bruchertseifer, H.; Cripps, R.; Guentay, S.; Jaeckel, B.

    2003-01-01

    Iodine radionuclides constitute one of the most important fission products of uranium and plutonium. If the volatile forms would be released into the environment during a severe accident, a potential health hazard would then ensue. Understanding its behaviour is an important prerequisite for planning appropriate mitigation measures. Improved and extensive knowledge of the main iodine species and their reactions important for the release and retention processes in the reactor containment is thus mandatory. The aim of PSI's radiolytical studies is to improve the current thermodynamic and kinetic databases and the models for iodine used in severe accident computer codes. Formation of sparingly soluble silver iodide (AgI) in a PWR containment sump can substantially reduce volatile iodine fraction in the containment atmosphere. However, the effectiveness is dependent on its radiation stability. The direct radiolytic decomposition of AgI and the effect of impurities on iodine volatilisation were experimentally determined at PSI using a remote-controlled and automated high activity 188W/Re generator (40 GBq/ml). Low molecular weight organic iodides are difficult to be retained in engineered safety systems. Investigation of radiolytic decomposition of methyl iodide in aqueous solutions, combined with an on-line analysis of iodine species is currently under investigation at PSI.

  1. Nondestructive assay of fission products in spent-fuel assemblies using gamma and photoneutron activation

    NASA Astrophysics Data System (ADS)

    Lakosi, L.; Veres, Á.

    1990-12-01

    Hard γ-radiation (above 1.078 MeV) from spent reactor fuel was detected by means of excitation of 115In to its 4.5 h half-life metastable state induced by the (γ, γ') reaction and subsequent counting of the 336 keV isomeric transition. Resonance-energy quanta were produced by Compton scattering in the source, i.e. the spent fuel itself. The sensitivity of the activation method above 1.67 MeV γ-energy was enhanced by introducing a Be photoneutron converter in order to produce neutrons for exploiting their much larger activation cross sections. For short cooling times (10-40 d) the hard-γ signature of the fuel was due to the fission product 140Ba 140La, detection of which facilitated monitoring of the reactor power which existed in the core just before reactor shutdown. A linear relationship was found between the γ-signal and the fissile content in the fuel. For 100-1000 d cooled fuel the 144Ce 144Pr content could be detected, which was only sensitive to the cooling time. Spent-fuel assemblies of both a research and a power reactor were assayed by these novel methods for reactor operational and nuclear-material safeguard purposes.

  2. Determination of Fission Product Yields of 235U, 238U and 239Pu for Neutron Energies from 0.5 to 14.8 MeV

    NASA Astrophysics Data System (ADS)

    Gooden, Matthew; Arnold, Charles; Becker, John; Bhatia, Chitra; Bhike, Megha; Fowler, Malcolm; Howell, Calvin; Kelley, John; Stoyer, Mark; Tonchev, Anton; Tornow, Werner; Vieira, Dave; Wilhelmy, Jerry

    2014-03-01

    A joint TUNL-LANL-LLNL collaboration has been formed to study the issue of possible energy dependences for certain fission product isotopes. Work has been carried out at the TUNL 10 MV Tandem accelerator which produces nearly mono-energetic neutrons via either 2H(d,n)3He,3H(d,n)4He,or3H(p,n)3He reactions. Three dual fission ionization chambers dedicated to 235U, 238U and 239Pu thick target foils and thin monitor foils respectively, were exposed to the neutron beams. After irradiation, thick target foils were gamma counted over a period of 1-2 months and characteristic gamma rays from fission products were recorded using HPGe detectors at TUNL's low background counting area. Using the dual fission chambers, relative fission product yield were determined at a high precision of 2-3 % as well as absolute fission product yields at a lower precision of 5-6 %. Preliminary results will be presented for a number of fission product isotopes over the incident neutron energy range of 0.5 to 14.8 MeV.

  3. Experimental investigations on the chemical state of solid fission-product elements in U3Si2

    NASA Astrophysics Data System (ADS)

    Ugajin, M.; Itoh, A.

    1994-10-01

    The uranium silicide U3Si2 has a congruent melting point of 1665 C and possesses higher uranium density (11.3 g U/cc) and higher thermal conductivity than the uranium dioxide currently used in light water reactors. U3Si2 is in use as a research reactor fuel (US Nuclear Regulatory Commission, NUREG-1313, July, 1988), representing a potentiality for power reactor fuel. A first attempt is made in this study to predict the chemical state of the solid fission-product elements comprising zirconium, molybdenum, rare earth elements, alkaline earth metals and elements of the platinum group. Ternary phase equilibria in the U-Mo-Si and U-Ru-Si systems are also investigated to supplement the fission product chemistry in U3Si2.

  4. Partition of actinides and fission products between metal and molten salt phases: Theory, measurement, and application to IFR pyroprocess development

    SciTech Connect

    Ackerman, J.P.; Johnson, T.R.

    1993-10-01

    The chemical basis of Integral Fast Reactor fuel reprocessing (pyroprocessing) is partition of fuel, cladding, and fission product elements between molten LiCl-KCl and either a solid metal phase or a liquid cadmium phase. The partition reactions are described herein, and the thermodynamic basis for predicting distributions of actinides and fission products in the pyroprocess is discussed. The critical role of metal-phase activity coefficients, especially those of rare earth and the transuranic elements, is described. Measured separation factors, which are analogous to equilibrium constants but which involve concentrations rather than activities, are presented. The uses of thermodynamic calculations in process development are described, as are computer codes developed for calculating material flows and phase compositions in pyroprocessing.

  5. Measurement of Airborne Fission Products in Chapel Hill, NC, USA from the Kukushima Dai-ichi Reactor Accident

    SciTech Connect

    MacMullin, S.; Giovanetti, G. K.; Green, M. P.; Henning, R.; Holmes, R.; Vorren, K.

    2012-01-01

    We present measurement results of airborne fission products in Chapel Hill, NC, USA, from 62 d following the March 11, 2011, accident at the Fukushima Dai-ichi nuclear power plant. Airborne particle samples were collected daily in air filters and radio-assayed with two high-purity germanium (HPGe) detectors. The fission products 131I and 137Cs were measured with maximum activity concentrations of 4.2 0.6 mBq/m3 and 0.42 0.07 mBq/m3 respectively. Additional activity from 131,132I, 134,136,137Cs and 132Te were measured in the same air filters using a low-background HPGe detector at the Kimballton Underground Research Facility (KURF).

  6. Aqueous Biphasic Systems Based on Salting-Out Polyethylene Glycol or Ionic Solutions: Strategies for Actinide or Fission Product Separations

    SciTech Connect

    Rogers, Robin D.; Gutowski, Keith E.; Griffin, Scott T.; Holbrey, John D.

    2004-03-29

    Aqueous biphasic systems can be formed by salting-out (with kosmotropic, waterstructuring salts) water soluble polymers (e.g., polyethylene glycol) or aqueous solutions of a wide range of hydrophilic ionic liquids based on imidazolium, pyridinium, phosphonium and ammonium cations. The use of these novel liquid/liquid biphases for separation of actinides or other fission products associated with nuclear wastes (e.g., pertechnetate salts) has been demonstrated and will be described in this presentation.

  7. Use of an ions thruster to dispose of type II long-lived fission products into outer space

    SciTech Connect

    Takahashi, H.; Yu, A.

    1997-04-01

    To dispose of long-lived fission products (LLFPs) into outer space, an ions thruster can be used instead of a static accelerator. The specifications of the ions thrusters which are presently studies for space propulsion are presented, and their usability discussed. Using of a rocket with an ions thruster for disposing of the LLFPs directly into the sun required a larger amount of energy than does the use of an accelerator.

  8. Deep Atomic Binding (DAB) Approach in Interpretation of Fission Products Behavior in Human Body, and Health Consequences

    SciTech Connect

    Ajlouni, Abdul-Wali M.S.

    2006-07-01

    According to models used to predict health effects of fission products enter the human body, a large number of fatalities, malignancies, thyroid cancer, born (genetic) defects,...etc.. But the actual data after Chernobyl and TMI accidents, and nuclear detonations in USA and Marshal Islands, were not consistent with these models. According to DAB, these data could be interpreted, and conflicts between former models predictions and actual field data explained. (author)

  9. Independent analysis of selected core-concrete interaction and fission product release experiments with CORCON-MOD2 and VANESA

    SciTech Connect

    Greene, G.A.; Sanborn, Y.

    1986-01-01

    The discrepancies between experimental findings and the Reactor Safety Study predictions, as well as the rapidly developing data base enabling phenomenological modeling of core-concrete interactions and ex-vessel fission product release, have prompted the development of several new computer models of core-concrete interactions and fission product release during severe accidents. Two such models are the CORCON-MOD2 model of core-concrete interactions and the VANESA model of ex-vessel aerosol and fission product release during core-concrete interactions. The final judge on the adequacy of the development of models of core debris-concrete interactions is, of course, comparison of the model predictions with the results of experiments. The research into ex-vessel core debris behavior differs from research into many aspects of reactor accidents in that there are many experimental results for comparison. Comparisons of code predictions with results of tests using realistic temperatures and conditions should provide an indication of the progress that has been made and, with appropriately chosen tests, an indication of work that needs to be done. 9 refs., 6 figs., 5 tabs.

  10. 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.

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

    DOE PAGES

    Knowles, Justin R.; Skutnik, Steven E.; Glasgow, David C.; Kapsimalis, Roger J.

    2016-06-23

    Rapid non-destructive assay methods for trace fissile material analysis are needed in both nuclear forensics and safeguards communities. To address these needs, research at the High Flux Isotope Reactor Neutron Activation Analysis laboratory has developed a generalized non-destructive assay method to characterize materials containing fissile isotopes. This method relies on gamma-ray emissions from short-lived fission products and capitalizes off 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 holistic characterization of isotopic identification,more » mass ratios, and absolute mass determination. Successful single fissile isotope mass recoveries of less than 6% error have been conducted on standards of 235U and 239Pu as low as 12 nanograms in less than 10 minutes. Additionally, mixtures of fissile isotope standards containing 235U and 239Pu have been characterized as low as 229 nanograms of fissile mass with less than 12% error. The generalizability of this method is illustrated by evaluating different fissile isotopes, mixtures of fissile isotopes, and two different irradiation positions in the reactor. Furthermore, it is anticipated that this method will be expanded to characterize additional fissile nuclides, utilize various irradiation sources, and account for increasingly complex sample matrices.« less

  12. Energetics of gaseous and volatile fission products in molten U-10Zr alloy: A density functional theory study

    NASA Astrophysics Data System (ADS)

    Wang, Ning; Tian, Jie; Jiang, Tao; Yang, Yanqiu; Hu, Sheng; Peng, Shuming; Yan, Liuming

    2015-11-01

    Gaseous and volatile fission products have a number of adverse effects on the safety and efficiency of the U-10Zr alloy fuel. The theoretical calculations were applied to investigate the energetics related to the formation, nucleation, and degassing of gaseous and volatile fission products (Kr, Xe and I) in molten U-10Zr alloy. The molecular dynamics (MD) simulations were applied to generate equilibrium configurations. The density functional theory (DFT) calculations were used to build atomistic models including molten U-10Zr alloy as well as its fission products substituted systems. The vacancy formation in liquid U-10Zr alloy were studied using DFT calculations, with average Gibbs free formation energies at 8.266 and 6.333 eV for U- and Zr-vacancies, respectively. And the interaction energies were -1.911 eV, -2.390 eV, and -1.826 eV for the I-I, Xe-Xe, and Kr-Kr interaction in lattice when two of the adjacent uranium atoms were substituted by gaseous atoms. So it could be concluded that the interaction between I, Kr, and Xe in lattice is powerful than the interaction between these two atoms and the other lattice atoms in U-10Zr.

  13. Development of fission-products transport model in severe-accident scenarios for Scdap/Relap5

    NASA Astrophysics Data System (ADS)

    Honaiser, Eduardo Henrique Rangel

    The understanding and estimation of the release of fission products during a severe accident became one of the priorities of the nuclear community after 1980, with the events of the Three-mile Island unit 2 (TMI-2), in 1979, and Chernobyl accidents, in 1986. Since this time, theoretical developments and experiments have shown that the primary circuit systems of light water reactors (LWR) have the potential to attenuate the release of fission products, a fact that had been neglected before. An advanced tool, compatible with nuclear thermal-hydraulics integral codes, is developed to predict the retention and physical evolution of the fission products in the primary circuit of LWRs, without considering the chemistry effects. The tool embodies the state-of-the-art models for the involved phenomena as well as develops new models. The capabilities acquired after the implementation of this tool in the Scdap/Relap5 code can be used to increase the accuracy of probability safety assessment (PSA) level 2, enhance the reactor accident management procedures and design new emergency safety features.

  14. Influences of Zr, Ce and Ba fission products on the surface properties of UO2: Atomistic simulations

    NASA Astrophysics Data System (ADS)

    Xiao, Hongxing; Long, Chongsheng; Tian, Xiaofeng; Chen, Hongsheng

    2016-07-01

    Molecular dynamics (MD) simulations with a shell-core model have been carried out to investigate the influences of Zr, Ce and Ba fission products on the surface properties of UO2. Simulation results indicate that (i) the presence of these fission products will change the surface energy of three low-index surfaces in UO2; (ii) the individual addition of Ce has no significant effect on the surface energy, while the individual addition of Ba will dramatically decrease the surface energy of UO2 by approximately 18% on (100) surface, 7% on (110) surface and 9% on (111) surface with the Ba contents ranging from 0 to 12.5 mol% at 300 K, which is obviously contrary to the Zr; (iii) the combined additions of Zr, Ce and Ba fission products will continuously increase the surface energy of UO2 (100), (110) and (111) surfaces; (iv) the structures of the three low-index surfaces in pure UO2 as well as U0.8(Zr, Ce, Ba)0.2O2 are dramatically disturbed after the free relaxation; (v) The nearest O atoms move towards the Zr and Ce atoms center by about 0.21 Å and 0.12 Å but move away from the Ba atom center by about 0.27 Å.

  15. Licensing topical report: the measurement and modelling of time-dependent fission product release from failed HTGR fuel particles under accident conditions

    SciTech Connect

    Myers, B.F.; Morrissey, R.E.

    1980-04-01

    The release of fission products from failed fuel particles was measured under simulated accident (core heatup) conditions. A generic model and specific model parameters that describe delayed fission product release from the kernels of failed HTGR fuel particles were developed from the experimental results. The release of fission products was measured from laser-failed BISO ThO/sub 2/ and highly enriched (HEU) TRISO UC/sub 2/ particles that had been irradiated to a range of kernel burnups. The burnups were 0.25, 1.4, and 15.7% FIMA for ThO/sub 2/ particles and 23.5 and 74% FIMA for UC/sub 2/ particles. The fission products measured were nuclides of xenon, iodine, krypton, tellurium, and cesium.

  16. Fission products silver, palladium, and cadmium identification in neutron-irradiated SiC TRISO particles using a Cs-Corrected HRTEM

    NASA Astrophysics Data System (ADS)

    van Rooyen, I. J.; Olivier, E. J.; Neethling, J. H.

    2016-08-01

    Electron microscopy investigations of selected coated particles from the first advanced gas reactor experiment at Idaho National Laboratory provided important information on fission product distribution and chemical composition in the silicon-carbide (SiC) layer. Silver precipitates were nano-sized, and therefore high-resolution transmission electron microscopy (HRTEM) was used to provide more information at the atomic level. Based on gamma-ray analysis, this particle which was irradiated to an average burnup of 19.38% fissions per initial metal atom, may have released as much as 10% of its available Ag-110 m inventory during irradiation. The HRTEM investigation focused on silver, palladium, and cadmium due to interest in silver transport mechanisms and possible correlation with palladium and silver previously found. Palladium, silver, and cadmium were found to co-exist in some of the SiC grain boundaries and triple junctions. This study confirmed palladium both at inter and intragranular sites. Phosphor was identified in SiC grain boundaries and triple points.

  17. NEUTRON CROSS SECTION EVALUATIONS OF FISSION PRODUCTS BELOW THE FAST ENERGY REGION

    SciTech Connect

    OH,S.Y.; CHANG,J.; MUGHABGHAB,S.

    2000-05-11

    Neutron cross section evaluations of the fission-product isotopes, {sup 95}Mo, {sup 99}Tc, {sup 101}Ru, {sup 103}Rh, {sup 105}Pd, {sup 109}Ag, {sup 131}Xe, {sup 133}Cs, {sup 141}Pr, {sup 141}Nd, {sup 147}Sm, {sup 149}Sm, {sup 150}Sm, {sup 151}Sm, {sup 152}Sm, {sup 153}Eu, {sup 155}Gd, and {sup 157}Gd were carried out below the fast neutron energy region within the framework of the BNL-KAERI international collaboration. In the thermal energy region, the energy dependence of the various cross-sections was calculated by applying the multi-level Breit-Wigner formalism. In particular, the strong energy dependence of the coherent scattering lengths of {sup 155}Gd and {sup 157}Gd were determined and were compared with recent calculations of Lynn and Seeger. In the resonance region, the recommended resonance parameters, reported in the BNL compilation, were updated by considering resonance parameter information published in the literature since 1981. The s-wave and, if available, p-wave reduced neutron widths were analyzed in terms of the Porter-Thomas distribution to determine the average level spacings and the neutron strength functions. Average radiative widths were also calculated from measured values of resolved energy resonances. The average resonance parameters determined in this study were compared with those in the BNL and other compilations, as well as the ENDF/B-VI, JEF-2.2, and JENDL-3.2 data libraries. The unresolved capture cross sections of these isotopes, computed with the determined average resonance parameters, were compared with measurements, as well as the ENDF/B-VI evaluations. To achieve agreement with the measurements, in a few cases minor adjustments in the average resonance parameters were made. Because of astrophysical interest, the Maxwellian capture cross sections of these nuclides at a neutron temperature of 30 keV were computed and were compared with other compilations and evaluations.

  18. Evaluation of Fission Product Critical Experiments and Associated Biases for Burnup Credit Validation

    SciTech Connect

    Mueller, Don; Rearden, Bradley T; Reed, Davis Allan

    2010-01-01

    One of the challenges associated with implementation of burnup credit is the validation of criticality calculations used in the safety evaluation; in particular the availability and use of applicable critical experiment data. The purpose of the validation is to quantify the relationship between reality and calculated results. Validation and determination of bias and bias uncertainty require the identification of sets of critical experiments that are similar to the criticality safety models. A principal challenge for crediting fission products (FP) in a burnup credit safety evaluation is the limited availability of relevant FP critical experiments for bias and bias uncertainty determination. This paper provides an evaluation of the available critical experiments that include FPs, along with bounding, burnup-dependent estimates of FP biases generated by combining energy dependent sensitivity data for a typical burnup credit application with the nuclear data uncertainty information distributed with SCALE 6. A method for determining separate bias and bias uncertainty values for individual FPs and illustrative results is presented. Finally, a FP bias calculation method based on data adjustment techniques and reactivity sensitivity coefficients calculated with the SCALE sensitivity/uncertainty tools and some typical results is presented. Using the methods described in this paper, the cross-section bias for a representative high-capacity spent fuel cask associated with the ENDF/B-VII nuclear data for 16 most important stable or near stable FPs is predicted to be no greater than 2% of the total worth of the 16 FPs, or less than 0.13 % k/k.

  19. Recoil-α-fission and recoil-α-α-fission events observed in the reaction 48Ca + 243Am

    NASA Astrophysics Data System (ADS)

    Forsberg, U.; Rudolph, D.; Andersson, L.-L.; Di Nitto, A.; Düllmann, Ch. E.; Fahlander, C.; Gates, J. M.; Golubev, P.; Gregorich, K. E.; Gross, C. J.; Herzberg, R.-D.; Heßberger, F. P.; Khuyagbaatar, J.; Kratz, J. V.; Rykaczewski, K.; Sarmiento, L. G.; Schädel, M.; Yakushev, A.; Åberg, S.; Ackermann, D.; Block, M.; Brand, H.; Carlsson, B. G.; Cox, D.; Derkx, X.; Dobaczewski, J.; Eberhardt, K.; Even, J.; Gerl, J.; Jäger, E.; Kindler, B.; Krier, J.; Kojouharov, I.; Kurz, N.; Lommel, B.; Mistry, A.; Mokry, C.; Nazarewicz, W.; Nitsche, H.; Omtvedt, J. P.; Papadakis, P.; Ragnarsson, I.; Runke, J.; Schaffner, H.; Schausten, B.; Shi, Yue; Thörle-Pospiech, P.; Torres, T.; Traut, T.; Trautmann, N.; Türler, A.; Ward, A.; Ward, D. E.; Wiehl, N.

    2016-09-01

    Products of the fusion-evaporation reaction 48Ca + 243Am were studied with the TASISpec set-up at the gas-filled separator TASCA at the GSI Helmholtzzentrum für Schwerionenforschung, Darmstadt, Germany. Amongst the detected thirty correlated α-decay chains associated with the production of element Z = 115, two recoil-α-fission and five recoil- α- α-fission events were observed. The latter five chains are similar to four such events reported from experiments performed at the Dubna gas-filled separator, and three such events reported from an experiment at the Berkeley gas-filled separator. The four chains observed at the Dubna gas-filled separator were assigned to start from the 2n-evaporation channel 289115 due to the fact that these recoil- α- α-fission events were observed only at low excitation energies. Contrary to this interpretation, we suggest that some of these recoil- α- α-fission decay chains, as well as some of the recoil- α- α-fission and recoil-α-fission decay chains reported from Berkeley and in this article, start from the 3n-evaporation channel 288115.

  20. Analyzing Losses: Transuranics into Waste and Fission Products into Recycled Fuel

    SciTech Connect

    Steven J. Piet; Nick R. Soelberg; Samuel E. Bays; Robert E. Cherry; Layne F. Pincock; Eric L. Shaber; Melissa C. Teague; Gregory M. Teske; Kurt G. Vedros; Candido Pereira; Denia Djokic

    2010-11-01

    All mass streams from separations and fuel fabrication are products that must meet criteria. Those headed for disposal must meet waste acceptance criteria (WAC) for the eventual disposal sites corresponding to their waste classification. Those headed for reuse must meet fuel or target impurity limits. A “loss” is any material that ends up where it is undesired. The various types of losses are linked in the sense that as the loss of transuranic (TRU) material into waste is reduced, often the loss or carryover of waste into TRU or uranium is increased. We have analyzed four separation options and two fuel fabrication options in a generic fuel cycle. The separation options are aqueous uranium extraction plus (UREX+1), electrochemical, Atomics International reduction oxidation separation (AIROX), and melt refining. UREX+1 and electrochemical are traditional, full separation techniques. AIROX and melt refining are taken as examples of limited separations, also known as minimum fuel treatment. The fuels are oxide and metal. To define a generic fuel cycle, a fuel recycling loop is fed from used light water reactor (LWR) uranium oxide fuel (UOX) at 51 MWth-day/kg-iHM burnup. The recycling loop uses a fast reactor with TRU conversion ratio (CR) of 0.50. Excess recovered uranium is put into storage. Only waste, not used fuel, is disposed – unless the impurities accumulate to a level so that it is impossible to make new fuel for the fast reactor. Impurities accumulate as dictated by separation removal and fission product generation. Our model approximates adjustment to fast reactor fuel stream blending of TRU and U products from incoming LWR UOX and recycling FR fuel to compensate for impurity accumulation by adjusting TRU:U ratios. Our mass flow model ignores postulated fuel impurity limits; we compare the calculated impurity values with those limits to identify elements of concern. AIROX and melt refining cannot be used to separate used LWR UOX-51 because they cannot

  1. 17 CFR 229.1204 - (Item 1204) Oil and gas production, production prices and production costs.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... noted. Instruction 2 to Item 1204: Production of natural gas should include only marketable production of natural gas on an “as sold” basis. Production will include dry, residue, and wet gas, depending on whether liquids have been extracted before the registrant transfers title. Flared gas, injected gas,...

