Initial Back-to-Back Fission Chamber Testing in ATRC

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

Benjamin Chase; Troy Unruh; Joy Rempe

2014-06-01

Development and testing of in-pile, real-time neutron sensors for use in Materials Test Reactor experiments is an ongoing project at Idaho National Laboratory. The Advanced Test Reactor National Scientific User Facility has sponsored a series of projects to evaluate neutron detector options in the Advanced Test Reactor Critical Facility (ATRC). Special hardware was designed and fabricated to enable testing of the detectors in the ATRC. Initial testing of Self-Powered Neutron Detectors and miniature fission chambers produced promising results. Follow-on testing required more experiment hardware to be developed. The follow-on testing used a Back-to-Back fission chamber with the intent to providemore » calibration data, and a means of measuring spectral indices. As indicated within this document, this is the first time in decades that BTB fission chambers have been used in INL facilities. Results from these fission chamber measurements provide a baseline reference for future measurements with Back-to-Back fission chambers.« less

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

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

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

DOE PAGES

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.; ...

2017-02-26

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

Nuclear Design of the HOMER-15 Mars Surface Fission Reactor

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

Poston, David I.

2002-07-01

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

Deep-Earth reactor: Nuclear fission, helium, and the geomagnetic field

PubMed Central

Hollenbach, D. F.; Herndon, J. M.

2001-01-01

Geomagnetic field reversals and changes in intensity are understandable from an energy standpoint as natural consequences of intermittent and/or variable nuclear fission chain reactions deep within the Earth. Moreover, deep-Earth production of helium, having 3He/4He ratios within the range observed from deep-mantle sources, is demonstrated to be a consequence of nuclear fission. Numerical simulations of a planetary-scale geo-reactor were made by using the SCALE sequence of codes. The results clearly demonstrate that such a geo-reactor (i) would function as a fast-neutron fuel breeder reactor; (ii) could, under appropriate conditions, operate over the entire period of geologic time; and (iii) would function in such a manner as to yield variable and/or intermittent output power. PMID:11562483

Deep-Earth reactor: nuclear fission, helium, and the geomagnetic field.

PubMed

Hollenbach, D F; Herndon, J M

2001-09-25

Geomagnetic field reversals and changes in intensity are understandable from an energy standpoint as natural consequences of intermittent and/or variable nuclear fission chain reactions deep within the Earth. Moreover, deep-Earth production of helium, having (3)He/(4)He ratios within the range observed from deep-mantle sources, is demonstrated to be a consequence of nuclear fission. Numerical simulations of a planetary-scale geo-reactor were made by using the SCALE sequence of codes. The results clearly demonstrate that such a geo-reactor (i) would function as a fast-neutron fuel breeder reactor; (ii) could, under appropriate conditions, operate over the entire period of geologic time; and (iii) would function in such a manner as to yield variable and/or intermittent output power.

Neutron-flux profile monitor for use in a fission reactor

DOEpatents

Kopp, M.K.; Valentine, K.H.

1981-09-15

A neutron flux monitor is provided which consists of a plurality of fission counters arranged as spaced-apart point detectors along a delay line. As a fission event occurs in any one of the counters, two delayed current pulses are generated at the output of the delay line. The time separation of the pulses identifies the counter in which the particular fission event occurred. Neutron flux profiles of reactor cores can be more accurately measured as a result.

Neutron flux profile monitor for use in a fission reactor

DOEpatents

Kopp, Manfred K.; Valentine, Kenneth H.

1983-01-01

A neutron flux monitor is provided which consists of a plurality of fission counters arranged as spaced-apart point detectors along a delay line. As a fission event occurs in any one of the counters, two delayed current pulses are generated at the output of the delay line. The time separation of the pulses identifies the counter in which the particular fission event occured. Neutron flux profiles of reactor cores can be more accurately measured as a result.

Radiation Effects in Fission and Fusion Reactors

NASA Astrophysics Data System (ADS)

Odette, G. Robert; Wirth, Brian D.

Since the prediction of "Wigner disease" [1] and the subsequent observation of anisotropic growth of the graphite used in the Chicago Pile, the effects of radiation on materials has been an important technological concern. The broad field of radiation effects impacts many critical advanced technologies, ranging from semiconductor processing to severe materials degradation in nuclear reactor environments. Radiation effects also occur in many natural environments, ranging from deep space to inside the Earth's crust. As selected examples that involve many basic phenomena that cross-cut and illustrate the broader impacts of radiation exposure on materials, this article focuses on modeling microstructural changes in iron-based ferritic alloys under high-energy neutron irradiation relevant to light water fission reactor pressure vessels. We also touch briefly on radiation effects in structural alloys for fusion reactor first wall and blanket structures; in this case the focus is on modeling the evolution of self-interstitial atom clusters and dislocation loops. Note, since even the narrower topic of structural materials for nuclear energy applications encompass a vast literature dating from 1942, the references included in this article are primarily limited to these two narrower subjects. Thus, the references cited here are presented as examples, rather than comprehensive bibliographies. However, the interested reader is referred to proceedings of continuing symposia series that have been sponsored by several organizations, several monographs [2-4] and key journals (e.g., Journal of Nuclear Materials, Radiation Effects and Defects in Solids).

Monitoring system for a liquid-cooled nuclear fission reactor

DOEpatents

DeVolpi, Alexander

1987-01-01

A monitoring system for detecting changes in the liquid levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting changes in the density of the liquid in these regions. A plurality of gamma radiation detectors are used, arranged vertically along the outside of the reactor vessel, and collimator means for each detector limits the gamma-radiation it receives as emitting from only isolated regions of the vessel. Excess neutrons produced by the fission reaction will be captured by the water coolant, by the steel reactor walls, or by the fuel or control structures in the vessel. Neutron capture by steel generates gamma radiation having an energy level of the order of 5-12 MeV, whereas neutron capture by water provides an energy level of approximately 2.2 MeV, and neutron capture by the fission fuel or its cladding provides an energy level of 1 MeV or less. The intensity of neutron capture thus changes significantly at any water-metal interface. Comparative analysis of adjacent gamma detectors senses changes from the normal condition with liquid coolant present to advise of changes in the presence and/or density of the coolant at these specific regions. The gamma detectors can also sense fission-product gas accumulation at the reactor head to advise of a failure of fuel-pin cladding.

Control system for a small fission reactor

DOEpatents

Burelbach, J.P.; Kann, W.J.; Saiveau, J.G.

1985-02-08

A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired.

Future Scenarios for Fission Based Reactors

NASA Astrophysics Data System (ADS)

David, S.

2005-04-01

The coming century will see the exhaustion of standard fossil fuels, coal, gas and oil, which today represent 75% of the world energy production. Moreover, their use will have caused large-scale emission of greenhouse gases (GEG), and induced global climate change. This problem is exacerbated by a growing world energy demand. In this context, nuclear power is the only GEG-free energy source available today capable of responding significantly to this demand. Some scenarios consider a nuclear energy production of around 5 Gtoe in 2050, wich would represent a 20% share of the world energy supply. Present reactors generate energy from the fission of U-235 and require around 200 tons of natural Uranium to produce 1GWe.y of energy, equivalent to the fission of one ton of fissile material. In a scenario of a significant increase in nuclear energy generation, these standard reactors will consume the whole of the world's estimated Uranium reserves in a few decades. However, natural Uranium or Thorium ore, wich are not themselves fissile, can produce a fissile material after a neutron capture ( 239Pu and 233U respectively). In a breeder reactor, the mass of fissile material remains constant, and the fertile ore is the only material to be consumed. In this case, only 1 ton of natural ore is needed to produce 1GWe.y. Thus, the breeding concept allows optimal use of fertile ore and development of sustainable nuclear energy production for several thousand years into the future. Different sustainable nuclear reactor concepts are studied in the international forum "generation IV". Different types of coolant (Na, Pb and He) are studied for fast breeder reactors based on the Uranium cycle. The thermal Thorium cycle requires the use of a liquid fuel, which can be reprocessed online in order to extract the neutron poisons. This paper presents these different sustainable reactors, based on the Uranium or Thorium fuel cycles and will compare the different options in term of fissile