  2. 17 CFR 229.1204 - (Item 1204) Oil and gas production, production prices and production costs.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... noted. Instruction 2 to Item 1204: Production of natural gas should include only marketable production of natural gas on an “as sold” basis. Production will include dry, residue, and wet gas, depending on whether liquids have been extracted before the registrant transfers title. Flared gas, injected gas,...

  3. Fuel efficient hydrodynamic containment for gas core fission reactor rocket propulsion. Final report, September 30, 1992--May 31, 1995

    SciTech Connect

    Sforza, P.M.; Cresci, R.J.

    1997-05-31

    Gas core reactors can form the basis for advanced nuclear thermal propulsion (NTP) systems capable of providing specific impulse levels of more than 2,000 sec., but containment of the hot uranium plasma is a major problem. The initial phase of an experimental study of hydrodynamic confinement of the fuel cloud in a gas core fission reactor by means of an innovative application of a base injection stabilized recirculation bubble is presented. The development of the experimental facility, a simulated thrust chamber approximately 0.4 m in diameter and 1 m long, is described. The flow rate of propellant simulant (air) can be varied up to about 2 kg/sec and that of fuel simulant (air, air-sulfur hexafluoride) up to about 0.2 kg/sec. This scale leads to chamber Reynolds numbers on the same order of magnitude as those anticipated in a full-scale nuclear rocket engine. The experimental program introduced here is focused on determining the size, geometry, and stability of the recirculation region as a function of the bleed ratio, i.e. the ratio of the injected mass flux to the free stream mass flux. A concurrent CFD study is being carried out to aid in demonstrating that the proposed technique is practical.

  4. Actinide, Activation Product and Fission Product Decay Data for Reactor-based Applications

    SciTech Connect

    Perry, R.J.; Dean, C.J.; Nichols, A.L.

    2014-06-15

    The UK Activation Product Decay Data Library was first released in September 1977 as UK-PADD1, to be followed by regular improvements on an almost yearly basis up to the assembly of UKPADD6.12 in March 2013. Similarly, the UK Heavy Element and Actinide Decay Data Library followed in December 1981 as UKHEDD1, with the implementation of various modifications leading to UKHEDD2.6, February 2008. Both the data content and evaluation procedures are defined, and the most recent evaluations are described in terms of specific radionuclides and the resulting consistency of their recommended decay-data files. New versions of the UKPADD and UKHEDD libraries are regularly submitted to the NEA Data Bank for possible inclusion in the JEFF library.

  5. Characteristics of gas dynamics of flow lasers excited by fission fragments of uranium nuclei

    SciTech Connect

    Borovkov, V.V.; Lazhintsev, B.V.; Sizov, A.N.

    1995-12-01

    The conceptual design of a nuclear-pumped cw laser is put forward. Alternation of laser cells with plane uranium layers and heat exchangers (radiators) in a shared gas loop can reduce the gas velocity down to {approximately} 10 m s{sup {minus}1}. The results are reported of an investigation of optical inhomogeneities which appear in He and Ar due to excitation of the active medium in a prototype flow laser. It is shown that, in a section perpendicular to the plane of the uranium layers, a pumping inhomogeneity creates a positive parabolic gas lens and in a section parallel to these layers an optical gas wedge is formed. A vortex zone is observed in the gas flow at the exits from heat exchangers. Simulation experiments demonstrate that this effect increases tens of times the thermal diffusivity of the gas and results in considerable refractive losses of radiation in the effective heat-exchange region. Methods of compensating for optical inhomogeneities and for reducing the influence of vortices are proposed. 17 refs., 5 figs.

  6. Microstructural Characterization of Irradiated U-7Mo/Al-5Si Dispersion to High Fission Density

    SciTech Connect

    J. Gan; B. D. Miller; D. D. Keiser, Jr.; A. B. Robinson; J. W. Madden; P. G. Medvedev; D. M. Wachs

    2014-11-01

    The fuel development program for research and test reactors calls for improved knowledge on the effect of microstructure on fuel performance in reactors. This work summarizes the recent TEM microstructural characterization of an irradiated U-7Mo/Al-5Si dispersion fuel plate (R3R050) irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory to 5.2×1021 fissions/cm3. While a large fraction of the fuel grains is decorated with large bubbles, there is no evidence showing interlinking of these large bubbles at the specified fission density. The attachment of solid fission product precipitates to the bubbles is likely the result of fission product diffusion into these bubbles. The process of fission gas bubble superlattice collapse appears through bubble coalescence. The results are compared with the previous TEM work of the dispersion fuels irradiated to lower fission density from the same fuel plate.

  7. Fission Product Yields for 14 MeV Neutrons on 235U, 238U and 239Pu

    NASA Astrophysics Data System (ADS)

    Mac Innes, M.; Chadwick, M. B.; Kawano, T.

    2011-12-01

    We report cumulative fission product yields (FPY) measured at Los Alamos for 14 MeV neutrons on 235U, 238U and 239Pu. The results are from historical measurements made in the 1950s-1970s, not previously available in the peer reviewed literature, although an early version of the data was reported in the Ford and Norris review. The results are compared with other measurements and with the ENDF/B-VI England and Rider evaluation. Compared to the Laurec (CEA) data and to ENDF/B-VI evaluation, good agreement is seen for 235U and 238U, but our FPYs are generally higher for 239Pu. The reason for the higher plutonium FPYs compared to earlier Los Alamos assessments reported by Ford and Norris is that we update the measured values to use modern nuclear data, and in particular the 14 MeV 239Pu fission cross section is now known to be 15-20% lower than the value assumed in the 1950s, and therefore our assessed number of fissions in the plutonium sample is correspondingly lower. Our results are in excellent agreement with absolute FPY measurements by Nethaway (1971), although Nethaway later renormalized his data down by 9% having hypothesized that he had a normalization error. The new ENDF/B-VII.1 14 MeV FPY evaluation is in good agreement with our data.

  8. How EIA Estimates Natural Gas Production

    EIA Publications

    2004-01-01

    The Energy Information Administration (EIA) publishes estimates monthly and annually of the production of natural gas in the United States. The estimates are based on data EIA collects from gas producing states and data collected by the U. S. Minerals Management Service (MMS) in the Department of Interior. The states and MMS collect this information from producers of natural gas for various reasons, most often for revenue purposes. Because the information is not sufficiently complete or timely for inclusion in EIA's Natural Gas Monthly (NGM), EIA has developed estimation methodologies to generate monthly production estimates that are described in this document.

  9. Effects of radiation and fission product incorporation in a yttria-stabilized zirconia based inert matrix fuel

    NASA Astrophysics Data System (ADS)

    Zhu, Sha

    This work has investigated the irradiation and incorporation effects of fission products in a yttria-stabilized zirconia (YSZ) based inert matrix fuel (IMF). The concept of inert matrix fuel is based on a new strategy for disposition of plutonium generated from the reprocessing of commercial nuclear fuel and the dismantling of nuclear weapons, i.e. using uranium-free oxides to "burn" plutonium and other actinides (Np, Cm, and Am) in reactors. This approach allows direct disposal, without reprocessing, after once-through burn-up. YSZ and MgAl2O4-YSZ composites are among the potential ceramics for IMF due to their high chemical durability and radiation resistance. The research involved investigating the production, nature, and accumulation of irradiation-induced defects, the behavior of the fission products in the ceramics, the structural stability and amorphization resistance of the YSZ during implantation. Ion implantations were conducted with 200--400 keV Cs+, Sr+, I+, Xe+ and Ti+ up to fluences of 1 x 1017/cm 2 at both room temperature and temperatures of 600--700°C. Thermal annealing was subsequently completed after room temperature ion implantations. In situ and ex situ transmission electron microscopy (TEM), optical absorption spectroscopy, photo-luminescence spectroscopy, and electron paramagnetic resonance (EPR) spectroscopy were employed to characterize the irradiation induced defect evolution and analyze the defect structures. Various irradiation effects were observed and determined in the experiments, such as point defects (F type and V type color centers), defect clusters (dislocation loops), cavities (voids and bubbles), the crystalline-to-amorphous transition, and the phase transformation from fluorite to pyrochlore structure. The ion irradiation-induced amorphization mechanism, the retention ability of the fission products, and structural stability of YSZ are discussed in terms of ion incorporation effects, implanted ion radii, and the solubility

  10. Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

    NASA Astrophysics Data System (ADS)

    Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

    2014-09-01

    One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo99 used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 106 cm-1) in a tube, their delta reactivities are the still

  11. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses: Criticality (keff) Predictions

    DOE PAGES

    Scaglione, John M.; Mueller, Don E.; Wagner, John C.

    2014-12-01

    One of the most important remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation—in particular, the availability and use of applicable measured data to support validation, especially for fission products (FPs). Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. In this study, this paper describes a validation approach for commercial spent nuclear fuel (SNF) criticality safety (keff) evaluations based on best-available data and methodsmore » and applies the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The criticality validation approach utilizes not only available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion program to support validation of the principal actinides but also calculated sensitivities, nuclear data uncertainties, and limited available FP LCE data to predict and verify individual biases for relevant minor actinides and FPs. The results demonstrate that (a) sufficient critical experiment data exist to adequately validate keff calculations via conventional validation approaches for the primary actinides, (b) sensitivity-based critical experiment selection is more appropriate for generating accurate application model bias and uncertainty, and (c) calculated sensitivities and nuclear data uncertainties can be used for generating conservative estimates of bias for minor actinides and FPs. Results based on the SCALE 6.1 and the ENDF/B-VII.0 cross-section libraries indicate that a conservative estimate of the bias for the minor actinides and FPs is 1.5% of their worth within the

  12. Atomistic modeling of intrinsic and radiation-enhanced fission gas (Xe) diffusion in UO2±x: Implications for nuclear fuel performance modeling

    NASA Astrophysics Data System (ADS)

    Andersson, D. A.; Garcia, P.; Liu, X.-Y.; Pastore, G.; Tonks, M.; Millett, P.; Dorado, B.; Gaston, D. R.; Andrs, D.; Williamson, R. L.; Martineau, R. C.; Uberuaga, B. P.; Stanek, C. R.

    2014-08-01

    Based on density functional theory (DFT) and empirical potential calculations, the diffusivity of fission gas atoms (Xe) in UO2 nuclear fuel has been calculated for a range of non-stoichiometry (i.e. UO2±x), under both out-of-pile (no irradiation) and in-pile (irradiation) conditions. This was achieved by first deriving expressions for the activation energy that account for the type of trap site that the fission gas atoms occupy, which includes the corresponding type of mobile cluster, the charge state of these defects and the chemistry acting as boundary condition. In the next step DFT calculations were used to estimate migration barriers and internal energy contributions to the thermodynamic properties and calculations based on empirical potentials were used to estimate defect formation and migration entropies (i.e. pre-exponentials). The diffusivities calculated for out-of-pile conditions as function of the UO2±x non-stoichiometry were used to validate the accuracy of the diffusion models and the DFT calculations against available experimental data. The Xe diffusivity is predicted to depend strongly on the UO2±x non-stoichiometry due to a combination of changes in the preferred Xe trap site and in the concentration of uranium vacancies enabling Xe diffusion, which is consistent with experiments. After establishing the validity of the modeling approach, it was used for studying Xe diffusion under in-pile conditions, for which experimental data is very scarce. The radiation-enhanced Xe diffusivity is compared to existing empirical models. Finally, the predicted fission gas diffusion rates were implemented in the BISON fuel performance code and fission gas release from a Risø fuel rod irradiation experiment was simulated.

  13. Atomistic modeling of intrinsic and radiation-enhanced fission gas (Xe) diffusion in UO2 +/- x: Implications for nuclear fuel performance modeling

    SciTech Connect

    Giovanni Pastore; Michael R. Tonks; Derek R. Gaston; Richard L. Williamson; David Andrs; Richard Martineau

    2014-03-01

    Based on density functional theory (DFT) and empirical potential calculations, the diffusivity of fission gas atoms (Xe) in UO2 nuclear fuel has been calculated for a range of non-stoichiometry (i.e. UO2x), under both out-of-pile (no irradiation) and in-pile (irradiation) conditions. This was achieved by first deriving expressions for the activation energy that account for the type of trap site that the fission gas atoms occupy, which includes the corresponding type of mobile cluster, the charge state of these defects and the chemistry acting as boundary condition. In the next step DFT calculations were used to estimate migration barriers and internal energy contributions to the thermodynamic properties and calculations based on empirical potentials were used to estimate defect formation and migration entropies (i.e. pre-exponentials). The diffusivities calculated for out-of-pile conditions as function of the UO2x nonstoichiometrywere used to validate the accuracy of the diffusion models and the DFT calculations against available experimental data. The Xe diffusivity is predicted to depend strongly on the UO2x non-stoichiometry due to a combination of changes in the preferred Xe trap site and in the concentration of uranium vacancies enabling Xe diffusion, which is consistent with experiments. After establishing the validity of the modeling approach, it was used for studying Xe diffusion under in-pile conditions, for which experimental data is very scarce. The radiation-enhanced Xe diffusivity is compared to existing empirical models. Finally, the predicted fission gas diffusion rates were implemented in the BISON fuel performance code and fission gas release from a Risø fuel rod irradiation experiment was simulated. 2014 Elsevier B.V. All rights

  14. FEMAXI-V benchmarking study on peak temperature and fission gas release prediction of PWR rod fuel

    SciTech Connect

    Suwardi; Dewayatna, W.; Briyatmoko, B.

    2012-06-06

    The present paper reports a study of FEMAXI-V code and related report on code benchmarking. Capabilities of the FEMAXI-V code to predict the thermal and fission gas release have been tested on MOX fuels in LWRs which has been done in SCK{center_dot}CEN and Belgonucleaire by using PRIMO MOX rod BD8 irradiation experiment after V Sobolev as reported O. J. Ott. Base irradiation in the BR3 reactor, the BD8 rod was transported to CEA-Saclay for irradiation in the OSIRIS reactor (ramp power excursion). The irradiation device used for the PRIMO ramps was the ISABELLE 1 loop, installed on a movable structure of the core periphery. The power variations were obtained by inwards/backwards movements of the loop in the core water. The preconditioning phase for rod BD8 occurred at a peak power level of 189 W/cm with a hold time of 27 hours. The subsequent power excursion rate amounted to 77 W/ (cm.min), reaching a terminal peak power level of 395 W/cm that lasted for 20 hours.

  15. ConocoPhillips Gas Hydrate Production Test

    SciTech Connect

    Schoderbek, David; Farrell, Helen; Howard, James; Raterman, Kevin; Silpngarmlert, Suntichai; Martin, Kenneth; Smith, Bruce; Klein, Perry

    2013-06-30

    Work began on the ConocoPhillips Gas Hydrates Production Test (DOE award number DE-NT0006553) on October 1, 2008. This final report summarizes the entire project from January 1, 2011 to June 30, 2013.

  16. Glass Ceramic Waste Forms for Combined CS+LN+TM Fission Products Waste Streams

    SciTech Connect

    Crum, Jarrod V.; Turo, Laura A.; Riley, Brian J.; Tang, Ming; Kossoy, Anna; Sickafus, Kurt E.

    2010-09-23

    In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.[1] The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development of a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both combined waste

  17. Use of Information Theory Concepts for Developing Contaminated Site Detection Method: Case for Fission Product and Actinides Accumulation Modeling

    SciTech Connect

    Harbachova, N.V.; Sharavarau, H.A.

    2006-07-01

    Information theory concepts and their fundamental importance for environmental pollution analysis in light of experience of Chernobyl accident in Belarus are discussed. An information and dynamic models of the radionuclide composition formation in the fuel of the Nuclear Power Plant are developed. With the use of code DECA numerical calculation of actinides (58 isotopes are included) and fission products (650 isotopes are included) activities has been carried out and their dependence with the fuel burn-up of the RBMK-type reactor have been investigated. (authors)

  18. Cement As a Waste Form for Nuclear Fission Products: The Case of (90)Sr and Its Daughters.

    PubMed

    Dezerald, Lucile; Kohanoff, Jorge J; Correa, Alfredo A; Caro, Alfredo; Pellenq, Roland J-M; Ulm, Franz J; Saúl, Andrés

    2015-11-17

    One of the main challenges faced by the nuclear industry is the long-term confinement of nuclear waste. Because it is inexpensive and easy to manufacture, cement is the material of choice to store large volumes of radioactive materials, in particular the low-level medium-lived fission products. It is therefore of utmost importance to assess the chemical and structural stability of cement containing radioactive species. Here, we use ab initio calculations based on density functional theory (DFT) to study the effects of (90)Sr insertion and decay in C-S-H (calcium-silicate-hydrate) in order to test the ability of cement to trap and hold this radioactive fission product and to investigate the consequences of its β-decay on the cement paste structure. We show that (90)Sr is stable when it substitutes the Ca(2+) ions in C-S-H, and so is its daughter nucleus (90)Y after β-decay. Interestingly, (90)Zr, daughter of (90)Y and final product in the decay sequence, is found to be unstable compared to the bulk phase of the element at zero K but stable when compared to the solvated ion in water. Therefore, cement appears as a suitable waste form for (90)Sr storage.

  19. Cement As a Waste Form for Nuclear Fission Products: The Case of (90)Sr and Its Daughters.

    PubMed

    Dezerald, Lucile; Kohanoff, Jorge J; Correa, Alfredo A; Caro, Alfredo; Pellenq, Roland J-M; Ulm, Franz J; Saúl, Andrés

    2015-11-17

    One of the main challenges faced by the nuclear industry is the long-term confinement of nuclear waste. Because it is inexpensive and easy to manufacture, cement is the material of choice to store large volumes of radioactive materials, in particular the low-level medium-lived fission products. It is therefore of utmost importance to assess the chemical and structural stability of cement containing radioactive species. Here, we use ab initio calculations based on density functional theory (DFT) to study the effects of (90)Sr insertion and decay in C-S-H (calcium-silicate-hydrate) in order to test the ability of cement to trap and hold this radioactive fission product and to investigate the consequences of its β-decay on the cement paste structure. We show that (90)Sr is stable when it substitutes the Ca(2+) ions in C-S-H, and so is its daughter nucleus (90)Y after β-decay. Interestingly, (90)Zr, daughter of (90)Y and final product in the decay sequence, is found to be unstable compared to the bulk phase of the element at zero K but stable when compared to the solvated ion in water. Therefore, cement appears as a suitable waste form for (90)Sr storage. PMID:26513644

  20. RADIOLYTIC GAS PRODUCTION RATES OF POLYMERS EXPOSED TO TRITIUM GAS

    SciTech Connect

    Clark, E.

    2013-08-31

    Data from previous reports on studies of polymers exposed to tritium gas is further analyzed to estimate rates of radiolytic gas production. Also, graphs of gas release during tritium exposure from ultrahigh molecular weight polyethylene (UHMW-PE), polytetrafluoroethylene (PTFE, a trade name is Teflon®), and Vespel® polyimide are re-plotted as moles of gas as a function of time, which is consistent with a later study of tritium effects on various formulations of the elastomer ethylene-propylene-diene monomer (EPDM). These gas production rate estimates may be useful while considering using these polymers in tritium processing systems. These rates are valid at least for the longest exposure times for each material, two years for UHMW-PE, PTFE, and Vespel®, and fourteen months for filled and unfilled EPDM. Note that the production “rate” for Vespel® is a quantity of H{sub 2} produced during a single exposure to tritium, independent of length of time. The larger production rate per unit mass for unfilled EPDM results from the lack of filler- the carbon black in filled EPDM does not produce H{sub 2} or HT. This is one aspect of how inert fillers reduce the effects of ionizing radiation on polymers.

  1. Summary Report: Glass-Ceramic Waste Forms for Combined Fission Products

    SciTech Connect

    Crum, Jarrod V.; Riley, Brian J.; Turo, Laura A.; Tang, Ming; Kossoy, Anna

    2011-09-23

    Glass-ceramic waste form development began in FY 2010 examining two combined waste stream options: (1) alkaline earth (CS) + lanthanide (Ln), and (2) + transition metal (TM) fission-product waste streams generated by the uranium extraction (UREX+) separations process. Glass-ceramics were successfully developed for both options however; Option 2 was selected over Option 1, at the conclusion of 2010, because Option 2 immobilized all three waste streams with only a minimal decrease in waste loading. During the first year, a series of three glass (Option 2) were fabricated that varied waste loading-WL (42, 45, and 50 mass%) at fixed molar ratios of CaO/MoO{sub 3} and B{sub 2}O{sub 3}/alkali both at 1.75. These glass-ceramics were slow cooled and characterized in terms of phase assemblage and preliminary irradiation stability. This fiscal year, further characterization was performed on the FY 2010 Option 2 glass-ceramics in terms of: static leach testing, phase analysis by transmission electron microscopy (TEM), and irradiation stability (electron and ion). Also, a new series of glass-ceramics were developed for Option 2 that varied the additives: Al{sub 2}O{sub 3} (0-6 mass%), molar ratio of CaO/MoO{sub 3} and B{sub 2}O{sub 3}/alkali (1.75 to 2.25) and waste loading (50, 55, and 60 mass%). Lastly, phase pure powellite and oxyapatite were synthesized for irradiation studies. Results of this fiscal year studies showed compositional flexibility, chemical stability, and radiation stability in the current glass-ceramic system. First, the phase assemblages and microstructure of all of the FY 2010 and 2011 glass-ceramics are very similar once subjected to the slow cool heat treatment. The phases identified in these glass-ceramics were oxyapatite, powellite, cerianite, and ln-borosilicate. This shows that variations in waste loading or additives can be accommodated without drastically changing the phase assemblage of the waste form, thus making the processing and performance

  2. Comparison of various hours living fission products for absolute power density determination in VVER-1000 mock up in LR-0 reactor.

    PubMed

    Košťál, Michal; Švadlenková, Marie; Koleška, Michal; Rypar, Vojtěch; Milčák, Ján

    2015-11-01

    Measuring power level of zero power reactor is a quite difficult task. Due to the absence of measurable cooling media heating, it is necessary to employ a different method. The gamma-ray spectroscopy of fission products induced within reactor operation is one of possible ways of power determination. The method is based on the proportionality between fission product buildup and released power. The (92)Sr fission product was previously preferred as nuclide for LR-0 power determination for short-time irradiation experiments. This work aims to find more appropriate candidates, because the (92)Sr, however suitable, has a short half-life, which limits the maximal measurable amount of fuel pins within a single irradiation batch. The comparison of various isotopes is realized for (92)Sr, (97)Zr, (135)I, (91)Sr, and (88)Kr. The comparison between calculated and experimentally determined (C/E-1 values) net peak areas is assessed for these fission products. Experimental results show that studied fission products, except (88)Kr, are in comparable agreement with (92)Sr results. Since (91)Sr has notably higher half-life than (92)Sr, (91)Sr seems to be more appropriate marker in experiments with a large number of measured fuel pins.

  3. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

    SciTech Connect

    Chadwick, M.B.; Herman, M.; Author : Chadwick,M.B.; Herman,M.; Oblozinsky,P.; Dunn,M.E.; Danon,Y.; Kahler,A.C.; Smith,D.L.; Pritychenko,B.; Arbanas,G.; Arcilla,R.; Brewer,R.; Brown,D.A.; Capote,R.; Carlson,A.D.; Cho,Y.S.; Derrien,H.; Guber,K.; Hale,G.M.; Hoblit,S.; Holloway,S.: Johnson,T.D.; Kawano,T.; Kiedrowski,B.C.; Kim,H.; Kunieda,S.; Larson,N.M.; Leal,L.; Lestone,J.P.; Little,R.C.; McCutchan,E.A.; MacFarlane,R.E.; MacInnes,M.; Mattoon,C.M.; McKnight,R.D.; Mughabghab,S.F.; Nobre,G.P.A.; Palmiotti,G.; Palumbo,A.; Pigni,M.T.; Pronyaev,V.G.; Sayer,R.O.; Sonzogni,A.A.; Summers,N.C.; Talou,P.; Thompson,I.J.; Trkov,A.; Vogt,R.L.; van der Marck,S.C.; Wallner,A.; White,M.C.; Wiarda,D.; Young,P.G.

    2011-12-01

    The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides {sup 235,238}U and {sup 239}Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on {sup 239}Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0

  4. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

    NASA Astrophysics Data System (ADS)

    Chadwick, M. B.; Herman, M.; Obložinský, P.; Dunn, M. E.; Danon, Y.; Kahler, A. C.; Smith, D. L.; Pritychenko, B.; Arbanas, G.; Arcilla, R.; Brewer, R.; Brown, D. A.; Capote, R.; Carlson, A. D.; Cho, Y. S.; Derrien, H.; Guber, K.; Hale, G. M.; Hoblit, S.; Holloway, S.; Johnson, T. D.; Kawano, T.; Kiedrowski, B. C.; Kim, H.; Kunieda, S.; Larson, N. M.; Leal, L.; Lestone, J. P.; Little, R. C.; McCutchan, E. A.; MacFarlane, R. E.; MacInnes, M.; Mattoon, C. M.; McKnight, R. D.; Mughabghab, S. F.; Nobre, G. P. A.; Palmiotti, G.; Palumbo, A.; Pigni, M. T.; Pronyaev, V. G.; Sayer, R. O.; Sonzogni, A. A.; Summers, N. C.; Talou, P.; Thompson, I. J.; Trkov, A.; Vogt, R. L.; van der Marck, S. C.; Wallner, A.; White, M. C.; Wiarda, D.; Young, P. G.

    2011-12-01

    The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides 235,238U and 239Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range

  5. Fission fragment driven neutron source

    DOEpatents

    Miller, Lowell G.; Young, Robert C.; Brugger, Robert M.

    1976-01-01

    Fissionable uranium formed into a foil is bombarded with thermal neutrons in the presence of deuterium-tritium gas. The resulting fission fragments impart energy to accelerate deuterium and tritium particles which in turn provide approximately 14 MeV neutrons by the reactions t(d,n).sup.4 He and d(t,n).sup.4 He.

  6. Evolution of gas saturation and relative permeability during gas production from hydrate-bearing sediments: Gas invasion vs. gas nucleation

    NASA Astrophysics Data System (ADS)

    Jang, Jaewon; Santamarina, J. Carlos

    2014-01-01

    Capillarity and both gas and water permeabilities change as a function of gas saturation. Typical trends established in the discipline of unsaturated soil behavior are used when simulating gas production from hydrate-bearing sediments. However, the evolution of gas saturation and water drainage in gas invasion (i.e., classical soil behavior) and gas nucleation (i.e., gas production) is inherently different: micromodel experimental results show that gas invasion forms a continuous flow path while gas nucleation forms isolated gas clusters. Complementary simulations conducted using tube networks explore the implications of the two different desaturation processes. In spite of their distinct morphological differences in fluid displacement, numerical results show that the computed capillarity-saturation curves are very similar in gas invasion and nucleation (the gas-water interface confronts similar pore throat size distribution in both cases); the relative water permeability trends are similar (the mean free path for water flow is not affected by the topology of the gas phase); and the relative gas permeability is slightly lower in nucleation (delayed percolation of initially isolated gas-filled pores that do not contribute to gas conductivity). Models developed for unsaturated sediments can be used for reservoir simulation in the context of gas production from hydrate-bearing sediments, with minor adjustments to accommodate a lower gas invasion pressure Po and a higher gas percolation threshold.

  7. Assessment of Fission Product Cross-Section Data for Burnup Credit Applications

    SciTech Connect

    Leal, Luiz C; Derrien, Herve; Dunn, Michael E; Mueller, Don

    2007-12-01

    Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance. Moreover, DOE, NRC, and EPRI have noted the need for additional scientific and technical data to justify expanding PWR burnup credit to include fission product (FP) nuclides and enable burnup credit implementation for boiling-water reactor (BWR) spent nuclear fuel (SNF). The criticality safety assessment needed for burnup credit applications will utilize computational analyses of packages containing SNF with FP nuclides. Over the years, significant efforts have been devoted to the nuclear data evaluation of major isotopes pertinent to reactor applications (i.e., uranium, plutonium, etc.); however, efforts to evaluate FP cross-section data in the resonance region have been less thorough relative to actinide data. In particular, resonance region cross-section measurements with corresponding R-matrix resonance analyses have not been performed for FP nuclides. Therefore, the objective of this work is to assess the status and performance of existing FP cross-section and cross-section uncertainty data in the resonance region for use in burnup credit analyses. Recommendations for new cross-section measurements and/or evaluations are made based on the data assessment. The assessment focuses on seven primary FP isotopes (103Rh, 133Cs, 143Nd, 149Sm, 151Sm, 152Sm, and 155Gd) that impact reactivity analyses of transportation packages and two FP isotopes (153Eu and 155Eu) that impact prediction of 155Gd concentrations. Much of the assessment work was completed in 2005, and the assessment focused on the latest FP cross-section evaluations available in the

  8. Stochastic simulation of fission product activity in primary coolant due to fuel rod failures in typical PWRs under power transients

    NASA Astrophysics Data System (ADS)

    Javed Iqbal, M.; Mirza, Nasir M.; Mirza, Sikander M.

    2008-01-01

    During normal operation of PWRs, routine fuel rods failures result in release of radioactive fission products (RFPs) in the primary coolant of PWRs. In this work, a stochastic model has been developed for simulation of failure time sequences and release rates for the estimation of fission product activity in primary coolant of a typical PWR under power perturbations. In the first part, a stochastic approach is developed, based on generation of fuel failure event sequences by sampling the time dependent intensity functions. Then a three-stage model based deterministic methodology of the FPCART code has been extended to include failure sequences and random release rates in a computer code FPCART-ST, which uses state-of-the-art LEOPARD and ODMUG codes as its subroutines. The value of the 131I activity in primary coolant predicted by FPCART-ST code has been found in good agreement with the corresponding values measured at ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with constant release option have also been found to have good agreement with corresponding experimental values for time dependent 135I, 135Xe and 89Kr concentrations in primary coolant measured during EDITHMOX-1 experiments.

  9. Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors

    SciTech Connect

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel [1]. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as PuFP. The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the PuFP composition. The FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  10. Measurement of fission products β decay properties using a total absorption spectrometer

    NASA Astrophysics Data System (ADS)

    Zakari-Issoufou, A.-A.; Porta, A.; Fallot, M.; Algora, A.; Tain, J. L.; Valencia, E.; Rice, S.; Agramunt, J.; Äystö, J.; Bowry, M.; Bui, V. M.; Caballero-Folch, R.; Cano-Ott, D.; Eloma, V.; Estévez, E.; Farrelly, G. F.; Garcia, A.; Gelletly, W.; Gomez-Hornillos, M. B.; Gorlychev, V.; Hakala, J.; Jokinen, A.; Jordan, M. D.; Kankainen, A.; Kondev, F. G.; Martinez, T.; Mendoza, E.; Molina, F.; Moore, I.; Perez, A.; Podolyak, Zs.; Penttilä, H.; Regan, P. H.; Rissanen, J.; Rubio, B.; Weber, C.

    2013-12-01

    In a nuclear reactor, the β decay of fission fragments is at the origin of decay heat and antineutrino flux. These quantities are not well known while they are very important for reactor safety and for our understanding of neutrino physics. One reason for the discrepancies observed in the estimation of the decay heat and antineutrinos flux coming from reactors could be linked with the Pandemonium effect. New measurements have been performed at the JYFL facility of Jyväskylä with a Total Absorption Spectrometer (TAS) in order to circumvent this effect. An overview of the TAS technique and first results from the 2009 measurement campaign will be presented.

  11. Spontaneous Fission

    DOE R&D Accomplishments Database

    Segre, Emilio

    1950-11-22

    The first attempt to discover spontaneous fission in uranium was made by [Willard] Libby, who, however, failed to detect it on account of the smallness of effect. In 1940, [K. A.] Petrzhak and [G. N.] Flerov, using more sensitive methods, discovered spontaneous fission in uranium and gave some rough estimates of the spontaneous fission decay constant of this substance. Subsequently, extensive experimental work on the subject has been performed by several investigators and will be quoted in the various sections. [N.] Bohr and [A.] Wheeler have given a theory of the effect based on the usual ideas of penetration of potential barriers. On this project spontaneous fission has been studied for the past several years in an effort to obtain a complete picture of the phenomenon. For this purpose the spontaneous fission decay constants {lambda} have been measured for separated isotopes of the heavy elements wherever possible. Moreover, the number {nu} of neutrons emitted per fission has been measured wherever feasible, and other characteristics of the spontaneous fission process have been studied. This report summarizes the spontaneous fission work done at Los Alamos up to January 1, 1945. A chronological record of the work is contained in the Los Alamos monthly reports.

  12. Molecular dynamics study of fission gas bubble nucleation in UO2

    NASA Astrophysics Data System (ADS)

    Liu, X.-Y.; Andersson, D. A.

    2015-07-01

    Molecular dynamics (MD) simulations are used to study helium and xenon gas bubble nucleation in UO2. For helium bubbles, the pressure release mechanism is by creating defects on the oxygen sublattice. Helium atoms diffuse away from the bubbles into nearby bulk UO2, thus forming a diffuse interface. For xenon bubbles, over-pressurized bubbles containing xenon can displace uranium atoms, which tend to aggregate around the xenon bubble as a pressure release mechanism. MD simulations of xenon atoms in pre-existing voids suggest that xenon atoms and the replaced uranium atoms occur in a 1:1 ratio, although kinetic factors may reduce that ratio depending on availability of xenon atoms and vacancies around the bubble. Finally, MD simulations suggest that for small bubbles (1-5 xenon atoms), the xenon bubble nucleus at UO2 grain-boundaries has much lower formation energy compared to that of bubbles of similar sizes in the bulk. However, when the xenon bubble grows into larger sizes, this energy difference is reduced.

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

    PubMed

    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

    2016-07-19

    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 (95)Zr and (97)Zr. By measuring both the perturbation of the (95)Mo/(96)Mo and (97)Mo/(96)Mo isotopic ratios and the total amount of molybdenum in the Trinity nuclear debris samples, it is possible to calculate the original concentrations of the (95)Zr and (97)Zr 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

  14. Quantification of Uncertainties due to 235,238U, 239,240,241Pu and Fission Products Nuclear Data Uncertainties for a PWR Fuel Assembly

    NASA Astrophysics Data System (ADS)

    da Cruz, D. F.; Rochman, D.; Koning, A. J.

    2014-04-01

    Uncertainty analysis on reactivity and discharged inventory for a typical PWR fuel element as a result of uncertainties in 235,238U, 239,240,241Pu, and fission products nuclear data was performed. The Total Monte-Carlo (TMC) method was applied using the deterministic transport code DRAGON. The nuclear data used in this study is from the JEFF-3.1 evaluations, with the exception of the nuclear data files for U, Pu and fission products isotopes, which are taken from the nuclear data library TENDL-2012. Results show that the calculated total uncertainty in keff (as result of uncertainties in nuclear data of the considered isotopes) is virtually independent on fuel burnp and amounts to 700 pcm. The uncertainties in inventory of the discharged fuel is dependent on the element considered and lies in the range 1-15% for most fission products, and is below 5% for the most important actinides.

  15. Exact Solution of Fractional Diffusion Model with Source Term used in Study of Concentration of Fission Product in Uranium Dioxide Particle

    NASA Astrophysics Data System (ADS)

    Fang, Chao; Cao, Jian-Zhu; Sun, Li-Feng

    2011-05-01

    The exact solution of fractional diffusion model with a location-independent source term used in the study of the concentration of fission product in spherical uranium dioxide (UO2) particle is built. The adsorption effect of the fission product on the surface of the UO2 particle and the delayed decay effect are also considered. The solution is given in terms of Mittag—Leffler function with finite Hankel integral transformation and Laplace transformation. At last, the reduced forms of the solution under some special physical conditions, which is used in nuclear engineering, are obtained and corresponding remarks are given to provide significant exact results to the concentration analysis of nuclear fission products in nuclear reactor.

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

    PubMed

    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

    2016-07-19

    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 (95)Zr and (97)Zr. By measuring both the perturbation of the (95)Mo/(96)Mo and (97)Mo/(96)Mo isotopic ratios and the total amount of molybdenum in the Trinity nuclear debris samples, it is possible to calculate the original concentrations of the (95)Zr and (97)Zr 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.

  17. The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors: A preliminary assessment of experiments HRB-17, HFR-B1, HFR-K6 and KORA

    SciTech Connect

    Myers, B.F.

    1995-09-01

    The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors has been measured in different laboratories under both irradiation and post irradiation conditions. The data from experiments HRB-17, HFR-B1, HFR-K6, and in the KORA facility are compared to assess their consistency and complimentarily. The experiments are consistent under comparable experimental conditions and reveal two general mechanisms involving exposed fuel kernels embedded in carbonaceous materials. One is manifest as a strong dependence of fission gas release on the partial pressure of water vapor below 1 kPa and the other, as a weak dependence above 1 kPa.

  18. Mitochondrial fusion but not fission regulates larval growth and synaptic development through steroid hormone production

    PubMed Central

    Sandoval, Hector; Yao, Chi-Kuang; Chen, Kuchuan; Jaiswal, Manish; Donti, Taraka; Lin, Yong Qi; Bayat, Vafa; Xiong, Bo; Zhang, Ke; David, Gabriela; Charng, Wu-Lin; Yamamoto, Shinya; Duraine, Lita; Graham, Brett H; Bellen, Hugo J

    2014-01-01

    Mitochondrial fusion and fission affect the distribution and quality control of mitochondria. We show that Marf (Mitochondrial associated regulatory factor), is required for mitochondrial fusion and transport in long axons. Moreover, loss of Marf leads to a severe depletion of mitochondria in neuromuscular junctions (NMJs). Marf mutants also fail to maintain proper synaptic transmission at NMJs upon repetitive stimulation, similar to Drp1 fission mutants. However, unlike Drp1, loss of Marf leads to NMJ morphology defects and extended larval lifespan. Marf is required to form contacts between the endoplasmic reticulum and/or lipid droplets (LDs) and for proper storage of cholesterol and ecdysone synthesis in ring glands. Interestingly, human Mitofusin-2 rescues the loss of LD but both Mitofusin-1 and Mitofusin-2 are required for steroid-hormone synthesis. Our data show that Marf and Mitofusins share an evolutionarily conserved role in mitochondrial transport, cholesterol ester storage and steroid-hormone synthesis. DOI: http://dx.doi.org/10.7554/eLife.03558.001 PMID:25313867

  19. True ternary fission

    NASA Astrophysics Data System (ADS)

    Vijayaraghavan, K. R.; Balasubramaniam, M.; von Oertzen, W.

    2015-04-01

    The study of the ternary fission of nuclei has received new interest recently. It is of general interest for nuclear dynamics, although the process is very rare. In the present work, we discuss the possibilities of true ternary fission (fragment masses A >30 ) in 252Cf for different mass splits. These mass splits are strongly favored in a collinear geometry. Based on the three cluster model (TCM), it is shown that the true ternary fission into fragments with almost equal masses is one of the possible fission modes in 252Cf . For general decays it is shown that the formation of the lightest fragment at the center has the highest probability. Further the formation of tin isotopes and/or other closed shell fragments are favored. For the decay products the presence of closed shell nuclei among the three fragments enhances the decay probabilities.

  20. Comparison of pka energy spectra, gas-atom production and damage energy deposition in neutron irradiation at various facilities

    NASA Astrophysics Data System (ADS)

    Nishiguchi, R.; Shimomura, Y.; Hahn, P. A.; Guinan, M. W.; Kiritani, M.

    1991-03-01

    By dividing neutron-energy spectrum into four energy groups, (I) <10 eV, (II) 10 eV to 0.1 MeV, (III) 0.1 MeV to 10 MeV and (IV) > 10 MeV, contributions to damage parameters (PKA spectrum, damage energy and gas-atom production) from each of the energy group were calculated for neutron irradiations at various facilities with the SPECTER code developed by Greenwood and Smither [1]. The normalized PKA spectra and the gas-atom productions were compared to examine differences in damage parameters. Such comparisons were carried out among (1) irradiations at various positions in different fission reactors (i.e. KUR, JOYO and FFTF-MOTA), and among (2) those at various fission reactors. Damage parameters were also calculated at STARFIRE fusion reactor and RTNS-II. A possible method to correlate damages at different fission reactors is discussed. It is suggested that damages in fusion reactor can be simulated by the superposition of irradiations with fission and D-T neutrons.

  1. A Multigroup Reaction Cross-Section Collapsing Code and Library of 154-Group Fission-Product Cross Sections.

    1983-03-23

    Version 01/02 The code reads multigroup cross sections from a compatible data file and collapses user-selected reaction cross sections to any few-group structure using one of a variety of user neutron flux spectrum options given below: Option Flux description 1 Built-in function including Maxwellian, fission, fusion and slowing-down regions and requiring user-specified parameters and energy-region boundaries. 2 Set of log-log flux-energy interpolation points read from input cross-section data file. 3 Set of log-log flux-energy interpolationmore » points read from user-supplied card input. 4 - 6 Histogram flux values read from user-supplied card input in arbitrary group structure in units of flux-per unit-energy, flux-per-unit lethargy, or integral group flux. LAFPX-E may be used to collapse any set of multigroup reaction cross sections furnished in the required format. However, the code was developed for, and is furnished with, a library of 154-group fission-product cross sections processed from ENDF/B-IV with a typical light water reactor (LWR) flux spectrum and temperature. Four-group radiative capture cross sections produced for LWR calculations are tabulated in the code documentation and are incorporated in the EPRI-CINDER data library, RSIC Code Package CCC-309.« less

  2. Antrim gas play, production expanding in Michigan

    SciTech Connect

    Not Available

    1994-05-30

    Devonian Antrim shale gas, the Michigan basin's dominant hydrocarbon play in terms of number of wells drilled for several years, shows every sign of continuing at a busy pace. About 3,500 Antrim completions now yield 350 MMcfd, more than 60% of Michigan's gas production. The outlook is for Antrim production to climb in the next 2--3 years to 500--600 MMcfd, about 1% of US gas output. These delivery numbers, slow decline rates, and expected producing life of 20--30 years has snagged pipelines attention. The growing production overtaxed local gathering facilities last fall, and the play recently got its first interstate outlet. Completion and production technology advances are improving well performance and trimming costs. Several hundred wells a year are likely to be drilled during the next few years. Production increases are coming from new wells, deepenings, and workovers. Numerous pipeline/gathering projects are planned in the area to handle the growing Antrim volumes. The paper discusses the development of this resource, efforts to extend the play, geology and production, drilling programs, and gas transportation.

  3. Tracy Gas field, new shallow production

    SciTech Connect

    Sterling, R.; Nahama, R. )

    1991-02-01

    The Tracy Gas field was the first commercial gas production in California. Amerada Hess found production in the Cretaceous Tracy Formation in 1934. Five production wells ultimately produced 13 BCF by 1959. In 1989, Nahama and Weagant Energy Company (NWEC) purchased seismic data acquired by Shell Oil in 1980. The data was reprocessed for amplitude vs. offset (AVO) analysis. The data showed a bright spot at 1.0 section that exhibit an AVO response indicative of the presence of gas. Log data suggested the possibility of gas in the based Miocene Zilch Sands in the Amerada F.D.L. {number sign}4 well drilled in Sec. 15. T2S, R5E drilled in 1934. Husky Oil found gas production in the Zilch with the Rossi {number sign}1 well drilled in 1980. The most obvious problem with following up Husky's well was the J. M. Young Cordoza {number sign}1, a dry hole, that was misplotted on all of the commercially available base maps. Nahama and Weagant Energy Company drilled the Tracy 1-15 well in January 1990 directionally to test the Zilch sands in Sec. 15 almost exactly where the dry hole had been misplotted, resulting in a discovery of 48 net ft of pay in the Zilch. The Miocene gas is present in basal Zilch sands that were deposited unconformably on the Cretaceous Blewett Formation. The gas is located on the west flank of the Tracy anticline which is formed by a 90{degree} bend in the Vernalis fault, a thrust fault with over 1,600 ft of vertical movement. The Vernalis fault had movement as recently as the Pliocene, based on seismic data. Due to the relative youth of the fault, deeper potential in the Winters and Lathrop formations is believed not to exist on the Tracy anticline.

  4. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses--Criticality (keff) Predictions

    SciTech Connect

    Scaglione, John M; Mueller, Don; Wagner, John C

    2011-01-01

    One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation - in particular, the availability and use of applicable measured data to support validation, especially for fission products. Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. U.S. Nuclear Regulatory Commission (NRC) staff have noted that the rationale for restricting their Interim Staff Guidance on burnup credit (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issue of validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach (both depletion and criticality) for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the criticality (k{sub eff}) validation approach, and resulting observations and recommendations. Validation of the isotopic composition (depletion) calculations is addressed in a companion paper at this conference. For criticality validation, the approach is to utilize (1) available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion (HTC) program to support validation of the principal actinides and (2) calculated sensitivities, nuclear data uncertainties, and the limited available fission

  5. Deep Atomic Binding (DAB) Hypothesis: A New Approach of Fission Product Chemistry

    SciTech Connect

    Ajlouni, Abdul-Wali M.S.

    2006-07-01

    Former studies assumed that, after fission process occurs, the highly ionized new born atoms (20-22 positive charge), ionize the media in which they pass through before becoming stable atoms in a manner similar to 4-MeV ?-particles. Via ordinary chemical reactions with the surroundings, each stable atom has a probability to form chemical compound. Since there are about 35 different elemental atoms created through fission processes, a large number of chemical species were suggested to be formed. But, these suggested chemical species were not found in the environment after actual releases of FP during accidents like TMI (USA, 1979), and Chernobyl (former USSR, 1986), also the models based on these suggested reactions and species could not interpret the behavior of these actual species. It is assumed here that the ionization states of the new born atoms and the long term high temperature were not dealt with in an appropriate way and they were the reasons of former models failure. Our new approach of Deep Atomic Binding (DAB) based on the following: 1-The new born atoms which are highly ionized, 10-12 electrons associated with each nucleus, having a large probability to create bonds between them to form molecules. These bonds are at the L, or M shells, and we call it DAB. 2-The molecules stay in the reactor at high temperatures for long periods, so they undergo many stages of composition and decomposition to form giant molecules. By applying DAB approach, field data from Chernobyl, TMI and nuclear detonations could be interpreted with a wide coincidence resulted. (author)

  6. Interconnecting compressors control coalbed gas production

    SciTech Connect

    Payton, R.; Niederhofer, J. )

    1992-10-05

    This paper reports that centralized compressors afford Taurus Exploration Inc.'s coalbed gas operations optimum control of gas production. Unlike satellite stations, the centralized system allows methane gas to e shifted from station to station via the interconnecting low-pressure pipeline network. The operations area encompasses approximately 40,000 acres, about 40 miles southwest of Birmingham, Ala. The project includes about 250-miles of low-pressure gas flow lines to almost 400 wells. The centralized system is less costly than a satellite station to build and operate. Unlike a satellite station that requires each compressor to have a complete set of ancillary equipment, the centralized system requires only one suction manifold, one dehydration setup, and one metering facility for every five compressor sets.