Curved Waveguide Based Nuclear Fission for Small, Lightweight Reactors

NASA Technical Reports Server (NTRS)

Coker, Robert; Putnam, Gabriel

2012-01-01

The focus of the presented work is on the creation of a system of grazing incidence, supermirror waveguides for the capture and reuse of fission sourced neutrons. Within research reactors, neutron guides are a well known tool for directing neutrons from the confined and hazardous central core to a more accessible testing or measurement location. Typical neutron guides have rectangular, hollow cross sections, which are crafted as thin, mirrored waveguides plated with metal (commonly nickel). Under glancing angles with incoming neutrons, these waveguides can achieve nearly lossless transport of neutrons to distant instruments. Furthermore, recent developments have created supermirror surfaces which can accommodate neutron grazing angles up to four times as steep as nickel. A completed system will form an enclosing ring or spherical resonator system to a coupled neutron source for the purpose of capturing and reusing free neutrons to sustain and/or accelerate fission. While grazing incidence mirrors are a known method of directing and safely using neutrons, no method has been disclosed for capture and reuse of neutrons or sustainment of fission using a circular waveguide structure. The presented work is in the process of fabricating a functional, highly curved, neutron supermirror using known methods of Ni-Ti layering capable of achieving incident reflection angles up to four times steeper than nickel alone. Parallel work is analytically investigating future geometries, mirror compositions, and sources for enabling sustained fission with applicability to the propulsion and energy goals of NASA and other agencies. Should research into this concept prove feasible, it would lead to development of a high energy density, low mass power source potentially capable of sustaining fission with a fraction of the standard critical mass for a given material and a broadening of feasible materials due to reduced rates of release, absorption, and non-fission for neutrons. This

Preliminary Comparison of Radioactive Waste Disposal Cost for Fusion and Fission Reactors

NASA Astrophysics Data System (ADS)

Seki, Yasushi; Aoki, Isao; Yamano, Naoki; Tabara, Takashi

1997-09-01

The environmental and economic impact of radioactive waste (radwaste) generated from fusion power reactors using five types of structural materials and a fission reactor has been evaluated and compared. Possible radwaste disposal scenario of fusion radwaste in Japan is considered. The exposure doses were evaluated for the skyshine of gamma-ray during the disposal operation, groundwater migration scenario during the institutional control period of 300 years and future site use scenario after the institutional period. The radwaste generated from a typical light water fission reactor was evaluated using the same methodology as for the fusion reactors. It is found that radwaste from the fusion reactors using F82H and SiC/SiC composites without impurities could be disposed by the shallow land disposal presently applied to the low level waste in Japan. The disposal cost of radwaste from five fusion power reactors and a typical light water reactor were roughly evaluated and compared.

Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Chang, G. S.; Lillo, M. A.

2009-08-01

The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y

ICP-MS analysis of fission product diffusion in graphite for High-Temperature Gas-Cooled Reactors

NASA Astrophysics Data System (ADS)

Carter, Lukas M.

Release of radioactive fission products from nuclear fuel during normal reactor operation or in accident scenarios is a fundamental safety concern. Of paramount importance are the understanding and elucidation of mechanisms of chemical interaction, nuclear interaction, and transport phenomena involving fission products. Worldwide efforts to reduce fossil fuel dependence coupled with an increasing overall energy demand have generated renewed enthusiasm toward nuclear power technologies, and as such, these mechanisms continue to be the subjects of vigorous research. High-Temperature Gas-Cooled Reactors (HTGRs or VHTRs) remain one of the most promising candidates for the next generation of nuclear power reactors. An extant knowledge gap specific to HTGR technology derives from an incomplete understanding of fission product transport in major core materials under HTGR operational conditions. Our specific interest in the current work is diffusion in reactor graphite. Development of methods for analysis of diffusion of multiple fission products is key to providing accurate models for fission product release from HTGR core components and the reactor as a whole. In the present work, a specialized diffusion cell has been developed and constructed to facilitate real-time diffusion measurements via ICP-MS. The cell utilizes a helium gas-jet system which transports diffusing fission products to the mass spectrometer using carbon nanoparticles. The setup was designed to replicate conditions present in a functioning HTGR, and can be configured for real-time release or permeation measurements of single or multiple fission products from graphite or other core materials. In the present work, we have analyzed release rates of cesium in graphite grades IG-110, NBG-18, and a commercial grade of graphite, as well as release of iodine in IG-110. Additionally we have investigated infusion of graphite samples with Cs, I, Sr, Ag, and other surrogate fission products for use in release or

Fission-Produced 99Mo Without a Nuclear Reactor.