  7. Thermal transport in UO2 with defects and fission products by molecular dynamics simulations

    SciTech Connect

    Liu, Xiang-Yang; Cooper, Michael William Donald; Mcclellan, Kenneth James; Lashley, Jason Charles; Byler, Darrin David; Stanek, Christopher Richard; Andersson, Anders David Ragnar

    2015-10-14

    The importance of the thermal transport in nuclear fuel has motivated a wide range of experimental and modelling studies. In this report, the reduction of thermal transport in UO2 due to defects and fission products has been investigated using non-equilibrium MD simulations, with two sets of empirical potentials for studying the degregation of UO2 thermal conductivity including a Buckingham type interatomic potential and a recently developed EAM type interatomic potential. Additional parameters for U5+ and Zr4+ in UO2 have been developed for the EAM potential. The thermal conductivity results from MD simulations are then corrected for the spin-phonon scattering through Callaway model formulations. To validate the modelling results, comparison was made with experimental measurements on single crystal hyper-stoichiometric UO2+x samples.

  8. [Certain characteristics of the perinatal development of children whose parents were exposed to the action of uranium fission products].

    PubMed

    Evtushenko, N N

    1991-01-01

    Analysis was made of the character of the perinatal development of 238 children. In 65 of these, the parents were exposed to the action of uranium fission products in the basin of the river Techa in the Chelyabinsk region. The parents were radiated approximately 5 years before the children's birth. The mean equivalent dose for endocrine glands (including gonads) of the parents amounted to about 12 centiSievert U. As compared to the control group, the basic one manifested a higher incidence of placenta and umbilical cord pathology. In the group of the radiated persons' progeny, the number of small-weight children appeared greater; a larger physiological reduction of the body weight was recorded as were later times of funic residue falling off. PMID:1788018

  9. Determination of critical assembly absolute power using post-irradiation activation measurement of week-lived fission products.

    PubMed

    Košťál, Michal; Švadlenková, Marie; Milčák, Ján; Rypar, Vojtěch; Koleška, Michal

    2014-07-01

    The work presents a detailed comparison of calculated and experimentally determined net peak areas of longer-living fission products after 100 h irradiation on a reactor with power of ~630 W and several days cooling. Specifically the nuclides studied are (140)Ba, (103)Ru, (131)I, (141)Ce, (95)Zr. The good agreement between the calculated and measured net peak areas, which is better than in determination using short lived (92)Sr, is reported. The experiment was conducted on the VVER-1000 mock-up installed on the LR-0 reactor. The Monte Carlo approach has been used for calculations. The influence of different data libraries on results of calculation is discussed as well.

  10. Collection of fission and activation product elements from fresh and ocean waters: a comparison of traditional and novel sorbents

    SciTech Connect

    Johnson, Bryce E.; Santschi, Peter H.; Addleman, Raymond S.; Douglas, Matthew; Davidson, Joseph D.; Fryxell, Glen E.; Schwantes, Jon M.

    2010-04-01

    Monitoring natural waters for the inadvertent release of radioactive fission products produced as a result of nuclear power generation downstream from these facilities is essential for maintaining water quality. To this end, we evaluated sorbents for simultaneous in-situ large volume extraction of radionuclides with both soft (e.g., Ag) and hard metal (e.g., Co, Zr, Nb, Ba, and Cs) or anionic (e.g., Ru, Te, Sb) character. In this study, we evaluated a number of conventional and novel nanoporous sorbents in both fresh and salt waters. In most cases, the nanoporous sorbents demonstrated enhanced retention of analytes. Salinity had significant effects upon sorbent performance and was most significant for hard cations, specifically Cs and Ba. The presence of natural organic matter had little effect on the ability of chemisorbents to extract target elements.

  11. Bio-gas production from alligator weeds

    NASA Technical Reports Server (NTRS)

    Latif, A.

    1976-01-01

    Laboratory experiments were conducted to study the effect of temperature, sample preparation, reducing agents, light intensity and pH of the media, on bio-gas and methane production from the microbial anaerobic decomposition of alligator weeds (Alternanthera philoxeroides. Efforts were also made for the isolation and characterization of the methanogenic bacteria.

  12. New Methodology for Natural Gas Production Estimates

    EIA Publications

    2010-01-01

    A new methodology is implemented with the monthly natural gas production estimates from the EIA-914 survey this month. The estimates, to be released April 29, 2010, include revisions for all of 2009. The fundamental changes in the new process include the timeliness of the historical data used for estimation and the frequency of sample updates, both of which are improved.

  13. Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

    SciTech Connect

    Wagner, J.C.

    2001-08-02

    This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses.

  14. Energy Dependence of Fission Product Yields from 235U, 238U and 239Pu for Incident Neutron Energies Between 0.5 and 14.8 MeV

    NASA Astrophysics Data System (ADS)

    Gooden, M.; Arnold, C.; Bredeweg, T.; Vieira, D.; Wilhelmy, J.; Tonchev, A.; Stoyer, M.; Bhike, M.; Krishichayan, F.; Tornow, W.; Fowler, M.

    2015-10-01

    Under a joint collaboration between TUNL-LANL-LLNL, a set of absolute fission product yield measurements has been performed. The energy dependence of a number of cumulative fission product yields (FPY) have been measured using quasi-monoenergetic neutron beams for three actinide targets, 235U, 238U and 239Pu, between 0.5 and 14.8 MeV. The FPYs were measured by a combination of fission counting using specially designed dual-fission chambers and ?-ray counting. Each dual-fission chamber is a back-to-back ionization chamber encasing an activation target in the center with thin deposits of the same target isotope in each chamber. This method allows for the direct measurement of the total number of fissions in the activation target with no reference to the fission cross-section, thus reducing uncertainties. ?-ray counting of the activation target was performed on well-shielded HPGe detectors over a period of 2 months post irradiation to properly identify fission products. Reported are absolute cumulative fission product yields for incident neutron energies of 0.5, 1.37, 2.4, 3.6, 4.6, 5.5, 7.5, 8.9 and 14.8 MeV. These results are compared to previous measurements and theoretical estimates. This work was performed under the auspices of the USDoE by Los Alamos National Security, LLC under Contract DE-AC52-06NA25396.

  15. Natural gas production and consumption 1979

    SciTech Connect

    Not Available

    1981-01-01

    Total marketed production of natural gas in the United States during 1979 was 20,471 billion cubic feet, an increase of approximately 497 billion cubic feet, or 2.5 percent over 1978. Texas and Louisiana, the two leading producing states, accounted for 70.5 percent of total 1979 marketed production. In 1979, deliveries of natural gas to residential, commercial, industrial, electric utilities, and other consumers totaled 18,141 billion cubic feet. Total consumption, which includes lease, plant, and pipeline fuel in addition to deliveries to consumers, was 20,241 billion cubic feet in 1979 compared to 19,627 billion cubic feet in 1978, an increase of 3.1 percent. Movements of natural gas into and out of each state are presented. Louisiana accounted for the largest quantity of net deliveries, 5,107 billion cubic feet, followed by Texas and Oklahoma with net deliveries of 2,772 billion cubic feet and 914 billion cubic feet, respectively. Imports of natural gas by pipeline from Canada and as liquefied natural gas (LNG) from Algeria totaled 1,253 billion cubic feet in 1979. Total imports increased 288 billion cubic feet, or 29.8 percent, from 1978 levels. Exports of LNG to Japan and pipeline shipments to Canada and Mexico increased 6.0 percent from 52.5 billion cubic feet in 1978 to 55.7 billion cubic feet in 1979. LNG shipments to Japan accounted for 92.1 percent of total exports in 1979.

  16. Heterogeneous UO2 fuel irradiated up to a high burn-up: Investigation of the HBS and of fission product releases

    NASA Astrophysics Data System (ADS)

    Noirot, J.; Lamontagne, J.; Nakae, N.; Kitagawa, T.; Kosaka, Y.; Tverberg, T.

    2013-11-01

    measured in these areas remained low, ˜5%. The intermediate phase is a release area with moderate release. The average fraction of gas measured in the bubbles was around 20%. In the 235U-rich agglomerates, the Xe SIMS total was very low. This part is a release area. Sharp transitions between initial microstructure and the HBS, often inside one grain. Increase in the resulting grain size with increasing distance from the pellet periphery. The grain sizes are in fact consistent with the MOX measurements [2]. Increase in the bubble size with the increasing distance from the pellet periphery, consistent with the MOX measurements. Smaller bubbles tend to be found in the peripheral part of the 235U-rich agglomerates rather than in their central part. Sharp transition, around 0.5R, between the peripheral area where the conventional form of HBS forms in the 235U-rich agglomerates and the central part where much larger bubbles form and where the grain size is also clearly larger. Xe concentration of 0.1 wt% outside the bubbles in the HBS areas is consistent with the [2] MOX measurements at equivalent local burn-ups. The heterogeneous MOX fuels examinations have firmly established that the HBS can extend outside the Pu-rich agglomerates due to the implantation of fission products around these agglomerates. Similarly, it has been shown that the small Pu-rich agglomerates can remain with the initial microstructure even if there is a similar actual local burn-up, a large rate of fission products being implanted outside the agglomerates themselves so that the local fission product concentration remains low.In this 235U heterogeneous UO2, the Xe and Nd concentration levels reached at the HBS formation limit ranged between 0.8 wt% and 1.1 wt% for Xe and between 0.63 wt% and 0.83 wt% for Nd. These ranges are similar to what was reported in [23] for the UO2 rim. These limits are, however, slightly higher than those found for Pu-rich agglomerates in heterogeneous MOX fuels in [2] or in [24

  17. 21 CFR 173.350 - Combustion product gas.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 21 Food and Drugs 3 2014-04-01 2014-04-01 false Combustion product gas. 173.350 Section 173.350... Combustion product gas. The food additive combustion product gas may be safely used in the processing and... manufactured by the controlled combustion in air of butane, propane, or natural gas. The combustion...

  18. I-NERI ANNUAL TECHNICAL PROGRESS REPORT: 2006-002-K, Separation of Fission Products from Molten LiCl-KCl Salt Used for Electrorefining of Metal Fuels

    SciTech Connect

    S. Frank

    2009-09-01

    An attractive alternative to the once-through disposal of electrorefiner salt is to selectively remove the active fission products from the salt and recycle the salt back to the electrorefiner (ER). This would allow salt reuse for some number of cycles before ultimate disposal of the salt in a ceramic waste form. Reuse of ER salt would, thus, greatly reduce the volume of ceramic waste produced during the pyroprocessing of spent nuclear fuel. This final portion of the joint I-NERI research project is to demonstrate the separation of fission products from molten ER salt by two methods previously selected during phase two (FY-08) of this project. The two methods selected were salt/zeolite contacting and rare-earth fission product precipitation by oxygen bubbling. The ER salt used in these tests came from the Mark-IV electrorefiner used to anodically dissolved driver fuel from the EBR-II reactor on the INL site. The tests were performed using the Hot Fuel Dissolution Apparatus (HFDA) located in the main cell of the Hot Fuels Examination Facility (HFEF) at the Materials and Fuels complex on the INL site. Results from these tests were evaluated during a joint meeting of KAERI and INL investigators to provide recommendations as to the future direction of fission product removal from electrorefiner salt that accumulate during spent fuel treatment. Additionally, work continued on kinetic measurements of surrogate quaternary salt systems to provide fundamental kinetics on the ion exchange system and to expand the equilibrium model system developed during the first two phases of this project. The specific objectives of the FY09 I-NERI research activities at the INL include the following: • Perform demonstration tests of the selected KAERI precipitation and INL salt/zeolite contacting processes for fission product removal using radioactive, fission product loaded ER salt • Continue kinetic studies of the quaternary Cs/Sr-LiCl-KCl system to determine the rate of ion

  19. SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

    DOE PAGES

    Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; Allen, Todd R.

    2016-08-15

    Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy’s Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representativemore » of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Finally, insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.« less

  20. SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

    NASA Astrophysics Data System (ADS)

    Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; Allen, Todd R.

    2016-11-01

    Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy's Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representative of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.

  1. Designing an upgrade of the Medley setup for light-ion production and fission cross-section measurements

    NASA Astrophysics Data System (ADS)

    Jansson, K.; Gustavsson, C.; Al-Adili, A.; Hjalmarsson, A.; Andersson-Sundén, E.; Prokofiev, A. V.; Tarrío, D.; Pomp, S.

    2015-09-01

    Measurements of neutron-induced fission cross-sections and light-ion production are planned in the energy range 1-40 MeV at the upcoming Neutrons For Science (NFS) facility. In order to prepare our detector setup for the neutron beam with continuous energy spectrum, a simulation software was written using the Geant4 toolkit for both measurement situations. The neutron energy range around 20 MeV is troublesome when it comes to the cross-sections used by Geant4 since data-driven cross-sections are only available below 20 MeV but not above, where they are based on semi-empirical models. Several customisations were made to the standard classes in Geant4 in order to produce consistent results over the whole simulated energy range. Expected uncertainties are reported for both types of measurements. The simulations have shown that a simultaneous precision measurement of the three standard cross-sections H(n,n), 235U(n,f) and 238U(n,f) relative to each other is feasible using a triple layered target. As high resolution timing detectors for fission fragments we plan to use Parallel Plate Avalanche Counters (PPACs). The simulation results have put some restrictions on the design of these detectors as well as on the target design. This study suggests a fissile target no thicker than 2 μm (1.7 mg/cm2) and a PPAC foil thickness preferably less than 1 μm. We also comment on the usability of Geant4 for simulation studies of neutron reactions in this energy range.

  2. Mobile neutron/gamma waste assay system for characterization of waste containing transuranics, uranium, and fission/activation products

    SciTech Connect

    Davidson, D.R.; Haggard, D.; Lemons, C.

    1994-12-31

    A new integrated neutron/gamma assay system has been built for measuring 55-gallon drums at Pacific Northwest Laboratory. The system is unique because it allows simultaneous measurement of neutrons and gamma-rays. This technique also allows measurement of transuranics (TRU), uranium, and fission/activation products, screening for shielded Special Nuclear Material prior to disposal, and critically determinations prior to transportation. The new system is positioned on a platform with rollers and installed inside a trailer or large van to allow transportation of the system to the waste site instead of movement of the drums to the scanner. The ability to move the system to the waste drums is particularly useful for drum retrieval programs common to all DOE sites and minimizes transportation problems on the site. For longer campaigns, the system can be moved into a facility. The mobile system consists of two separate subsystems: a passive Segmented Gamma Scanner (SGS) and a {open_quotes}clam-shell{close_quotes} passive neutron counter. The SGS with high purity germanium detector and {sup 75}Se transmission source simultaneously scan the height of the drum allowing identification of unshieled {open_quotes}hot spots{close_quotes} in the drum or segments where the matrix is too dense for the transmission source to penetrate. Dense segments can flag shielding material that could be used to hide plutonium or uranium during the gamma analysis. The passive nuetron counter with JSR-12N Neutron Coincidence Analyzer measures the coincident neutrons from the spontaneous fission of even isotopes of plutonium. Because high-density shielding produces minimal absorption of neutrons, compared to gamma rays, the passive neutron portion of the system can detect shielded SNM. Measurements to evaluate the performance of the system are still underway at Pacific Northwest Laboratory.

  3. Ionizing radiation accelerates Drp1-dependent mitochondrial fission, which involves delayed mitochondrial reactive oxygen species production in normal human fibroblast-like cells

    SciTech Connect

    Kobashigawa, Shinko; Suzuki, Keiji; Yamashita, Shunichi

    2011-11-04

    Highlights: Black-Right-Pointing-Pointer We report first time that ionizing radiation induces mitochondrial dynamic changes. Black-Right-Pointing-Pointer Radiation-induced mitochondrial fission was caused by Drp1 localization. Black-Right-Pointing-Pointer We found that radiation causes delayed ROS from mitochondria. Black-Right-Pointing-Pointer Down regulation of Drp1 rescued mitochondrial dysfunction after radiation exposure. -- Abstract: Ionizing radiation is known to increase intracellular level of reactive oxygen species (ROS) through mitochondrial dysfunction. Although it has been as a basis of radiation-induced genetic instability, the mechanism involving mitochondrial dysfunction remains unclear. Here we studied the dynamics of mitochondrial structure in normal human fibroblast like cells exposed to ionizing radiation. Delayed mitochondrial O{sub 2}{sup {center_dot}-} production was peaked 3 days after irradiation, which was coupled with accelerated mitochondrial fission. We found that radiation exposure accumulated dynamin-related protein 1 (Drp1) to mitochondria. Knocking down of Drp1 expression prevented radiation induced acceleration of mitochondrial fission. Furthermore, knockdown of Drp1 significantly suppressed delayed production of mitochondrial O{sub 2}{sup {center_dot}-}. Since the loss of mitochondrial membrane potential, which was induced by radiation was prevented in cells knocking down of Drp1 expression, indicating that the excessive mitochondrial fission was involved in delayed mitochondrial dysfunction after irradiation.

  4. Spontaneous fission

    SciTech Connect

    Hoffman, D.C.

    1993-09-01

    The spontaneous fission (SF) of the heaviest actinides and the transactinides is of particular interest because of the dramatic changes in properties observed in the region of the heavy fermion isotopes and for still heavier elements. The existing experimental information on SF properties including half-life systematics, fragment kinetic-energy and mass-yield distributions, prompt neutron emission, and gamma emission will be reviewed. Possibility for extending studies of SF properties to other regions are considered and the potential for obtaining additional information about low-energy fission properties is discussed.

  5. Benchmarking nuclear fission theory

    DOE PAGES

    Bertsch, G. F.; Loveland, W.; Nazarewicz, W.; Talou, P.

    2015-05-14

    We suggest a small set of fission observables to be used as test cases for validation of theoretical calculations. Thus, the purpose is to provide common data to facilitate the comparison of different fission theories and models. The proposed observables are chosen from fission barriers, spontaneous fission lifetimes, fission yield characteristics, and fission isomer excitation energies.

  6. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Isotopic Composition Predictions

    SciTech Connect

    Radulescu, Georgeta; Gauld, Ian C; Ilas, Germina; Wagner, John C

    2011-01-01

    The expanded use of burnup credit in the United States (U.S.) for storage and transport casks, particularly in the acceptance of credit for fission products, has been constrained by the availability of experimental fission product data to support code validation. The U.S. Nuclear Regulatory Commission (NRC) staff has noted that the rationale for restricting the Interim Staff Guidance on burnup credit for storage and transportation casks (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issues of burnup credit criticality validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the isotopic composition (depletion) validation approach and resulting observations and recommendations. Validation of the criticality calculations is addressed in a companion paper at this conference. For isotopic composition validation, the approach is to determine burnup-dependent bias and uncertainty in the effective neutron multiplication factor (keff) due to bias and uncertainty in isotopic predictions, via comparisons of isotopic composition predictions (calculated) and measured isotopic compositions from destructive radiochemical assay utilizing as much assay data as is available, and a best-estimate Monte Carlo based method. This paper (1) provides a detailed description of the burnup credit isotopic validation approach and its technical bases, (2) describes the application of the approach for

  7. Spallation reaction study for fission products in nuclear waste: Cross section measurements for 137Cs and 90Sr on proton and deuteron

    NASA Astrophysics Data System (ADS)

    Wang, H.; Otsu, H.; Sakurai, H.; Ahn, D. S.; Aikawa, M.; Doornenbal, P.; Fukuda, N.; Isobe, T.; Kawakami, S.; Koyama, S.; Kubo, T.; Kubono, S.; Lorusso, G.; Maeda, Y.; Makinaga, A.; Momiyama, S.; Nakano, K.; Niikura, M.; Shiga, Y.; Söderström, P.-A.; Suzuki, H.; Takeda, H.; Takeuchi, S.; Taniuchi, R.; Watanabe, Ya.; Watanabe, Yu.; Yamasaki, H.; Yoshida, K.

    2016-03-01

    We have studied spallation reactions for the fission products 137Cs and 90Sr for the purpose of nuclear waste transmutation. The spallation cross sections on the proton and deuteron were obtained in inverse kinematics for the first time using secondary beams of 137Cs and 90Sr at 185 MeV/nucleon at the RIKEN Radioactive Isotope Beam Factory. The target dependence has been investigated systematically, and the cross-section differences between the proton and deuteron are found to be larger for lighter spallation products. The experimental data are compared with the PHITS calculation, which includes cascade and evaporation processes. Our results suggest that both proton- and deuteron-induced spallation reactions are promising mechanisms for the transmutation of radioactive fission products.

  8. 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

  9. Bimodal fission

    SciTech Connect

    Hulet, E.K.

    1989-04-19

    In recent years, we have measured the mass and kinetic-energy distributions from the spontaneous fission of /sup 258/Fm, /sup 259/Md, /sup 260/Md, /sup 258/No, /sup 262/No, and /sup 260/(104). All are observed to fission with a symmetrical division of mass, whereas the total-kinetic-energy (TKE) distributions strongly deviated from the Gaussian shape characteristically found in the fission of all other actinides. When the TKE distributions are resolved into two Gaussians the constituent peaks lie near 200 and near 233 MeV. We conclude two modes or bimodal fission is occurring in five of the six nuclides studied. Both modes are possible in the same nuclides, but one generally predominates. We also conclude the low-energy but mass-symmetrical mode is likely to extend to far heavier nuclei; while the high-energy mode will be restricted to a smaller region, a region of nuclei defined by the proximity of the fragments to the strong neutron and proton shells in /sup 132/Sn. 16 refs., 7 figs., 1 tab.

  10. 21 CFR 173.350 - Combustion product gas.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 21 Food and Drugs 3 2012-04-01 2012-04-01 false Combustion product gas. 173.350 Section 173.350... CONSUMPTION Specific Usage Additives § 173.350 Combustion product gas. The food additive combustion product gas may be safely used in the processing and packaging of the foods designated in paragraph (c)...

  11. 21 CFR 173.350 - Combustion product gas.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 21 Food and Drugs 3 2011-04-01 2011-04-01 false Combustion product gas. 173.350 Section 173.350... CONSUMPTION Specific Usage Additives § 173.350 Combustion product gas. The food additive combustion product gas may be safely used in the processing and packaging of the foods designated in paragraph (c)...

  12. 21 CFR 173.350 - Combustion product gas.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 21 Food and Drugs 3 2013-04-01 2013-04-01 false Combustion product gas. 173.350 Section 173.350... CONSUMPTION Specific Usage Additives § 173.350 Combustion product gas. The food additive combustion product gas may be safely used in the processing and packaging of the foods designated in paragraph (c)...

  13. 21 CFR 173.350 - Combustion product gas.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 21 Food and Drugs 3 2010-04-01 2009-04-01 true Combustion product gas. 173.350 Section 173.350... CONSUMPTION Specific Usage Additives § 173.350 Combustion product gas. The food additive combustion product gas may be safely used in the processing and packaging of the foods designated in paragraph (c)...

  14. CO Methanation for Synthetic Natural Gas Production.

    PubMed

    Kambolis, Anastasios; Schildhauer, Tilman J; Kröcher, Oliver

    2015-01-01

    Energy from woody biomass could supplement renewable energy production towards the replacement of fossil fuels. A multi-stage process involving gasification of wood and then catalytic transformation of the producer gas to synthetic natural gas (SNG) represents progress in this direction. SNG can be transported and distributed through the existing pipeline grid, which is advantageous from an economical point of view. Therefore, CO methanation is attracting a great deal of attention and much research effort is focusing on the understanding of the process steps and its further development. This short review summarizes recent efforts at Paul Scherrer Institute on the understanding of the reaction mechanism, the catalyst deactivation, and the development of catalytic materials with benign properties for CO methanation. PMID:26598405

  15. Mesoscale Benchmark Demonstration Problem 1: Mesoscale Simulations of Intra-granular Fission Gas Bubbles in UO2 under Post-irradiation Thermal Annealing

    SciTech Connect

    Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert; Gao, Fei; Sun, Xin; Tonks, Michael; Biner, Bullent; Millet, Paul; Tikare, Veena; Radhakrishnan, Balasubramaniam; Andersson , David

    2012-04-11

    A study was conducted to evaluate the capabilities of different numerical methods used to represent microstructure behavior at the mesoscale for irradiated material using an idealized benchmark problem. The purpose of the mesoscale benchmark problem was to provide a common basis to assess several mesoscale methods with the objective of identifying the strengths and areas of improvement in the predictive modeling of microstructure evolution. In this work, mesoscale models (phase-field, Potts, and kinetic Monte Carlo) developed by PNNL, INL, SNL, and ORNL were used to calculate the evolution kinetics of intra-granular fission gas bubbles in UO2 fuel under post-irradiation thermal annealing conditions. The benchmark problem was constructed to include important microstructural evolution mechanisms on the kinetics of intra-granular fission gas bubble behavior such as the atomic diffusion of Xe atoms, U vacancies, and O vacancies, the effect of vacancy capture and emission from defects, and the elastic interaction of non-equilibrium gas bubbles. An idealized set of assumptions was imposed on the benchmark problem to simplify the mechanisms considered. The capability and numerical efficiency of different models are compared against selected experimental and simulation results. These comparisons find that the phase-field methods, by the nature of the free energy formulation, are able to represent a larger subset of the mechanisms influencing the intra-granular bubble growth and coarsening mechanisms in the idealized benchmark problem as compared to the Potts and kinetic Monte Carlo methods. It is recognized that the mesoscale benchmark problem as formulated does not specifically highlight the strengths of the discrete particle modeling used in the Potts and kinetic Monte Carlo methods. Future efforts are recommended to construct increasingly more complex mesoscale benchmark problems to further verify and validate the predictive capabilities of the mesoscale modeling

  16. Bio gas oil production from waste lard.

    PubMed

    Hancsók, Jeno; Baladincz, Péter; Kasza, Tamás; Kovács, Sándor; Tóth, Csaba; Varga, Zoltán

    2011-01-01

    Besides the second generations bio fuels, one of the most promising products is the bio gas oil, which is a high iso-paraffin containing fuel, which could be produced by the catalytic hydrogenation of different triglycerides. To broaden the feedstock of the bio gas oil the catalytic hydrogenation of waste lard over sulphided NiMo/Al(2)O(3) catalyst, and as the second step, the isomerization of the produced normal paraffin rich mixture (intermediate product) over Pt/SAPO-11 catalyst was investigated. It was found that both the hydrogenation and the decarboxylation/decarbonylation oxygen removing reactions took place but their ratio depended on the process parameters (T = 280-380°C, P = 20-80 bar, LHSV = 0.75-3.0  h(-1) and H(2)/lard ratio: 600  Nm(3)/m(3)). In case of the isomerization at the favourable process parameters (T = 360-370°C, P = 40-50 bar, LHSV = 1.0  h(-1) and H(2)/hydrocarbon ratio: 400  Nm(3)/m(3)) mainly mono-branching isoparaffins were obtained. The obtained products are excellent Diesel fuel blending components, which are practically free of heteroatoms. PMID:21403875

  17. Isotope ratio analysis of actinides, fission products, and geolocators by high-efficiency multi-collector thermal ionization mass spectrometry

    NASA Astrophysics Data System (ADS)

    Bürger, S.; Riciputi, L. R.; Bostick, D. A.; Turgeon, S.; McBay, E. H.; Lavelle, M.

    2009-09-01

    A ThermoFisher "Triton" multi-collector thermal ionization mass spectrometer (MC-TIMS) was evaluated for trace and ultra-trace level isotope ratio analysis of actinides (uranium, plutonium, and americium), fission products and geolocators (strontium, cesium, and neodymium). Total efficiencies (atoms loaded to ions detected) of up to 0.5-2% for U, Pu, and Am, and 1-30% for Sr, Cs, and Nd can be reported employing resin bead load techniques onto flat ribbon Re filaments or resin beads loaded into a millimeter-sized cavity drilled into a Re rod. This results in detection limits of <0.1 fg (104 atoms to 105 atoms) for 239-242+244Pu, 233+236U, 241-243Am, 89,90Sr, and 134,135,137Cs, and <=1 pg for natural Nd isotopes (limited by the chemical processing blank) using a secondary electron multiplier (SEM) or multiple-ion counters (MICs). Relative standard deviations (RSD) as small as 0.1% and abundance sensitivities of 1 × 106 or better using a SEM are reported here. Precisions of RSD [approximate]0.01-0.001% using a multi-collector Faraday cup array can be achieved at sub-nanogram concentrations for strontium and neodymium and are suitable to gain crucial geolocation information. The analytical protocols reported herein are of particular value for nuclear forensic and nuclear safeguard applications.

  18. Isotope ratio analysis of actinides, fission products, and geolocators by high-efficiency multi-collector thermal ionization mass spectrometry

    SciTech Connect

    Bürger, Stefan; Riciputi, Lee R; Bostick, Debra A; Turgeon, Steven; McBay, Eddie H; Lavelle, Mark

    2009-01-01

    A ThermoFisher 'Triton' multi-collector thermal ionization mass spectrometer (MC-TIMS) was evaluated for trace and ultra-trace level isotoperatioanalysis of actinides (uranium, plutonium, and americium), fission products and geolocators (strontium, cesium, and neodymium). Total efficiencies (atoms loaded to ions detected) of up to 0.5-2% for U, Pu, and Am, and 1-30% for Sr, Cs, and Nd can be reported employing resin bead load techniques onto flat ribbon Re filaments or resin beads loaded into a millimeter-sized cavity drilled into a Re rod. This results in detection limits of <0.1 fg (10{sup 4} atoms to 10{sup 5} atoms) for {sup 239-242+244}Pu, {sup 233+236}U, {sup 241-243}Am, {sup 89,90}Sr, and {sup 134,135,137}Cs, and {le} 1 pg for natural Nd isotopes (limited by the chemical processing blank) using a secondary electron multiplier (SEM) or multiple-ion counters (MICs). Relative standard deviations (RSD) as small as 0.1% and abundance sensitivities of 1 x 10{sup 6} or better using a SEM are reported here. Precisions of RSD {approx} 0.01-0.001% using a multi-collector Faraday cup array can be achieved at sub-nanogram concentrations for strontium and neodymium and are suitable to gain crucial geolocation information. The analytical protocols reported herein are of particular value for nuclear forensic and nuclear safeguard applications.

  19. Development of a multi-layer diffusion couple to study fission product transport in β-SiC

    NASA Astrophysics Data System (ADS)

    Dwaraknath, S.; Was, G. S.

    2014-01-01

    A multi-layer diffusion couple was designed to study fission product diffusion behavior while avoiding the pitfalls of direct ion implantation. Thin films of highly anisotropic pyrolytic carbon (PyC) were deposited onto CVD β-SiC substrates. The PyC films were subsequently implanted with 400 keV silver, cesium, strontium, europium, or iodine at 22 °C to a dose of 1016 cm-2, such that the implanted species resided wholly within the PyC layer. The samples were then coated with 50 nm of SiC via plasma enhanced CVD (PECVD) to retain the implanted species during post-deposition annealing treatments. The design allows for high spatial resolution tracking of the implanted specie using Rutherford backscattering spectrometry. Annealing at 1100 °C for 10 h resulted in retention of 100% of implanted cesium, strontium, europium and iodine, and 70% of silver. This diffusion couple design provides the opportunity to determine diffusion coefficients of FPs in PyC and SiC and the role of the PyC/SiC interface in FP transport.

  20. Using tea as an artificial urine in a Canadian performance testing program for fission/activation products.

    PubMed

    Daka, Joseph N; Moodie, Gerry; DiNardo, Anthony; Kramer, Gary H

    2014-12-01

    In recent years, the National Calibration Reference Centre for Bioassay and In Vivo Monitoring (NCRC) at the Radiation Protection Bureau (RPB), Health Canada, has been conducting investigations with black tea to develop a matrix that can be used to replace urine in each of the following performance testing programs (PTP): (1) tritium, (2) carbon-14, (3) the DUAL (i.e., 3H/14C), and (4) fission/activation products (F/AP). A 1% tea solution with thimerosal, which had worked successfully for tritium, carbon-14, and the DUAL, was selected and tested for the F/AP PTP because of its similarity to urine in color and UV-VIS spectra. However, application of this tea to samples of the F/AP program containing 133Ba, 137Cs, 57Co, and 60Co produced precipitates, which was an unexpected result. Further experiments showed that replacement of thimerosal with an alcohol at about 5% eliminated the precipitation problem. The alcohol can be ethanol, methanol, or isopropanol. In the experiments, the 1% tea, preserved with alcohol, remained clear and stable for at least 100 d. The duration of each PTP for the NCRC is limited to 90 d. Application of the CNSC S-106 regulatory standard to the tea produced acceptable accuracy and precision results. It was concluded that a suitable tea matrix for the F/AP program had been found.

  1. Wet deposition of fission-product isotopes to North America from the Fukushima Dai-ichi incident, March 2011

    USGS Publications Warehouse

    Wetherbee, Gregory A.; Gay, David A.; Debey, Timothy M.; Lehmann, Christopher M.B.; Nilles, Mark A.

    2012-01-01

    Using the infrastructure of the National Atmospheric Deposition Program (NADP), numerous measurements of radionuclide wet deposition over North America were made for 167 NADP sites before and after the Fukushima Dai-ichi Nuclear Power Station incident of March 12, 2011. For the period from March 8 through April 5, 2011, wet-only precipitation samples were collected by NADP and analyzed for fission-product isotopes within whole-water and filterable solid samples by the United States Geological Survey using gamma spectrometry. Variable amounts of 131I, 134Cs, or 137Cs were measured at approximately 21% of sampled NADP sites distributed widely across the contiguous United States and Alaska. Calculated 1- to 2-week individual radionuclide deposition fluxes ranged from 0.47 to 5100 Becquerels per square meter during the sampling period. Wet deposition activity was small compared to measured activity already present in U.S. soil. NADP networks responded to this complex disaster, and provided scientifically valid measurements that are comparable and complementary to other networks in North America and Europe.

  2. Advances in Development of the Fission Product Extraction Process for the Separation of Cesium and Strontium from Spent Nuclear Fuel

    SciTech Connect

    JAck D. Law

    2007-09-01

    The Fission Product Extraction (FPEX) Process is being developed as part of the United States Department of Energy Advanced Fuel Cycle Initiative for the simultaneous separation of cesium (Cs) and strontium (Sr) from spent light water reactor (LWR) fuel. Separation of the Cs and Sr will reduce the short-term heat load in a geological repository, and when combined with the separation of americium (Am) and curium (Cm), could increase the capacity of the geological repository by a factor of approximately 100. The FPEX process is based on two highly specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. Results of flowsheet testing of the FPEX process with a simulated feed solution in 3.3-cm centrifugal contactors are detailed. Removal efficiencies, distribution coefficient data, coextraction of metals, and process hydrodynamic performance are discussed along with recommendations for future flowsheet testing with actual spent nuclear fuel.

  3. METHOD OF SEPARATING FISSION PRODUCTS FROM FUSED BISMUTH-CONTAINING URANIUM

    DOEpatents

    Wiswall, R.H.

    1958-06-24

    A process is described for removing metal selectively from liquid metal compositions. The method effects separation of flssion product metals selectively from dilute solution in fused bismuth, which contains uraniunn in solution without removal of more than 1% of the uranium. The process comprises contacting the fused bismuth with a fused salt composition consisting of sodium, potassium and lithium chlorides, adding to fused bismuth and molten salt a quantity of bismuth chloride which is stoichiometrically required to convert the flssion product metals to be removed to their chlorides which are more stable in the fused salt than in the molten metal and are, therefore, preferentially taken up in the fused salt phase.

  4. GGA+U study of uranium mononitride: A comparison of the U-ramping and occupation matrix schemes and incorporation energies of fission products

    NASA Astrophysics Data System (ADS)

    Claisse, Antoine; Klipfel, Marco; Lindbom, Niclas; Freyss, Michel; Olsson, Pär

    2016-09-01

    Uranium mononitride is studied in the DFT + U framework. Its ground state is investigated and a study of the incorporation of diverse fission products in the crystal is conducted. The U-ramping and occupation matrix control (OMC) schemes are used to eliminate metastable states. Beyond a certain amount of introduced correlation, the OMC scheme starts to find a lower total energy. The OMC scheme is chosen for the second part of this study. Furthermore, the influence of the magnetic ordering is studied using the U-ramping method, showing that antiferromagnetic order is the most stable one when the U parameter is larger than 1.75 eV. The effect on the density of states is investigated and elastic constants are provided for comparison with other methods and experiments. The incorporation energies of fission products in different defect configurations are calculated and these energies are corrected to take into account the limited size of the supercell.

  5. Future challenges for nuclear data research in fission (u)

    SciTech Connect

    Chadwick, Mark B

    2010-01-01

    I describe some high priority research areas in nuclear fission, where applications in nuclear reactor technologies and in modeling criticality in general are demanding higher accuracies in our databases. We focus on fission cross sections, fission neutron spectra, and fission product data.

  6. Fission meter

    DOEpatents

    Rowland, Mark S.; Snyderman, Neal J.

    2012-04-10

    A neutron detector system for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly into information that a first responder can use to discriminate materials. The system comprises counting neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source.

  7. A repository released-dose model for the evaluation of long-lived fission product transmutation effectiveness

    SciTech Connect

    Davidson, J.W.

    1995-07-01

    A methodology has been developed to quantify the total integrated dose due to a radionuclide species i emplaced in a geologic repository; the focus is on the seven long-lived fission products (LLFPs). The methodology assumes continuous exposure water contaminated with species i at the accessible environment (i.e., just beyond the geologic barrier afforded by the geologic repository). The dose integration is performed out to a reference post-release time. The integrated dose is a function of the total initial inventory of radionuclide i the repository, the time at which complete and instantaneous failure of the engineered barrier (e.g., waste canister) in, a geologic repository occurs, the fractional dissolution rate (from waste solid form) of radionuclide i in ground water, the ground water travel time to the accessible environment, the retardation factor (sorption on the geologic media) for radionuclide i, the time after radionuclide begins to enter the biosphere. In order to assess relative dose, the ratio of total integrated dose to that for a reference LLFP species j (e.g., {sup 99}Tc) was defined. This ratio is a measure of the relative benefit of transmutation of other LLFPs compared to {sup 99}Tc. This methodology was further developed in order to quantify the integrated dose reduction per neutron utilized for LLFP transmutation in accelerator-driven transmutation technologies (ADTT). This measure of effectiveness is a function of the integrated dose due to LLFP species i, the number of total captures in LLFP species i chain per LLFP nuclide fed to the chain at equilibrium, and the number of total captures in related transmutation product (TP) chains per capture in the LLFP species i chain. To assess relative transmutation effectiveness, the ratio of integrated dose reduction per neutron utilization to that for a reference LLFP species j (e.g., {sup 99}Tc) was defined. This relative measure of effectiveness was evaluated LLFP transmutation strategy.

  8. Correlation between Asian Dust and Specific Radioactivities of Fission Products Included in Airborne Samples in Tokushima, Shikoku Island, Japan, Due to the Fukushima Nuclear Accident

    SciTech Connect

    Sakama, M.; Nagano, Y.; Kitade, T.; Shikino, O.; Nakayama, S.

    2014-06-15

    Radioactive fission product {sup 131}I released from the Fukushima Daiichi Nuclear Power Plants (FD-NPP) was first detected on March 23, 2011 in an airborne aerosol sample collected at Tokushima, Shikoku Island, located in western Japan. Two other radioactive fission products, {sup 134}Cs and {sup 137}Cs were also observed in a sample collected from April 2 to 4, 2011. The maximum specific radioactivities observed in this work were about 2.5 to 3.5 mBq×m{sup -3} in a airborne aerosol sample collected on April 6. During the course of the continuous monitoring, we also made our first observation of seasonal Asian Dust and those fission products associated with the FDNPP accident concurrently from May 2 to 5, 2011. We found that the specific radioactivities of {sup 134}Cs and {sup 137}Cs decreased drastically only during the period of Asian Dust. And also, it was found that this trend was very similar to the atmospheric elemental concentration (ng×m{sup -3}) variation of stable cesium ({sup 133}Cs) quantified by elemental analyses using our developed ICP-DRC-MS instrument.

  9. Advanced electron microscopic techniques applied to the characterization of irradiation effects and fission product identification of irradiated TRISO coated particles from the AGR-1 experiment

    SciTech Connect

    Rooyen, I.J. van; Lillo, T.M.; Trowbridge, T.L.; Madden, J.M.; Wu, Y.Q.; Goran, D.

    2013-07-01

    Preliminary electron microscopy of coated fuel particles from the AGR-1 experiment was conducted using characterization techniques such as scanning electron microscopy (SEM), transmission electron microscopy (TEM), energy dispersive spectroscopy (EDS), and wavelength dispersive spectroscopy (WDS). Microscopic quantification of fission-product precipitates was performed. Although numerous micro- and nano-sized precipitates observed in the coating layers during initial SEM characterization of the cross-sections, and in subsequent TEM diffraction patterns, were indexed as UPd{sub 2}Si{sub 2}, no Ag was conclusively found. Additionally, characterization of these precipitates highlighted the difficulty of measuring low concentrations of Ag in precipitates in the presence of significantly higher concentrations of Pd and U. The electron microscopy team followed a multi-directional and phased approach in the identification of fission products in irradiated TRISO fuel. The advanced electron microscopy techniques discussed in this paper, not only demonstrate the usefulness of the equipment (methods) as relevant research tools, but also provide relevant scientific results which increase the knowledge about TRISO fuel particles microstructure and fission products transport.

  10. Correlation between Asian Dust and Specific Radioactivities of Fission Products Included in Airborne Samples in Tokushima, Shikoku Island, Japan, Due to the Fukushima Nuclear Accident

    NASA Astrophysics Data System (ADS)

    Sakama, M.; Nagano, Y.; Kitade, T.; Shikino, O.; Nakayama, S.

    2014-06-01

    Radioactive fission product 131I released from the Fukushima Daiichi Nuclear Power Plants (FD-NPP) was first detected on March 23, 2011 in an airborne aerosol sample collected at Tokushima, Shikoku Island, located in western Japan. Two other radioactive fission products, 134Cs and 137Cs were also observed in a sample collected from April 2 to 4, 2011. The maximum specific radioactivities observed in this work were about 2.5 to 3.5 mBq×m-3 in a airborne aerosol sample collected on April 6. During the course of the continuous monitoring, we also made our first observation of seasonal Asian Dust and those fission products associated with the FDNPP accident concurrently from May 2 to 5, 2011. We found that the specific radioactivities of 134Cs and 137Cs decreased drastically only during the period of Asian Dust. And also, it was found that this trend was very similar to the atmospheric elemental concentration (ng×m-3) variation of stable cesium (133Cs) quantified by elemental analyses using our developed ICP-DRC-MS instrument.

  11. Expanding Canadian natural gas production will strengthen growth of LP-gas industry

    SciTech Connect

    Hawkins, D.J. )

    1994-01-01

    In 1992, over 86% of Canadian propane and 70% of Canadian butane production originated in gas plants. Propane and butane production not recovered at gas plants is recovered in other processing facilities, primarily refineries and heavy oil upgraders. As a result, supplies of both products are largely tied to natural gas production, and the outlook for natural gas therefore provides the basis for any discussion on the outlook for gas processing and NGL industry infrastructure. The paper discusses gas processing, economies of scale, NGL supply, expected declines, industry structure and infrastructure, the two major centers of the Canadian NGL industry, new shippers, and required pipeline expansion.

  12. High rate of methane leakage from natural gas production

    NASA Astrophysics Data System (ADS)

    Balcerak, Ernie

    2013-10-01

    Natural gas production is growing as the United States seeks domestic sources of relatively clean energy. Natural gas combustion produces less carbon dioxide emissions than coal or oil for the amount of energy produced. However, one source of concern is that some natural gas leaks to the atmosphere from the extraction point, releasing methane, a potent greenhouse gas.

  13. Fifty years with nuclear fission

    SciTech Connect

    Behrens, J.W.; Carlson, A.D. )

    1989-01-01

    The news of the discovery of nucler fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fiftieth anniversary of its discovery by holding a topical meeting entitled, Fifty years with nuclear fission,'' in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent developments in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicating a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two full days of sessions (April 27 and 28) at the main sites of the NIST in Gaithersburg, Maryland. The wide range of topics covered by Volume 2 of this topical meeting included plenary invited, and contributed sessions entitled, Nuclear fission -- a prospective; reactors II; fission science II; medical and industrial applications by by-products; reactors and safeguards; general research, instrumentation, and by-products; and fission data, astrophysics, and space applications. The individual papers have been cataloged separately.

  14. Fifty years with nuclear fission

    SciTech Connect

    Behrens, J.W.; Carlson, A.D. )

    1989-01-01

    The news of the discovery of nuclear fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fifieth anniversary of its discovery by holding a topical meeting entitled, Fifty Years with Nuclear Fission,'' in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent development in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicated a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two fully days of sessions (April 27 and 28) at the main site of the NIST in Gaithersburg, Maryland. The wide range of topics covered in this Volume 1 by this topical meeting included plenary invited, and contributed sessions entitled: Preclude to the First Chain Reaction -- 1932 to 1942; Early Fission Research -- Nuclear Structure and Spontaneous Fission; 50 Years of Fission, Science, and Technology; Nuclear Reactors, Secure Energy for the Future; Reactors 1; Fission Science 1; Safeguards and Space Applications; Fission Data; Nuclear Fission -- Its Various Aspects; Theory and Experiments in Support of Theory; Reactors and Safeguards; and General Research, Instrumentation, and By-Product. The individual papers have been cataloged separately.

  15. Gas-well production decline in multiwell reservoirs

    SciTech Connect

    Aminian, K.; Ameri, S. ); Stark, J.J. ); Yost, A.B. II )

    1990-12-01

    This paper introduces a pseudosteady-state constant-pressure solution for gas wells. The solution was used to develop a type-curve-based method to history match and predict multiwell gas reservoir production. Good agreements between the predicted and actual gas well production rates were obtained.

  16. Effects of gas bubble production on heat transfer from a volumetrically heated liquid pool

    NASA Astrophysics Data System (ADS)

    Bull, Geoffrey R.