PubMed

Youker, Amanda J; Chemerisov, Sergey D; Tkac, Peter; Kalensky, Michael; Heltemes, Thad A; Rotsch, David A; Vandegrift, George F; Krebs, John F; Makarashvili, Vakho; Stepinski, Dominique C

2017-03-01

99 Mo, the parent of the widely used medical isotope 99m Tc, is currently produced by irradiation of enriched uranium in nuclear reactors. The supply of this isotope is encumbered by the aging of these reactors and concerns about international transportation and nuclear proliferation. Methods: We report results for the production of 99 Mo from the accelerator-driven subcritical fission of an aqueous solution containing low enriched uranium. The predominately fast neutrons generated by impinging high-energy electrons onto a tantalum convertor are moderated to thermal energies to increase fission processes. The separation, recovery, and purification of 99 Mo were demonstrated using a recycled uranyl sulfate solution. Conclusion: The 99 Mo yield and purity were found to be unaffected by reuse of the previously irradiated and processed uranyl sulfate solution. Results from a 51.8-GBq 99 Mo production run are presented. © 2017 by the Society of Nuclear Medicine and Molecular Imaging.

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

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

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-19

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritiummore » allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

NASA Astrophysics Data System (ADS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

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

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

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

Kaganas, G.; Rest, J.; Nuclear Engineering Division

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 fuelsmore » 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.« less

Development and Utilization of Space Fission Power Systems

NASA Technical Reports Server (NTRS)

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

2009-01-01

Space fission power systems could enable advanced civilian space missions. Terrestrially, thousands of fission systems have been operated since 1942. In addition, the US flew a space fission system in 1965, and the former Soviet Union flew 33 such systems prior to the end of the Cold War. Modern design and development practices, coupled with 65 years of experience with terrestrial reactors, could enable the affordable development of space fission power systems for near-term planetary surface applications.

Development and Utilization of Space Fission Power Systems

NASA Technical Reports Server (NTRS)

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

2008-01-01

Space fission power systems could enable advanced civilian space missions. Terrestrially, thousands of fission systems have been operated since 1942. In addition, the US flew a space fission system in 1965, and the former Soviet Union flew 33 such systems prior to the end of the Cold War. Modern design and development practices, coupled with 65 years of experience with terrestrial reactors, could enable the affordable development of space fission power systems for near-term planetary surface applications.

NEET Enhanced Micro-Pocket Fission Detector for High Temperature Reactors - FY16 Status Report

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

Unruh, Troy; Reichenberger, Michael; Stevenson, Sarah

2016-09-01

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core fission chambers and thermocouple. As discussed within this report,more » the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year two of this three year project. Highlights from research accomplishments include: • Continuation of a joint collaboration between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. • An updated parallel wire HT MPFD design was developed. • Program support for HT MPFD deployments was given to Accident Tolerant Fuels (ATF) and Advanced Gas-cooled Reactor (AGR) irradiation test programs. • Quality approved materials for HT MPFD construction were procured by irradiation test programs for upcoming deployments. • KSU improved and performed electrical contact and fissile material plating. �

Nuclear fission: a review of experimental advances and phenomenology

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

DOEpatents

Flint, O.

1961-01-10

Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

A thermodynamic approach for advanced fuels of gas-cooled reactors

NASA Astrophysics Data System (ADS)

Guéneau, C.; Chatain, S.; Gossé, S.; Rado, C.; Rapaud, O.; Lechelle, J.; Dumas, J. C.; Chatillon, C.

2005-09-01

For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO 2 gas formation during the chemical interaction of [UO 2± x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

Control system for a small fission reactor

DOEpatents

Burelbach, James P.; Kann, William J.; Saiveau, James G.