    Aqueous solutions of uranium salts may provide a new supply chain to fill potential shortfalls in the availability of the most common radiopharmaceuticals currently in use worldwide, including Tc99m which is a decay product of Mo99. The fissioning of the uranium in these solutions creates Mo99 but also generates large amounts of hydrogen and oxygen from the radiolysis of the water. When the dissolved gases reach a critical concentration, bubbles will form in the solution. Bubbles in the solution affect both the fission power and the heat transfer out of the solution. As a result, for safety and production calculations, the effects of the bubbles on heat transfer must be understood. A high aspect ratio tank was constructed to simulate a section of an annulus with heat exchangers on the inner and outer steel walls to provide cooling. Temperature measurements via thermocouples inside the tank and along the outside of the steel walls allowed the calculation of overall and local heat transfer coefficients. Different air injection manifolds allowed the exploration of various bubble characteristics and patterns on heat transfer from the pool. The manifold type did not appear to have significant impact on the bubble size distributions in water. However, air injected into solutions of magnesium sulfate resulted in smaller bubble sizes and larger void fractions than those in water at the same injection rates. One dimensional calculations provide heat transfer coefficient values as functions of the superficial gas velocity in the pool.

  17. Volatilities of ruthenium, iodine, and technetium on calcining fission product nitrate wastes

    SciTech Connect

    Rimshaw, S.J.; Case, F.N.

    1980-01-01

    Various high-level nitrate wastes were subjected to formic acid denitration. Formic acid reacts with the nitrate anion to yield noncondensable, inert gases according to the following equation: 4 HCOOH + 2 HNO/sub 3/ ..-->.. N/sub 2/O + 4 CO/sub 2/ + 5 H/sub 2/O. These gases can be scrubbed free of /sup 106/Ru, /sup 131/I, and /sup 99/Tc radioactivities prior to elimination from the plant by passage through HEPA filters. The formation of deleterious NO/sub x/ is avoided. Moreover, formic acid reduces ruthenium to a lower valence state with a sharp reduction in RuO/sub 4/ volatility during subsequent calcination of the pretreated waste. It is shown that a minimum of 3% of RuO/sub 4/ in an off-gas stream reacts with Davison silica gel (Grade 40) to give a fine RuO/sub 2/ aerosol having a particle size of 0.5 ..mu... This RuO/sub 2/ aerosol passes through water or weak acid scrub solutions but is trapped by a caustic scrub solution. Iodine volatilizes almost completely on calcining an acidic waste, and the iodine volatility increases with increasing calcination temperature. On calcining an alkaline sodium nitrate waste the iodine volatility is about an order of magnitude lower, with a relatively low iodine volatility of 0.39% at a calcination temperature of 250/sup 0/C and a moderate volatility of 9.5% at 600/sup 0/C. Volatilities of /sup 99/Tc were generally <1% on calcining acidic or basic wastes at temperatures of 250 to 600/sup 0/C. Data are presented to indicate that /sup 99/Tc concentrates in the alkaline sodium nitrate supernatant waste, with approx. 10 mg /sup 99/Tc being associated with each curie of /sup 137/Cs present in the waste. It is shown that lutidine (2,4 dimethyl-pyridine) extracts Tc(VII) quantitatively from alkaline supernatant wastes. The distribution coefficient (K/sub D/) for Tc(VII) going into the organic phase in the above system is 102 for a simulated West Valley waste and 191 for a simulated Savannah River Plant (SRP) waste.

  18. Ground movements associated with gas hydrate production

    SciTech Connect

    Siriwardane, H.J.

    1992-10-01

    The mechanics of ground movements during hydrate production can be more closely simulated by considering similarities with ground movements associated with subsidence in permafrost regions than with gob compaction in a longwall mine. The purpose of this research work is to investigate the potential strata movements associated with hydrate production by considering similarities with ground movements in permafrost regions. The work primarily involves numerical modeling of subsidence caused by hydrate dissociation. The investigation is based on the theories of continuum mechanics , thermomechanical behavior of frozen geo-materials, and principles of rock mechanics and geomechanics. It is expected that some phases of the investigation will involve the use of finite element method, which is a powerful computer-based method which has been widely used in many areas of science and engineering. Parametric studies will be performed to predict expected strata movements and surface subsidence for different reservoir conditions and properties of geological materials. The results from this investigation will be useful in predicting the magnitude of the subsidence problem associated with gas hydrate production. The analogy of subsidence in permafrost regions may provide lower bounds for subsidence expected in hydrate reservoirs. Furthermore, it is anticipated that the results will provide insight into planning of hydrate recovery operations.

  19. Beta decay of the fission product 125Sb and a new complete evaluation of absolute gamma ray transition intensities

    NASA Astrophysics Data System (ADS)

    Rajput, M. U.; Ali, N.; Hussain, S.; Mujahid, S. A.; MacMahon, D.

    2012-04-01

    The radionuclide 125Sb is a long-lived fission product, which decays to 125Te by negative beta emission with a half-life of 1008 day. The beta decay is followed by the emission of several gamma radiations, ranging from low to medium energy, that can suitably be used for high-resolution detector calibrations, decay heat calculations and in many other applications. In this work, the beta decay of 125Sb has been studied in detail. The complete published experimental data of relative gamma ray intensities in the beta decay of the radionuclide 125Sb has been compiled. The consistency analysis was performed and discrepancies found at several gamma ray energies. Evaluation of the discrepant data was carried out using Normalized Residual and RAJEVAL methods. The decay scheme balance was carried out using beta branching ratios, internal conversion coefficients, populating and depopulating gamma transitions to 125Te levels. The work has resulted in the consistent conversion factor equal to 29.59(13) %, and determined a new evaluated set of the absolute gamma ray emission probabilities. The work has also shown 22.99% of the delayed intensity fraction as outgoing from the 58 d isomeric 144 keV energy level and 77.01% of the prompt intensity fraction reaching to the ground state from the other excited states. The results are discussed and compared with previous evaluations. The present work includes additional experimental data sets which were not included in the previous evaluations. A new set of recommended relative and absolute gamma ray emission probabilities is presented.

  20. Microscopic description of complex nuclear decay: Multimodal fission

    NASA Astrophysics Data System (ADS)

    Staszczak, A.; Baran, A.; Dobaczewski, J.; Nazarewicz, W.

    2009-07-01

    Our understanding of nuclear fission, a fundamental nuclear decay, is still incomplete due to the complexity of the process. In this paper, we describe a study of spontaneous fission using the symmetry-unrestricted nuclear density functional theory. Our results show that the observed bimodal fission can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. We also predict a new phenomenon of trimodal spontaneous fission for some rutherfordium, seaborgium, and hassium isotopes.

  1. Microscopic description of complex nuclear decay: Multimodal fission

    SciTech Connect

    Staszczak, A.; Baran, A.; Dobaczewski, J.; Nazarewicz, W.

    2009-07-15

    Our understanding of nuclear fission, a fundamental nuclear decay, is still incomplete due to the complexity of the process. In this paper, we describe a study of spontaneous fission using the symmetry-unrestricted nuclear density functional theory. Our results show that the observed bimodal fission can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. We also predict a new phenomenon of trimodal spontaneous fission for some rutherfordium, seaborgium, and hassium isotopes.

  2. Gas Production Strategy of Underground Coal Gasification Based on Multiple Gas Sources

    PubMed Central

    Tianhong, Duan; Zuotang, Wang; Limin, Zhou; Dongdong, Li

    2014-01-01

    To lower stability requirement of gas production in UCG (underground coal gasification), create better space and opportunities of development for UCG, an emerging sunrise industry, in its initial stage, and reduce the emission of blast furnace gas, converter gas, and coke oven gas, this paper, for the first time, puts forward a new mode of utilization of multiple gas sources mainly including ground gasifier gas, UCG gas, blast furnace gas, converter gas, and coke oven gas and the new mode was demonstrated by field tests. According to the field tests, the existing power generation technology can fully adapt to situation of high hydrogen, low calorific value, and gas output fluctuation in the gas production in UCG in multiple-gas-sources power generation; there are large fluctuations and air can serve as a gasifying agent; the gas production of UCG in the mode of both power and methanol based on multiple gas sources has a strict requirement for stability. It was demonstrated by the field tests that the fluctuations in gas production in UCG can be well monitored through a quality control chart method. PMID:25114953

  3. Dynamical Aspects of Nuclear Fission

    NASA Astrophysics Data System (ADS)

    Kliman, J.; Itkis, M. G.; Gmuca, Š.

    2008-11-01

    Fission dynamics. Dependence of scission-neutron yield on light-fragment mass for [symbol]=1/2 [et al.]. Dynamics of capture quasifission and fusion-fission competition / L. Stuttgé ... [et al.] -- Fission-fission. The processes of fusion-fission and quasi-fission of superheavy nuclei / M. G. Itkis ... [et al.]. Fission and quasifission in the reactions [symbol]Ca+[symbol]Pb and [symbol]Ni+[symbol]W / G. N. Knyazheva ... [et al.]. Mass-energy characteristics of reactions [symbol]Fe+[symbol][symbol][symbol]266Hs and [symbol]Mg+[symbol]Cm[symbol][symbol]Hs at Coulomb barrier / L. Krupa ... [et al.]. Fusion of heavy ions at extreme sub-barrier energies / Ş. Mişicu and H. Esbensen. Fusion and fission dynamics of heavy nuclear system / V. Zagrebaev and W. Greiner. Time-dependent potential energy for fusion and fission processes / A. V. Karpov ... [et al.] -- Superheavy elements. Advances in the understanding of structure and production mechanisms for superheavy elements / W. Greiner and V. Zagrebaev. Fission barriers of heaviest nuclei / A. Sobiczewski ... [et al.]. Possibility of synthesizing doubly magic superheavy nuclei / Y Aritomo ... [et al.]. Synthesis of superheavy nuclei in [symbol]Ca-induced reactions / V. K. Utyonkov ... [et al.] -- Fragmentation. Production of neutron-rich nuclei in the nucleus-nucleus collisions around the Fermi energy / M. Veselský. Signals of enlarged core in [symbol]Al / Y. G. Ma ... [et al.] -- Exotic modes. New insight into the fission process from experiments with relativistic heavy-ion beams / K.-H. Schmidt ... [et al.]. New results for the intensity of bimodal fission in binary and ternary spontaneous fission of [symbol]Cf / C. Goodin ... [et al.]. Rare fission modes: study of multi-cluster decays of actinide nuclei / D. V. Kamanin ... [et al.]. Energy distribution of ternary [symbol]-particles in [symbol]Cf(sf) / M. Mutterer ... [et al.]. Preliminary results of experiment aimed at searching for collinear cluster tripartition of

  4. Behaviour of fission gas in the rim region of high burn-up UO 2 fuel pellets with particular reference to results from an XRF investigation

    NASA Astrophysics Data System (ADS)

    Mogensen, M.; Pearce, J. H.; Walker, C. T.

    1999-01-01

    XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8-54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU.

  5. Fission product separations testing

    SciTech Connect

    Bostick, D.A.; DePaoli, S.M.

    1997-10-01

    The initial goal of this task is to adequately understand the treatment needs of the end user in treating contaminated wastewater. These needs are then incorporated into the evaluation of new treatment technologies for wastewater treatment. Pertinent information is than supplied to the end user so that they can select a preferred process to meet their waste treatment needs. New sorbent materials, ion-exchange materials, or other processes of interest to DOE`s Office of Environmental Restoration (EM-40) will be evaluated initially for the removal of {sup 90}Sr and {sup 137}Cs from groundwater and process wastewater. Laboratory studies will strive to obtain a quantitative understanding of the behavior of these new materials and to evaluate their sorption efficiency in reference to a standard benchmark treatment technique. Testing of the new materials will begin by conducting scoping tests where new treatment materials are compared with standard, commercially available materials in batch shaker tests. Experimental data for the most promising sorbents will then be fit to an equilibrium model so that nuclide sorption can be predicted for variable wastewater composition. Additional testing with actual wastewater will be conducted with two or three of the most effective treatment methods. Once batch testing of a treatment method is completed, dynamic column tests will be performed to validate the equilibrium sorption model and to obtain the defining column operating parameters for scaling up the technology.

  6. Thermal reactor. [liquid silicon production from silane gas

    NASA Technical Reports Server (NTRS)

    Levin, H.; Ford, L. B. (Inventor)

    1982-01-01

    A thermal reactor apparatus and method of pyrolyticaly decomposing silane gas into liquid silicon product and hydrogen by-product gas is disclosed. The thermal reactor has a reaction chamber which is heated well above the decomposition temperature of silane. An injector probe introduces the silane gas tangentially into the reaction chamber to form a first, outer, forwardly moving vortex containing the liquid silicon product and a second, inner, rewardly moving vortex containing the by-product hydrogen gas. The liquid silicon in the first outer vortex deposits onto the interior walls of the reaction chamber to form an equilibrium skull layer which flows to the forward or bottom end of the reaction chamber where it is removed. The by-product hydrogen gas in the second inner vortex is removed from the top or rear of the reaction chamber by a vortex finder. The injector probe which introduces the silane gas into the reaction chamber is continually cooled by a cooling jacket.

  7. Advanced Space Fission Propulsion Systems

    NASA Technical Reports Server (NTRS)

    Houts, Michael G.; Borowski, Stanley K.

    2010-01-01

    -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.

  8. Production of biodiesel using expanded gas solvents

    SciTech Connect

    Ginosar, Daniel M; Fox, Robert V; Petkovic, Lucia M

    2009-04-07

    A method of producing an alkyl ester. The method comprises providing an alcohol and a triglyceride or fatty acid. An expanding gas is dissolved into the alcohol to form a gas expanded solvent. The alcohol is reacted with the triglyceride or fatty acid in a single phase to produce the alkyl ester. The expanding gas may be a nonpolar expanding gas, such as carbon dioxide, methane, ethane, propane, butane, pentane, ethylene, propylene, butylene, pentene, isomers thereof, and mixtures thereof, which is dissolved into the alcohol. The gas expanded solvent may be maintained at a temperature below, at, or above a critical temperature of the expanding gas and at a pressure below, at, or above a critical pressure of the expanding gas.

  9. Neutron production from (. cap alpha. ,n) reactions and spontaneous fission in ThO/sub 2/, UO/sub 2/, and (U,Pu)O/sub 2/ fuels

    SciTech Connect

    Perry, R.T.; Wilson, W.B.

    1981-06-01

    Available alpha-particle stopping cross-section and /sup 17/ /sup 18/O(..cap alpha..,n) cross-section data were adjusted, fitted, and used in calculating the thick-target neutron production function for alpha particles below 10 MeV in oxide fuels. The spent UO/sub 2/ function produced was folded with actinide decay spectra to determine (..cap alpha..,n) neutron production by each of 89 actinides. Spontaneous-fission (SF) neutron production for 40 actinides was calculated as the product of anti ..nu..(SF) and SF branching-fraction values accumulated or estimated from available data. These contributions and total neutron production in spent UO/sub 2/ fuel are tabulated and, when combined with any calculated inventory, describe the spent UO/sub 2/ neutron source. All data are tabulated and methodology is described to permit easy extension to specialized problems.

  10. GASCAP: Wellhead Gas Productive Capacity Model documentation, June 1993

    SciTech Connect

    Not Available

    1993-07-01

    The Wellhead Gas Productive Capacity Model (GASCAP) has been developed by EIA to provide a historical analysis of the monthly productive capacity of natural gas at the wellhead and a projection of monthly capacity for 2 years into the future. The impact of drilling, oil and gas price assumptions, and demand on gas productive capacity are examined. Both gas-well gas and oil-well gas are included. Oil-well gas productive capacity is estimated separately and then combined with the gas-well gas productive capacity. This documentation report provides a general overview of the GASCAP Model, describes the underlying data base, provides technical descriptions of the component models, diagrams the system and subsystem flow, describes the equations, and provides definitions and sources of all variables used in the system. This documentation report is provided to enable users of EIA projections generated by GASCAP to understand the underlying procedures used and to replicate the models and solutions. This report should be of particular interest to those in the Congress, Federal and State agencies, industry, and the academic community, who are concerned with the future availability of natural gas.

  11. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses: Criticality (keff) Predictions

    SciTech Connect

    Scaglione, John M.; Mueller, Don E.; Wagner, John C.

    2014-12-01

    One of the most important remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation—in particular, the availability and use of applicable measured data to support validation, especially for fission products (FPs). Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. In this study, this paper describes a validation approach for commercial spent nuclear fuel (SNF) criticality safety (keff) evaluations based on best-available data and methods and applies the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The criticality validation approach utilizes not only available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion program to support validation of the principal actinides but also calculated sensitivities, nuclear data uncertainties, and limited available FP LCE data to predict and verify individual biases for relevant minor actinides and FPs. The results demonstrate that (a) sufficient critical experiment data exist to adequately validate keff calculations via conventional validation approaches for the primary actinides, (b) sensitivity-based critical experiment selection is more appropriate for generating accurate application model bias and uncertainty, and (c) calculated sensitivities and nuclear data uncertainties can be used for generating conservative estimates of bias for minor actinides and FPs. Results based on the SCALE 6.1 and the ENDF/B-VII.0 cross-section libraries indicate that a conservative estimate of the bias for the minor actinides and FPs is 1.5% of their worth

  12. New results on the ternary fission of 243Cm

    NASA Astrophysics Data System (ADS)

    Heyse, J.; Wagemans, C.; Vermote, S.; Serot, O.; Geltenbort, P.; Soldner, T.; Van Gils, J.

    2005-11-01

    Ternary fission is an important source of He and tritium gas in nuclear reactors and used fuel elements. Therefore a systematic study of the ternary fission yields for 4He and tritons (t) is being performed. In recent years the influence of the excitation energy of the fissioning nucleus on the triton emission probability (t/B) has been investigated for different Cm and Cf isotopes. In this paper we report on new results on the neutron induced fission of 243Cm.

  13. Neutron Emission in Fission and Quasi-Fission

    NASA Astrophysics Data System (ADS)

    Itkis, I.; Bogatchev, A. A.; Chizhov, A. Yu.; Itkis, M. G.; Kliman, J.; Knyazheva, G. N.; Kondratiev, N. A.; Kozulin, E. M.; Korzyukov, I. V.; Krupa, L.; Oganessian, Yu. Ts.; Pokrovski, I. V.; Prokhorova, E. V.; Sagaidak, R. N.; Voskressenski, V. M.; Rusanov, A. Ya.; Corradi, L.; Stefanini, A. M.; Trotta, M.; Beghini, S.; Montagnoli, G.; Scarlassara, F.; Chubarian, G.; Hanappe, F.; Materna, T.; Dorvaux, O.; Rowley, N.; Stuttge, L.; Giardina, G.

    2005-09-01

    The work presents the results of the study of characteristics of the neutron emission in fission and quasi-fission of heavy and super-heavy nuclei, produced in the reactions with heavy ions. These experiments have been performed at the U-400 accelerator of the Flerov Laboratory of Nuclear Reactions (JINR), tandem accelerator in Legnaro (LNL) and VIVITRON accelerator in Strasbourg (IReS) with the use of the time-of-flight spectrometer of fission fragments CORSET and neutron multidetector DEMON. Mass-energy distributions (MED) of the 48Ca + 168Er, 208Pb, 238U and 18O + 208Pb reactions products at energies close to and below the Coulomb barrier have been studied. The pre- and post-fission neutron multiplicities as a function of the fragment mass have been obtained. A significant yield of the asymmetric component observed in the fragment mass distributions in the case of 18O + 208Pb reaction denotes the multimodal nature of the fission process. At the same time an increase in the yield of fragment masses ML ≅ 75-85 and MH ≅ 200-210 in the case 48Ca+208Pb, 238U reactions and ML ≅ 75-85 and MH ≅ 130-140 in the case 48Ca+168Er is rather connected with a quasi-fission process. The obtained neutron multiplicities dependences on fragment masses showed the validity of these assumptions.

  14. The role of off-line mass spectrometry in nuclear fission.

    PubMed

    De Laeter, J R

    1996-01-01

    The role of mass spectrometry in nuclear fission has been invaluable since 1940, when A. O. C. Nier separated microgram quantities of (235) U from (238) U, using a gas source mass spectrometer. This experiment enabled the fissionable nature of (235) U to be established. During the Manhattan Project, the mass spectrometer was used to measure the isotope abundances of uranium after processing in various separation systems, in monitoring the composition of the gaseous products in the Oak Ridge Diffusion Plant, and as a helium leak detector. Following the construction of the first reactor at the University of Chicago, it was necessary to unravel the nuclear systematics of the various fission products produced in the fission process. Off-line mass spectrometry was able to identify stable and long-lived isotopes produced in fission, but more importantly, was used in numerous studies of the distribution of mass of the cumulative fission yields. Improvements in sensitivity enabled off-line mass spectrometric studies to identify fine structure in the mass-yield curve and, hence, demonstrate the importance of shell structure in nuclear fission. Solid-source mass spectrometry was also able to measure the cumulative fission yields in the valley of symmetry in the mass-yield curve, and enabled spontaneous fission yields to be quantified. Apart from the accurate measurement of abundances, the stable isotope mass spectrometric technique has been invaluable in establishing absolute cumulative fission yields for many isotopes making up the mass-yield distribution curve for a variety of fissile nuclides. Extensive mass spectrometric studies of noble gases in primitive meteorites revealed the presence of fission products from the now extinct nuclide (244) Pu, and have eliminated the possibility of fission products from a super-heavy nuclide contributing to isotopic anomalies in meteoritic material. Numerous mass spectrometric studies of the isotopic and elemental abundances of

  15. Measurements of yields of fission products in the reaction of {sup 238}U with high-energy p, d and n beams

    SciTech Connect

    Nolen, J.A.; Ahmad, I.; Back, B.B.

    1995-08-01

    An experiment was performed at the Michigan State University cyclotron to determine the yields of neutron-rich fission products in the reaction of {sup 238}U with 100-MeV neutrons, 200-MeV deuterons and 200-MeV protons. Several 1-mm-thick {sup 238}U foils were irradiated for 100-second intervals sequentially for each configuration and the ten spectra were added for higher statistics. The three successive spectra, each for a 40 s period, were accumulated for each sample. Ten foils were irradiated. Successive spectra allowed us to determine approximate half-lives of the gamma peaks. Several arrangements, which were similar to the setup we plan to use in our radioactive beam proposal, were used for the production of fission products. For the high-energy neutron irradiation, U foils were placed after a 5-inch-long, 1-inch-diameter Be cylinder which stopped the 200-MeV deuteron beam generating 100-MeV neutrons. Arrangements for deuteron irradiation included direct irradiation of U foils, placing U foils after different lengths of (0.5 inch, 1.0 inch and 1.5 inch) 2-inch diameter U cylinder. Since the deuteron range in uranium is 17 mm, some of the irradiations were due to the secondary neutrons from the deuteron-induced fission of U. Similar arrangements were also used for the 200-MeV proton irradiation of the {sup 238}U foils. In all cases, several neutron-rich fission products were identified and their yields determined. In particular, we were able to observe Sn in all the runs and determine its yield. The data show that with our proposed radioactive device we will be able to produce more than 10{sup 12} {sup 132}Sn atoms per second in the target. Assuming an overall efficiency of 1 %, we will be able to deliver one particle nanoampere of {sup 132}Sn beam at a target location. Detailed analysis of the {gamma}-ray spectra is in progress.

  16. Fast-Mixed Spectrum Reactor progress report. Results of the FMSR Benchmark calculations and an assessment of current fission product libraries

    SciTech Connect

    Ludewig, H.; Durston, C.; Atefi, B.; Cerbone, R.J.

    1980-06-01

    As part of the Initial Feasibility Study of the Fast Mixed Spectrum Reactor, a series of benchmark calculations were made to determine the sensitivity of the physics analysis to differences in methods and data. Argonne National Laboratory (ANL), the Massachusetts Institute of Technology (MIT), and Oak Ridge National Laboratory (ORNL) were invited to participate with Brookhaven National Laboratory in the analysis of a FMSR model prescribed by BNL. Detailed comparisons are made including a comprehensive study on the adequacy of the fission product treatments.

  17. A suitability study of the fission product phantom and the bottle manikin absorption phantom for calibration of in vivo bioassay equipment for the DOELAP accreditation testing program

    SciTech Connect

    Olsen, P.C.; Lynch, T.P.

    1991-08-01

    Pacific Northwest laboratory (PNL) conducted an intercomparison study of the Fission Product phantom and the bottle manikin absorption (BOMAB) phantom for the US Department of Energy (DOE) to determine the consistency of calibration response of the two phantoms and their suitability for certification and use under a planned bioassay laboratory accreditation program. The study was initiated to determine calibration factors for both types of phantoms and to evaluate the suitability of their use in DOE Laboratory Accreditation Program (DOELAP) round-robin testing. The BOMAB was found to be more appropriate for the DOELAP testing program. 9 refs., 9 figs., 9 tabs.

  18. 78 FR 33051 - Notification of Proposed Production Activity, The Gas Company, LLC dba Hawai'i Gas, Subzone 9F...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-03

    ... Foreign-Trade Zones Board Notification of Proposed Production Activity, The Gas Company, LLC dba Hawai'i Gas, Subzone 9F (Synthetic Natural Gas), Kapolei, Hawaii The Gas Company, LLC dba Hawai'i Gas (Hawai'i Gas), operator of Subzone 9F, submitted a notification of proposed production activity to the...

  19. Mathematical analysis of intermittent gas injection model in oil production

    NASA Astrophysics Data System (ADS)

    Tasmi, Silvya, D. R.; Pudjo, S.; Leksono, M.; Edy, S.

    2016-02-01

    Intermittent gas injection is a method to help oil production process. Gas is injected through choke in surface and then gas into tubing. Gas forms three areas in tubing: gas column area, film area and slug area. Gas column is used to propel slug area until surface. A mathematical model of intermittent gas injection is developed in gas column area, film area and slug area. Model is expanding based on mass and momentum conservation. Using assume film thickness constant in tubing, model has been developed by Tasmi et. al. [14]. Model consists of 10 ordinary differential equations. In this paper, assumption of pressure in gas column is uniform. Model consist of 9 ordinary differential equations. Connection of several variables can be obtained from this model. Therefore, dynamics of all variables that affect to intermittent gas lift process can be seen from four equations. To study the behavior of variables can be analyzed numerically and mathematically. In this paper, simple mathematically analysis approach is used to study behavior of the variables. Variables that affect to intermittent gas injection are pressure in upstream valve and in gas column. Pressure in upstream valve will decrease when gas mass in valve greater than gas mass in choke. Dynamic of the pressure in the gas column will decrease and increase depending on pressure in upstream valve.

  20. Measuring micro-organism gas production

    NASA Technical Reports Server (NTRS)

    Wilkins, J. R.; Pearson, A. O.; Mills, S. M.

    1973-01-01

    Transducer, which senses pressure buildup, is easy to assemble and use, and rate of gas produced can be measured automatically and accurately. Method can be used in research, in clinical laboratories, and for environmental pollution studies because of its ability to detect and quantify rapidly the number of gas-producing microorganisms in water, beverages, and clinical samples.

  1. Analysis of eastern Devonian gas shales production data

    SciTech Connect

    Gatens, J.M.; Stanley, D.K.; Lancaster, D.E.; Lee, W.J.; Lane, H.S.; Watson, A.T.

    1989-05-01

    Production data from more than 800 Devonian shale wells have been analyzed. Permeability-thickness product and gas in place estimated from production data have been found to correlate with well performance. Empirical performance equations, production type curves, and an analytical dual-porosity model with automatic history-matching scheme were developed for the Devonian shale.

  2. Depressurization and electrical heating of hydrate sediment for gas production

    NASA Astrophysics Data System (ADS)

    Minagawa, H.

    2015-12-01

    As a part of a Japanese National hydrate research program (MH21, funded by METI), we performed a study on electrical heating of the hydrate core combined with depressurization for gas production. In-situ dissociation of natural gas hydrate is necessary for commercial recovery of natural gas from natural gas hydrate sediment. Thermal stimulation is an effective dissociation method, along with depressurization.To simulate methane gas production from methane hydrate layer, we investigated electrical heating of methane hydrate sediment. A decrease in core temperature due to the endothermic reaction of methane hydrate dissociation was suppressed and the core temperature increased between 1oC and 4oC above the control temperature with electric heating. A current density of 10A/m2 with depressurization would effectively dissociate hydrate. Therefore, depressurization and additional electrode heating of hydrate sediment saturated with electrolyte solution was confirmed to enable higher gas production from sediment with less electric power.

  3. First-principles study of fission product (Xe, Cs, Sr) incorporation and segregation in alkaline earth metal oxides, HfO2, and MgO-HfO2 interface

    SciTech Connect

    Liu, Xiang-yang; Uberuaga, Blas P; Sickafus, Kurt E

    2008-01-01

    In order to close the nuclear fuel cycle, advanced concepts for separating out fission products are necessary. One approach is to use a dispersion fuel form in which a fissile core is surrounded by an inert matrix that captures and immobilizes the fission products from the core. If this inert matrix can be easily separated from the fuel, via e.g. solution chemistry, the fission products can be separated from the fissile material. We examine a surrogate dispersion fuel composition, in which hafnia (HfO{sub 2}) is a surrogate for the fissile core and alkaline earth metal oxides are used as the inert matrix. The questions of fission product incorporation in these oxides and possible segregation behavior at interfaces are considered. Density functional theory based calculations for fission product elements (Xe, Sr, and Cs) in these oxides are carried out. We find smaller incorporation energy in hafnia than in MgO for Cs and Sr, and Xe if variation of charge state is allowed. We also find that this trend is reversed or reduced for alkaline earth metal oxides with large cation sizes. Model interfacial calculations show a strong tendency of segregation from bulk MgO to MgO-HfO{sub 2} interfaces.

  4. Integrated production of fuel gas and oxygenated organic compounds from synthesis gas

    DOEpatents

    Moore, Robert B.; Hegarty, William P.; Studer, David W.; Tirados, Edward J.

    1995-01-01

    An oxygenated organic liquid product and a fuel gas are produced from a portion of synthesis gas comprising hydrogen, carbon monoxide, carbon dioxide, and sulfur-containing compounds in a integrated feed treatment and catalytic reaction system. To prevent catalyst poisoning, the sulfur-containing compounds in the reactor feed are absorbed in a liquid comprising the reactor product, and the resulting sulfur-containing liquid is regenerated by stripping with untreated synthesis gas from the reactor. Stripping offgas is combined with the remaining synthesis gas to provide a fuel gas product. A portion of the regenerated liquid is used as makeup to the absorber and the remainder is withdrawn as a liquid product. The method is particularly useful for integration with a combined cycle coal gasification system utilizing a gas turbine for electric power generation.

  5. Method for the production of synthesis gas

    SciTech Connect

    Deuser, N.; Diemer, P.

    1982-09-28

    Crude gas obtained through partial oxidation of carboncontaining material is subjected to catalytic conversion in an apparatus containing a moistening-demoistering system connected by means of a combined water circulation. To reduce the required amount of steam and process heating, a partial stream of water coming from the moistener is branched off, heated by indirect heat exchange with converted, hot synthesis gas, and then united with the partial stream of water coming from the demoistener for re-delivery to the moistener.

  6. The seventh international conference on the chemistry and migration behavior of actinides and fission products in the Geosphere MIGRATION'99 abstracts

    SciTech Connect

    Palmer, C

    1999-09-01

    The Migration conferences focus on recent developments in the fundamental chemistry of actinides and fission products in natural aquifer systems, their interactions and migration in the geosphere, and the processes involved in modeling their geochemical behavior. The primary mode dissemination of technical information will be early evening poster sessions designed to encourage intensive communication between the authors and participants. Daily oral sessions will be opened with invited lectures followed by contributed papers within the scope of each session. Sessions cover: (A) Chemistry of actinides and fission products in natural aquatic systems: (1) Solubilities and dissolution reactions; (2) Complexation with inorganic and organic ligands; (3) Redox reactions; (4) Colloid formation; and (5) Experimental methods. (B) Geochemical interactions and transport phenomena: (1) Diffusion and migration in geologic media; (2) Sorption/desorption phenomena; (3) Natural analog studies; (4) Effects of biological activities and organic materials; (5) Colloid transport; (6) Radionuclides in soils; and (7) Soil-remediation chemistries. (C) Data base development and modeling: (1) Data selection and evaluation; (2) Data base management; (3) Geochemical models and modeling; (4) Application of models; and (5) Validation of modeling results.

  7. 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.

  8. 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. PMID:25537991

  9. Gas hearth products market fact base. Topical report, January 1996

    SciTech Connect

    1996-02-01

    The Gas Hearth Products Market Fact Base is an analysis of the U.S. gas log and fireplace markets. The study was undertaken to: determine current usage of and attitudes about fireplaces; identify barriers to acceptance of gas logs and fireplaces; determine the influence of service providers, and; identify important trends that can affect the markets for gas hearth products. The market fact base is based on four studies: a market analysis synthesizing primary and secondary research reports; in-depth interviews with market influencers from across the country (architects, contractors, interior designers, fireplace retailers and installers) and industry experts from gas utilities and trade associations; focus group meetings with consumers who own or intend to buy fireplaces, gas fireplace industry professionals, and editors of fireplace-related trade magazines, and; quantitative interviews with consumers in six U.S. cities.

  10. Process for production desulfurized of synthesis gas

    DOEpatents

    Wolfenbarger, James K.; Najjar, Mitri S.

    1993-01-01

    A process for the partial oxidation of a sulfur- and silicate-containing carbonaceous fuel to produce a synthesis gas with reduced sulfur content which comprises partially oxidizing said fuel at a temperature in the range of 1900.degree.-2600.degree. F. in the presence of a temperature moderator, an oxygen-containing gas and a sulfur capture additive which comprises a calcium-containing compound portion, a sodium-containing compound portion, and a fluoride-containing compound portion to produce a synthesis gas comprising H.sub.2 and CO with a reduced sulfur content and a molten slag which comprises (1) a sulfur-containing sodium-calcium-fluoride silicate phase; and (2) a sodium-calcium sulfide phase.

  11. Strategies for gas production from oceanic Class 3 hydrateaccumulations

    SciTech Connect

    Moridis, George J.; Reagan, Matthew T.

    2007-05-01

    Gas hydrates are solid crystalline compounds in which gasmolecules are lodged within the lattices of ice crystals. Vast amounts ofCH4 are trapped in gas hydrates, and a significant effort has recentlybegun to evaluate hydrate deposits as a potential energy source. Class 3hydrate deposits are characterized by an isolated Hydrate-Bearing Layer(HBL) that is not in contact with any hydrate-free zone of mobile fluids.The base of the HBL in Class 3 deposits may occur within or at the edgeof the zone of thermodynamic hydrate stability.In this numerical study oflong-term gas production from typical representatives of unfracturedClass 3 deposits, we determine that simple thermal stimulation appears tobe a slow and inefficient production method. Electrical heating and warmwater injection result in very low production rates (4 and 12 MSCFD,respectively) that are orders of magnitude lower than generallyacceptable standards of commercial viability of gas production fromoceanic reservoirs. However, production from depressurization-baseddissociation based on a constant well pressure appears to be a promisingapproach even in deposits characterized by high hydrate saturations. Thisapproach allows the production of very large volumes ofhydrate-originating gas at high rates (>15 MMSCFD, with a long-termaverage of about 8.1 MMSCFD for the reference case) for long times usingconventional technology. Gas production from hydrates is accompanied by asignificant production of water. However, unlike conventional gasreservoirs, the water production rate declines with time. The lowsalinity of the produced water may require care in its disposal. Becauseof the overwhelming advantage of depressurization-based methods, thesensitivity analysis was not extendedto thermal stimulation methods. Thesimulation results indicate that depressurization-induced gas productionfrom oceanic Class 3 deposits increases (and the corresponding waterto-gas ratio decreases) with increasing hydrate temperature

  12. Determination of the rod-wise fission gas release fraction in a complete fuel assembly using non-destructive gamma emission tomography

    NASA Astrophysics Data System (ADS)

    Holcombe, Scott; Andersson, Peter; Svärd, Staffan Jacobsson; Hallstadius, Lars

    2016-11-01

    A gamma tomography instrument has been developed at the Halden Boiling Water Reactor (HBWR) in cooperation between the Institute for Energy Technology, Westinghouse (Sweden) and Uppsala University. The instrument is used to record the gamma radiation field surrounding complete fuel assemblies and consists of a shielded enclosure with fixtures to accurately position the fuel and detector relative to each other. A High Purity Germanium detector is used for acquiring high-resolution spectroscopic data, allowing for analysis of multiple gamma-ray peaks. Using the data extracted from the selected peaks, tomographic reconstruction algorithms are used to reproduce the corresponding spatial gamma-ray source distributions within the fuel assembly. With this method, rod-wise data can be can be deduced without the need to dismantle the fuel. In this work, the tomographic device has been experimentally benchmarked for non-destructive rod-wise determination of the Fission Gas Release (FGR) fraction. Measurements were performed on the fuel-stack and gas-plenum regions of a complete fuel assembly, and quantitative tomographic reconstructions of the measurement data were performed in order to determine the rod-wise ratio of 85Kr in the gas plenum to 137Cs in the fuel stack. The rod-wise ratio of 85Kr/137Cs was, in turn, used to calculate the rod-wise FGR fraction. In connection to the tomographic measurements, the fuel rods were also measured individually using gamma scanning in order to provide an experimental benchmark for the tomographic method. Fuel rods from two donor driver fuel assemblies were placed into a nine-rod HBWR driver fuel assembly configuration. In order to provide a challenging measurement object and thus an appropriate benchmark for the tomographic method, five rods were taken from an assembly with a burnup of 51 MWd/kgUO2, and four rods were from an assembly with a burnup of 26 MWd/kgUO2. At the time of the measurements, the nine rods had cooled for

  13. Tempest gas turbine extends EGT product line

    SciTech Connect

    Chellini, R.

    1995-07-01

    With the introduction of the 7.8 MW (mechanical output) Tempest gas turbine, ECT has extended the company`s line of its small industrial turbines. The new Tempest machine, featuring a 7.5 MW electric output and a 33% thermal efficiency, ranks above the company`s single-shaft Typhoon gas turbine, rated 3.2 and 4.9 MW, and the 6.3 MW Tornado gas turbine. All three machines are well-suited for use in combined heat and power (CHP) plants, as demonstrated by the fact that close to 50% of the 150 Typhoon units sold are for CHP applications. This experience has induced EGT, of Lincoln, England, to announce the introduction of the new gas turbine prior to completion of the testing program. The present single-shaft machine is expected to be used mainly for industrial trial cogeneration. This market segment, covering the needs of paper mills, hospitals, chemical plants, ceramic industry, etc., is a typical local market. Cogeneration plants are engineered according to local needs and have to be assisted by local organizations. For this reason, to efficiently cover the world market, EGT has selected a number of associates that will receive from Lincoln completely engineered machine packages and will engineer the cogeneration system according to custom requirements. These partners will also assist the customer and dispose locally of the spares required for maintenance operations.

  14. Outer Continental Shelf Oil and Gas Leasing/Production Program

    SciTech Connect

    Not Available

    1988-01-01

    This annual report on the Outer Continental Shelf (OCS) Oil and Gas Leasing and Production program summarizes receipts and expenditures, and includes information on OCS safety violations as reported by the US Coast Guard. 3 figs., 12 tabs.

  15. Preliminary report on the commercial viability of gas production from natural gas hydrates

    USGS Publications Warehouse

    Walsh, M.R.; Hancock, S.H.; Wilson, S.J.; Patil, S.L.; Moridis, G.J.; Boswell, R.; Collett, T.S.; Koh, C.A.; Sloan, E.D.

    2009-01-01

    Economic studies on simulated gas hydrate reservoirs have been compiled to estimate the price of natural gas that may lead to economically viable production from the most promising gas hydrate accumulations. As a first estimate, $CDN2005 12/Mscf is the lowest gas price that would allow economically viable production from gas hydrates in the absence of associated free gas, while an underlying gas deposit will reduce the viability price estimate to $CDN2005 7.50/Mscf. Results from a recent analysis of the simulated production of natural gas from marine hydrate deposits are also considered in this report; on an IROR basis, it is $US2008 3.50-4.00/Mscf more expensive to produce marine hydrates than conventional marine gas assuming the existence of sufficiently large marine hydrate accumulations. While these prices represent the best available estimates, the economic evaluation of a specific project is highly dependent on the producibility of the target zone, the amount of gas in place, the associated geologic and depositional environment, existing pipeline infrastructure, and local tariffs and taxes. ?? 2009 Elsevier B.V.

  16. Energy Dependence of Fission Product Yields from 235U, 238U and 239Pu for Incident Neutron Energies Between 0.5 and 14.8 MeV

    NASA Astrophysics Data System (ADS)

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

    2016-01-01

    Fission Product Yields (FPY) have historically been one of the most observable features of the fission process. They are known to have strong variations that are dependent on the fissioning species, the excitation energy, and the angular momentum of the compound system. However, consistent and systematic studies of the variation of these FPY with energy have proved challenging. This is caused primarily by the nature of the experiments that have traditionally relied on radiochemical procedures to isolate specific fission products. Although radiochemical procedures exist that can isolate all products, each element presents specific challenges and introduces varying degrees of systematic errors that can make inter-comparison of FPY uncertain. Although of high importance in fields such as nuclear forensics and Stockpile Stewardship, accurate information about the energy dependence of neutron induced FPY are sparse, due primarily to the lack of suitable monoenergetic neutron sources. There is a clear need for improved data, and to address this issue, a collaboration was formed between Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL) and the Triangle Universities Nuclear Laboratory (TUNL) to measure the energy dependence of FPY for 235U, 238U and 239Pu. The measurements have been performed at TUNL, using a 10 MV Tandem Van de Graaff accelerator to produce monoenergetic neutrons at energies between 0.6 MeV to 14.8 MeV through a variety of reactions. The measurements have utilized a dual-fission chamber, with thin (10-100 μg/cm2) reference foils of similar material to a thick (100-400 mg) activation target held in the center between the chambers. This method allows for the accurate determination of the number of fissions that occurred in the thick target without requiring knowledge of the fission cross section or neutron fluence on target. Following activation, the thick target was removed from the dual-fission chamber and gamma

  17. Energy dependence of fission product yields from 235U, 238U and 239Pu for incident neutron energies between 0.5 and 14.8 MeV

    DOE PAGES

    Gooden, M. E.; Arnold, C. W.; Becker, J. A.; Bhatia, C.; Bhike, M.; Bond, E. M.; Bredeweg, T. A.; Fallin, B.; Fowler, M. M.; Howell, C. R.; et al

    2016-01-06

    In this study, Fission Product Yields (FPY) have historically been one of the most observable features of the fission process. They are known to have strong variations that are dependent on the fissioning species, the excitation energy, and the angular momentum of the compound system. However, consistent and systematic studies of the variation of these FPY with energy have proved challenging. This is caused primarily by the nature of the experiments that have traditionally relied on radiochemical procedures to isolate specific fission products. Although radiochemical procedures exist that can isolate all products, each element presents specific challenges and introduces varyingmore » degrees of systematic errors that can make inter-comparison of FPY uncertain. Although of high importance in fields such as nuclear forensics and Stockpile Stewardship, accurate information about the energy dependence of neutron induced FPY are sparse, due primarily to the lack of suitable monoenergetic neutron sources. There is a clear need for improved data, and to address this issue, a collaboration was formed between Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL) and the Triangle Universities Nuclear Laboratory (TUNL) to measure the energy dependence of FPY for 235U, 238U and 239Pu. The measurements have been performed at TUNL, using a 10 MV Tandem Van de Graaff accelerator to produce monoenergetic neutrons at energies between 0.6 MeV to 14.8 MeV through a variety of reactions. The measurements have utilized a dual-fission chamber, with thin (10-100 μg/cm2) reference foils of similar material to a thick (100-400 mg) activation target held in the center between the chambers. This method allows for the accurate determination of the number of fissions that occurred in the thick target without requiring knowledge of the fission cross section or neutron fluence on target. Following activation, the thick target was removed from the dual-fission

  18. Carbonyl Emissions From Oil and Gas Production Facilities

    NASA Astrophysics Data System (ADS)

    Lyman, S. N.; O'Neil, T.; Tran, T.

    2015-12-01

    A number of recent studies have targeted emissions of methane and other hydrocarbons from oil and gas exploration and production activity. These measurements are greatly increasing understanding of the atmospheric impacts of oil and gas development. Very few measurements exist, however, of emissions of formaldehyde and other carbonyls from oil and gas equipment. Carbonyls are toxic and serve as important ozone precursors, especially during winter ozone episodes in places like Utah's Uintah Basin. Current air quality models are only able to reproduce observed high wintertime ozone if they incorporate emissions inventories with very high carbonyl emissions. We measured carbonyl emissions from oil and gas equipment and facilities—including glycol dehydrators, liquid storage tanks, raw gas leaks, raw gas-burning engines, and produced water surface impoundments—in Rocky Mountain oil and gas fields. Carbonyl emissions from raw gas were below detection, but emissions of formaldehyde, acetaldehyde, and other carbonyls were detected from liquid storage tanks, glycol dehydrators, and other oil and gas equipment. In some cases, carbonyls may be formed from the degradation of methanol and other chemicals used in oil and gas production, but the collected data provide evidence for other non-combustion formation pathways. Raw gas-burning engines also emitted carbonyls. Emissions from all measured sources were a small fraction of total volatile organic compound emissions. We incorporated our measurements into an emissions inventory, used that inventory in an air quality model (WRF-SMOKE-CAMx), and were unable to reproduce observed high wintertime ozone. This could be because (1) emission sources we have not yet measured, including compressors, gas processing plants, and others, are large; (2) non-carbonyl emissions, especially those that quickly degrade into carbonyls during photochemical processing, are underestimated in the inventory; or (3) the air quality model is unable

  19. Spontaneous and induced emission of XeCl* excimer molecules under pumping of Xe – CCl{sub 4} and Ar – Xe – CCl{sub 4} gas mixtures with a low CCl{sub 4} content by fast electrons and uranium fission fragments

    SciTech Connect

    Mis'kevich, A I; Guo, J; Dyuzhov, Yu A

    2013-11-30

    The spontaneous and induced emission of XeCl* excimer molecules upon excitation of Xe – CCl{sub 4} and Ar – Xe – CCl{sub 4} gas mixtures with a low CCl{sub 4} content by high-energy charged particles [a pulsed high-energy electron beam and products of neutron nuclear reaction {sup 235}U(n, f)] has been experimentally studied. The electron energy was 150 keV, and the pump current pulse duration and amplitude were 5 ns and 5 A, respectively. The energy of fission fragments did not exceed 100 MeV, the duration of the neutron pump pulse was 200 μs, and the specific power contribution to the gas was about 300 W cm{sup -3}. Electron beam pumping in a cell 4 cm long with a cavity having an output mirror transmittance of 2.7% gives rise to lasing on the B → X transition in the XeCl* molecule (λ = 308 nm) with a gain α = 0.0085 cm{sup -1} and fluorescence efficiency η ≈ 10%. Pumping by fission fragments in a 250-cm-long cell with a cavity formed by a highly reflecting mirror and a quartz window implements amplified spontaneous emission (ASE) with an output power of 40 – 50 kW sr{sup -1} and a base ASE pulse duration of ∼200 ms. .

  20. Challenges, uncertainties and issues facing gas production from gas hydrate deposits

    SciTech Connect

    Moridis, G.J.; Collett, T.S.; Pooladi-Darvish, M.; Hancock, S.; Santamarina, C.; Boswell, R.; Kneafsey, T.; Rutqvist, J.; Kowalsky, M.; Reagan, M.T.; Sloan, E.D.; Sum, A.K.; Koh, C.

    2010-11-01

    The current paper complements the Moridis et al. (2009) review of the status of the effort toward commercial gas production from hydrates. We aim to describe the concept of the gas hydrate petroleum system, to discuss advances, requirement and suggested practices in gas hydrate (GH) prospecting and GH deposit characterization, and to review the associated technical, economic and environmental challenges and uncertainties, including: the accurate assessment of producible fractions of the GH resource, the development of methodologies for identifying suitable production targets, the sampling of hydrate-bearing sediments and sample analysis, the analysis and interpretation of geophysical surveys of GH reservoirs, well testing methods and interpretation of the results, geomechanical and reservoir/well stability concerns, well design, operation and installation, field operations and extending production beyond sand-dominated GH reservoirs, monitoring production and geomechanical stability, laboratory investigations, fundamental knowledge of hydrate behavior, the economics of commercial gas production from hydrates, and the associated environmental concerns.

  1. Geometry of membrane fission.

    PubMed

    Frolov, Vadim A; Escalada, Artur; Akimov, Sergey A; Shnyrova, Anna V

    2015-01-01

    Cellular membranes define the functional geometry of intracellular space. Formation of new membrane compartments and maintenance of complex organelles require division and disconnection of cellular membranes, a process termed membrane fission. Peripheral membrane proteins generally control membrane remodeling during fission. Local membrane stresses, reflecting molecular geometry of membrane-interacting parts of these proteins, sum up to produce the key membrane geometries of fission: the saddle-shaped neck and hour-glass hemifission intermediate. Here, we review the fundamental principles behind the translation of molecular geometry into membrane shape and topology during fission. We emphasize the central role the membrane insertion of specialized protein domains plays in orchestrating fission in vitro and in cells. We further compare individual to synergistic action of the membrane insertion during fission mediated by individual protein species, proteins complexes or membrane domains. Finally, we describe how local geometry of fission intermediates defines the functional design of the protein complexes catalyzing fission of cellular membranes. PMID:25062896

  2. Thorium-uranium fission radiography

    NASA Technical Reports Server (NTRS)

    Haines, E. L.; Weiss, J. R.; Burnett, D. S.; Woolum, D. S.

    1976-01-01

    Results are described for studies designed to develop routine methods for in-situ measurement of the abundance of Th and U on a microscale in heterogeneous samples, especially rocks, using the secondary high-energy neutron flux developed when the 650 MeV proton beam of an accelerator is stopped in a 42 x 42 cm diam Cu cylinder. Irradiations were performed at three different locations in a rabbit tube in the beam stop area, and thick metal foils of Bi, Th, and natural U as well as polished silicate glasses of known U and Th contents were used as targets and were placed in contact with mica which served as a fission track detector. In many cases both bare and Cd-covered detectors were exposed. The exposed mica samples were etched in 48% HF and the fission tracks counted by conventional transmitted light microscopy. Relative fission cross sections are examined, along with absolute Th track production rates, interaction tracks, and a comparison of measured and calculated fission rates. The practicality of fast neutron radiography revealed by experiments to data is discussed primarily for Th/U measurements, and mixtures of other fissionable nuclei are briefly considered.

  3. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    SciTech Connect

    Shamasundar, B.I.; Fehrenbach, M.E.

    1981-05-01

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations.

  4. Devices for the Production of Reference Gas Mixtures.

    PubMed

    Fijało, Cyprian; Dymerski, Tomasz; Gębicki, Jacek; Namieśnik, Jacek

    2016-09-01

    For many years there has been growing demand for gaseous reference materials, which is connected with development in many fields of science and technology. As a result, new methodological and instrumental solutions appear that can be used for this purpose. Appropriate quality assurance/quality control (QA/QC) must be used to make sure that measurement data are a reliable source of information. Reference materials are a significant element of such systems. In the case of gas samples, such materials are generally called reference gas mixtures. This article presents the application and classification of reference gas mixtures, which are a specific type of reference materials, and the methods for obtaining them are described. Construction solutions of devices for the production of reference gas mixtures are detailed, and a description of a prototype device for dynamic production of reference gas mixtures containing aroma compounds is presented. PMID:27437588

  5. Devices for the Production of Reference Gas Mixtures.

    PubMed

    Fijało, Cyprian; Dymerski, Tomasz; Gębicki, Jacek; Namieśnik, Jacek

    2016-09-01

    For many years there has been growing demand for gaseous reference materials, which is connected with development in many fields of science and technology. As a result, new methodological and instrumental solutions appear that can be used for this purpose. Appropriate quality assurance/quality control (QA/QC) must be used to make sure that measurement data are a reliable source of information. Reference materials are a significant element of such systems. In the case of gas samples, such materials are generally called reference gas mixtures. This article presents the application and classification of reference gas mixtures, which are a specific type of reference materials, and the methods for obtaining them are described. Construction solutions of devices for the production of reference gas mixtures are detailed, and a description of a prototype device for dynamic production of reference gas mixtures containing aroma compounds is presented.

  6. NOVEL REACTOR FOR THE PRODUCTION OF SYNTHESIS GAS

    SciTech Connect

    Vasilis Papavassiliou; Leo Bonnell; Dion Vlachos

    2004-12-01

    Praxair investigated an advanced technology for producing synthesis gas from natural gas and oxygen This production process combined the use of a short-reaction time catalyst with Praxair's gas mixing technology to provide a novel reactor system. The program achieved all of the milestones contained in the development plan for Phase I. We were able to develop a reactor configuration that was able to operate at high pressures (up to 19atm). This new reactor technology was used as the basis for a new process for the conversion of natural gas to liquid products (Gas to Liquids or GTL). Economic analysis indicated that the new process could provide a 8-10% cost advantage over conventional technology. The economic prediction although favorable was not encouraging enough for a high risk program like this. Praxair decided to terminate development.

  7. Evaluation of the gas production economics of the gas hydrate cyclic thermal injection model

    SciTech Connect

    Kuuskraa, V.A.; Hammersheimb, E.; Sawyer, W.

    1985-05-01

    The objective of the work performed under this directive is to assess whether gas hydrates could potentially be technically and economically recoverable. The technical potential and economics of recovering gas from a representative hydrate reservoir will be established using the cyclic thermal injection model, HYDMOD, appropriately modified for this effort, integrated with economics model for gas production on the North Slope of Alaska, and in the deep offshore Atlantic. The results from this effort are presented in this document. In Section 1, the engineering cost and financial analysis model used in performing the economic analysis of gas production from hydrates -- the Hydrates Gas Economics Model (HGEM) -- is described. Section 2 contains a users guide for HGEM. In Section 3, a preliminary economic assessment of the gas production economics of the gas hydrate cyclic thermal injection model is presented. Section 4 contains a summary critique of existing hydrate gas recovery models. Finally, Section 5 summarizes the model modification made to HYDMOD, the cyclic thermal injection model for hydrate gas recovery, in order to perform this analysis.

  8. Fission Yield Measurements by Inductively Coupled Plasma Mass-Spectrometry

    SciTech Connect

    Irina Glagolenko; Bruce Hilton; Jeffrey Giglio; Daniel Cummings; Karl Grimm; Richard McKnight

    2009-11-01

    Correct prediction of the fission products inventory in irradiated nuclear fuels is essential for accurate estimation of fuel burnup, establishing proper requirements for spent fuel transportation and storage, materials accountability and nuclear forensics. Such prediction is impossible without accurate knowledge of neutron induced fission yields. Unfortunately, the accuracy of the fission yields reported in the ENDF/B-VII.0 library is not uniform across all of the data and much of the improvement is desired for certain isotopes and fission products. We discuss our measurements of cumulative fission yields in nuclear fuels irradiated in thermal and fast reactor spectra using Inductively Coupled Plasma Mass Spectrometry.

  9. Compact fission counter for DANCE

    SciTech Connect

    Wu, C Y; Chyzh, A; Kwan, E; Henderson, R; Gostic, J; Carter, D; Bredeweg, T; Couture, A; Jandel, M; Ullmann, J

    2010-11-06

    The Detector for Advanced Neutron Capture Experiments (DANCE) consists of 160 BF{sub 2} crystals with equal solid-angle coverage. DANCE is a 4{pi} {gamma}-ray calorimeter and designed to study the neutron-capture reactions on small quantities of radioactive and rare stable nuclei. These reactions are important for the radiochemistry applications and modeling the element production in stars. The recognition of capture event is made by the summed {gamma}-ray energy which is equivalent of the reaction Q-value and unique for a given capture reaction. For a selective group of actinides, where the neutron-induced fission reaction competes favorably with the neutron capture reaction, additional signature is needed to distinguish between fission and capture {gamma} rays for the DANCE measurement. This can be accomplished by introducing a detector system to tag fission fragments and thus establish a unique signature for the fission event. Once this system is implemented, one has the opportunity to study not only the capture but also fission reactions. A parallel-plate avalanche counter (PPAC) has many advantages for the detection of heavy charged particles such as fission fragments. These include fast timing, resistance to radiation damage, and tolerance of high counting rate. A PPAC also can be tuned to be insensitive to {alpha} particles, which is important for experiments with {alpha}-emitting actinides. Therefore, a PPAC is an ideal detector for experiments requiring a fast and clean trigger for fission. A PPAC with an ingenious design was fabricated in 2006 by integrating amplifiers into the target assembly. However, this counter was proved to be unsuitable for this application because of issues related to the stability of amplifiers and the ability to separate fission fragments from {alpha}'s. Therefore, a new design is needed. A LLNL proposal to develop a new PPAC for DANCE was funded by NA22 in FY09. The design goal is to minimize the mass for the proposed counter

  10. US production of natural gas from tight reservoirs

    SciTech Connect

    Not Available

    1993-10-18

    For the purposes of this report, tight gas reservoirs are defined as those that meet the Federal Energy Regulatory Commission`s (FERC) definition of tight. They are generally characterized by an average reservoir rock permeability to gas of 0.1 millidarcy or less and, absent artificial stimulation of production, by production rates that do not exceed 5 barrels of oil per day and certain specified daily volumes of gas which increase with the depth of the reservoir. All of the statistics presented in this report pertain to wells that have been classified, from 1978 through 1991, as tight according to the FERC; i.e., they are ``legally tight`` reservoirs. Additional production from ``geologically tight`` reservoirs that have not been classified tight according to the FERC rules has been excluded. This category includes all producing wells drilled into legally designated tight gas reservoirs prior to 1978 and all producing wells drilled into physically tight gas reservoirs that have not been designated legally tight. Therefore, all gas production referenced herein is eligible for the Section 29 tax credit. Although the qualification period for the credit expired at the end of 1992, wells that were spudded (began to be drilled) between 1978 and May 1988, and from November 5, 1990, through year end 1992, are eligible for the tax credit for a subsequent period of 10 years. This report updates the EIA`s tight gas production information through 1991 and considers further the history and effect on tight gas production of the Federal Government`s regulatory and tax policy actions. It also provides some high points of the geologic background needed to understand the nature and location of low-permeability reservoirs.

  11. Effects of gas composition in headspace and bicarbonate concentrations in media on gas and methane production, degradability, and rumen fermentation using in vitro gas production techniques.

    PubMed

    Patra, Amlan Kumar; Yu, Zhongtang

    2013-07-01

    Headspace gas composition and bicarbonate concentrations in media can affect methane production and other characteristics of rumen fermentation in in vitro gas production systems, but these 2 important factors have not been evaluated systematically. In this study, these 2 factors were investigated with respect to gas and methane production, in vitro digestibility of feed substrate, and volatile fatty acid (VFA) profile using in vitro gas production techniques. Three headspace gas compositions (N2+ CO2+ H2 in the ratio of 90:5:5, CO2, and N2) with 2 substrate types (alfalfa hay only, and alfalfa hay and a concentrate mixture in a 50:50 ratio) in a 3×2 factorial design (experiment 1) and 3 headspace compositions (N2, N2 + CO2 in a 50:50 ratio, and CO2) with 3 bicarbonate concentrations (80, 100, and 120 mM) in a 3×3 factorial design (experiment 2) were evaluated. In experiment 1, total gas production (TGP) and net gas production (NGP) was the lowest for CO2, followed by N2, and then the gas mixture. Methane concentration in headspace gas after fermentation was greater for CO2 than for N2 and the gas mixture, whereas total methane production (TMP) and net methane production (NMP) were the greatest for CO2, followed by the gas mixture, and then N2. Headspace composition did not affect in vitro digestibility or the VFA profile, except molar percentages of propionate, which were greater for CO2 and N2 than for the gas mixture. Methane concentration in headspace gas, TGP, and NGP were affected by the interaction of headspace gas composition and substrate type. In experiment 2, increasing concentrations of CO2 in the headspace decreased TGP and NGP quadratically, but increased the concentrations of methane, NMP, and in vitro fiber digestibility linearly, and TMP quadratically. Fiber digestibility, TGP, and NGP increased linearly with increasing bicarbonate concentrations in the medium. Concentrations of methane and NMP were unaffected by bicarbonate concentration, but

  12. Production of Substitute Natural Gas from Coal

    SciTech Connect

    Andrew Lucero

    2009-01-31

    The goal of this research program was to develop and demonstrate a novel gasification technology to produce substitute natural gas (SNG) from coal. The technology relies on a continuous sequential processing method that differs substantially from the historic methanation or hydro-gasification processing technologies. The thermo-chemistry relies on all the same reactions, but the processing sequences are different. The proposed concept is appropriate for western sub-bituminous coals, which tend to be composed of about half fixed carbon and about half volatile matter (dry ash-free basis). In the most general terms the process requires four steps (1) separating the fixed carbon from the volatile matter (pyrolysis); (2) converting the volatile fraction into syngas (reforming); (3) reacting the syngas with heated carbon to make methane-rich fuel gas (methanation and hydro-gasification); and (4) generating process heat by combusting residual char (combustion). A key feature of this technology is that no oxygen plant is needed for char combustion.

  13. Gas and particulate phase products from the ozonolysis of acenaphthylene

    NASA Astrophysics Data System (ADS)

    Riva, Matthieu; Healy, Robert M.; Tomaz, Sophie; Flaud, Pierre-Marie; Perraudin, Emilie; Wenger, John C.; Villenave, Eric

    2016-10-01

    Polycyclic aromatic hydrocarbons (PAHs) are recognized as important secondary organic aerosol (SOA) precursors in the urban atmosphere. In this work, the gas-phase ozonolysis of acenaphthylene was investigated in an atmospheric simulation chamber using a proton transfer reaction time-of-flight-mass spectrometer (PTR-TOF-MS) and an aerosol time-of-flight-mass spectrometer (ATOFMS) for on-line characterization of the oxidation products in the gas and particle phases, respectively. SOA samples were also collected on filters and analyzed by ultra performance liquid chromatography/electrospray ionization high-resolution quadrupole time-of-flight mass spectrometry (UPLC/ESI-HR-QTOFMS) and gas chromatography/electron impact ionization-mass spectrometry (GC/EI-MS). The major gas-phase products included a range of oxygenated naphthalene derivatives such as 1,8-naphthalic anhydride, naphthalene 1,8-dicarbaldehyde and naphthaldehyde, as well as a secondary ozonide. Possible reaction mechanisms are proposed for the formation of these products and favoured pathways have been suggested. Many of these products were also found in the particle phase along with a range of oligomeric compounds. The same range of gas and particle phase products was observed in the presence and absence of excess cyclohexane, an OH scavenger, indicating that OH radical production from the ozonolysis of acenaphthylene is negligible. SOA yields in the range 23-37% were determined and indicate that acenaphthylene ozonolysis may contribute to part of the SOA observed in urban areas.

  14. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    SciTech Connect

    Mueller, Don E.; Marshall, William J.; Wagner, John C.; Bowen, Douglas G.

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  15. Water Resources and Natural Gas Production from the Marcellus Shale

    USGS Publications Warehouse

    Soeder, Daniel J.; Kappel, William M.

    2009-01-01

    The Marcellus Shale is a sedimentary rock formation deposited over 350 million years ago in a shallow inland sea located in the eastern United States where the present-day Appalachian Mountains now stand (de Witt and others, 1993). This shale contains significant quantities of natural gas. New developments in drilling technology, along with higher wellhead prices, have made the Marcellus Shale an important natural gas resource. The Marcellus Shale extends from southern New York across Pennsylvania, and into western Maryland, West Virginia, and eastern Ohio (fig. 1). The production of commercial quantities of gas from this shale requires large volumes of water to drill and hydraulically fracture the rock. This water must be recovered from the well and disposed of before the gas can flow. Concerns about the availability of water supplies needed for gas production, and questions about wastewater disposal have been raised by water-resource agencies and citizens throughout the Marcellus Shale gas development region. This Fact Sheet explains the basics of Marcellus Shale gas production, with the intent of helping the reader better understand the framework of the water-resource questions and concerns.

  16. On-Board Hydrogen Gas Production System For Stirling Engines

    DOEpatents

    Johansson, Lennart N.

    2004-06-29

    A hydrogen production system for use in connection with Stirling engines. The production system generates hydrogen working gas and periodically supplies it to the Stirling engine as its working fluid in instances where loss of such working fluid occurs through usage through operation of the associated Stirling engine. The hydrogen gas may be generated by various techniques including electrolysis and stored by various means including the use of a metal hydride absorbing material. By controlling the temperature of the absorbing material, the stored hydrogen gas may be provided to the Stirling engine as needed. A hydrogen production system for use in connection with Stirling engines. The production system generates hydrogen working gas and periodically supplies it to the Stirling engine as its working fluid in instances where loss of such working fluid occurs through usage through operation of the associated Stirling engine. The hydrogen gas may be generated by various techniques including electrolysis and stored by various means including the use of a metal hydride absorbing material. By controlling the temperature of the absorbing material, the stored hydrogen gas may be provided to the Stirling engine as needed.

  17. Gas Production from Hydrate-Bearing Sediments - Emergent Phenomena -

    SciTech Connect

    Jung, J.W.; Jang, J.W.; Tsouris, Costas; Phelps, Tommy Joe; Rawn, Claudia J; Santamarina, Carlos

    2012-01-01

    Even a small fraction of fine particles can have a significant effect on gas production from hydrate-bearing sediments and sediment stability. Experiments were conducted to investigate the role of fine particles on gas production using a soil chamber that allows for the application of an effective stress to the sediment. This chamber was instrumented to monitor shear-wave velocity, temperature, pressure, and volume change during CO{sub 2} hydrate formation and gas production. The instrumented chamber was placed inside the Oak Ridge National Laboratory Seafloor Process Simulator (SPS), which was used to control the fluid pressure and temperature. Experiments were conducted with different sediment types and pressure-temperature histories. Fines migrated within the sediment in the direction of fluid flow. A vuggy structure formed in the sand; these small cavities or vuggs were precursors to the development of gas-driven fractures during depressurization under a constant effective stress boundary condition. We define the critical fines fraction as the clay-to-sand mass ratio when clays fill the pore space in the sand. Fines migration, clogging, vugs, and gas-driven fracture formation developed even when the fines content was significantly lower than the critical fines fraction. These results show the importance of fines in gas production from hydrate-bearing sediments, even when the fines content is relatively low.

  18. Fission dynamics within time-dependent Hartree-Fock. II. Boost-induced fission

    NASA Astrophysics Data System (ADS)

    Goddard, Philip; Stevenson, Paul; Rios, Arnau

    2016-01-01

    Background: Nuclear fission is a complex large-amplitude collective decay mode in heavy nuclei. Microscopic density functional studies of fission have previously concentrated on adiabatic approaches based on constrained static calculations ignoring dynamical excitations of the fissioning nucleus and the daughter products. Purpose: We explore the ability of dynamic mean-field methods to describe induced fission processes, using quadrupole boosts in the nuclide 240Pu as an example. Methods: Following upon the work presented in Goddard et al. [Phys. Rev. C 92, 054610 (2015)], 10.1103/PhysRevC.92.054610, quadrupole-constrained Hartree-Fock calculations are used to create a potential energy surface. An isomeric state and a state beyond the second barrier peak are excited by means of instantaneous as well as temporally extended gauge boosts with quadrupole shapes. The subsequent deexcitation is studied in a time-dependent Hartree-Fock simulation, with emphasis on fissioned final states. The corresponding fission fragment mass numbers are studied. Results: In general, the energy deposited by the quadrupole boost is quickly absorbed by the nucleus. In instantaneous boosts, this leads to fast shape rearrangements and violent dynamics that can ultimately lead to fission. This is a qualitatively different process than the deformation-induced fission. Boosts induced within a finite time window excite the system in a relatively gentler way and do induce fission but with a smaller energy deposition. Conclusions: The fission products obtained using boost-induced fission in time-dependent Hartree-Fock are more asymmetric than the fragments obtained in deformation-induced fission or the corresponding adiabatic approaches.

  19. Valorization of Flue Gas by Combining Photocatalytic Gas Pretreatment with Microalgae Production.

    PubMed

    Eynde, Erik Van; Lenaerts, Britt; Tytgat, Tom; Blust, Ronny; Lenaerts, Silvia

    2016-03-01

    Utilization of flue gas for algae cultivation seems to be a promising route because flue gas from fossil-fuel combustion processes contains the high amounts of carbon (CO2) and nitrogen (NO) that are required for algae growth. NO is a poor nitrogen source for algae cultivation because of its low reactivity and solublilty in water and its toxicity for algae at high concentrations. Here, we present a novel strategy to valorize NO from flue gas as feedstock for algae production by combining a photocatalytic gas pretreatment unit with a microalgal photobioreactor. The photocatalytic air pretreatment transforms NO gas into NO2 gas and thereby enhances the absorption of NOx in the cultivation broth. The absorbed NOx will form NO2(-) and NO3(-) that can be used as a nitrogen source by algae. The effect of photocatalytic air pretreatment on the growth and biomass productivity of the algae Thalassiosira weissflogii in a semicontinuous system aerated with a model flue gas (1% CO2 and 50 ppm of NO) is investigated during a long-term experiment. The integrated system makes it possible to produce algae with NO from flue gas as the sole nitrogen source and reduces the NOx content in the exhaust gas by 84%.

  20. Valorization of Flue Gas by Combining Photocatalytic Gas Pretreatment with Microalgae Production.

    PubMed

    Eynde, Erik Van; Lenaerts, Britt; Tytgat, Tom; Blust, Ronny; Lenaerts, Silvia

    2016-03-01

    Utilization of flue gas for algae cultivation seems to be a promising route because flue gas from fossil-fuel combustion processes contains the high amounts of carbon (CO2) and nitrogen (NO) that are required for algae growth. NO is a poor nitrogen source for algae cultivation because of its low reactivity and solublilty in water and its toxicity for algae at high concentrations. Here, we present a novel strategy to valorize NO from flue gas as feedstock for algae production by combining a photocatalytic gas pretreatment unit with a microalgal photobioreactor. The photocatalytic air pretreatment transforms NO gas into NO2 gas and thereby enhances the absorption of NOx in the cultivation broth. The absorbed NOx will form NO2(-) and NO3(-) that can be used as a nitrogen source by algae. The effect of photocatalytic air pretreatment on the growth and biomass productivity of the algae Thalassiosira weissflogii in a semicontinuous system aerated with a model flue gas (1% CO2 and 50 ppm of NO) is investigated during a long-term experiment. The integrated system makes it possible to produce algae with NO from flue gas as the sole nitrogen source and reduces the NOx content in the exhaust gas by 84%. PMID:26838336