1986-01-01

A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired. In another embodiment, a plurality of flexible hollow tubes each containing a neutron absorber are positioned adjacent to one another in spaced relation around the periphery of the reactor vessel and inside the outer neutron reflector with reactivity controlled by the extension and compression of all or some of the coiled hollow tubes. Yet another embodiment of the invention envisions the neutron reflector in the form of an expandable coil spring positioned in an annular space between the reactor vessel and an outer neutron absorbing structure for controlling the neutron flux reflected back into the reactor vessel.

Energy production using fission fragment rockets

NASA Astrophysics Data System (ADS)

Chapline, G.; Matsuda, Y.

1991-08-01

Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: approximately twice the efficiency if the fission fragment energy can be directly converted into electricity; reduction of the buildup of a fission fragment inventory in the reactor could avoid a Chernobyl type disaster; and collection of the fission fragments outside the reactor could simplify the waste disposal problem.

THERMAL FISSION REACTOR COMPOSITIONS AND METHOD OF FABRICATING SAME

DOEpatents

Blainey, A.

1959-10-01

A body is presented for use in a thermal fission reactor comprising a sintered compressed mass of a substance of the group consisting of uranium, thorium, and oxides and carbides of uranium and thorium, enclosed in an envelope of a sintered, compacted, heat-conductive material of the group consisting of beryllium, zirconium, and oxides and carbides of beryllium and zirconium.

Space Fission Propulsion System Development Status

NASA Technical Reports Server (NTRS)

Houts, Mike; VanDyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Williams, Eric; Harper, Roger; Salvail, Pat; Hrbud, Ivana;

Phase 1 Space Fission Propulsion Energy Source Design

NASA Technical Reports Server (NTRS)

Houts, Mike; VanDyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Salvail, Pat; Hrbud, Ivana; Carter, Robert; Rodgers, Stephen L. (Technical Monitor)

2002-01-01

Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and operated. Studies conducted in fiscal year 2001 (IISTP, 2001) show that fission electric propulsion (FEP) systems with a specific mass at or below 50 kg/kWjet could enhance or enable numerous robotic outer solar system missions of interest. At the required specific mass, it is possible to develop safe, affordable systems that meet mission requirements. To help select the system design to pursue, eight evaluation criteria were identified: system integration, safety, reliability, testability, specific mass, cost, schedule, and programmatic risk. A top-level comparison of four potential concepts was performed: a Testable, Passive, Redundant Reactor (TPRR), a Testable Multi-Cell In-Core Thermionic Reactor (TMCT), a Direct Gas Cooled Reactor (DGCR), and a Pumped Liquid Metal Reactor.(PLMR). Development of any of the four systems appears feasible. However, for power levels up to at least 500 kWt (enabling electric power levels of 125-175 kWe, given 25-35% power conversion efficiency) the TPRR has advantages related to several criteria and is competitive with respect to all. Hardware-based research and development has further increased confidence in the TPRR approach. Successful development and utilization of a "Phase I" fission electric propulsion system will enable advanced Phase 2 and Phase 3 systems capable of providing rapid, affordable access to any point in the solar system.

Recent advances in nuclear fission theory: pre- and post-scission physics

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

Talou, Patrick; Kawano, Toshihiko; Bouland, Olivier

2010-01-01

Recent advances in the modeling of the nuclear fission process for data evaluation purposes are reviewed. In particular, it is stressed that a more comprehensive approach to fission data is needed if predictive capability is to be achieved. The link between pre- and post-scission data is clarified, and a path forward to evaluate those data in a consistent and comprehensive manner is presented. Two examples are given: (i) the modeling of fission cross-sections in the R-matrix formalism, for which results for Pu isotopes from 239 to 242 are presented; (ii) the modeling of prompt fission neutrons in the Monte Carlomore » Hauser-Feshbach framework. Results for neutron-induced fission on {sup 235}U are discussed.« less

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

NASA Technical Reports Server (NTRS)

Clement, J. D.

1973-01-01

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

Developments and Tendencies in Fission Reactor Concepts

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Fuji-Ie, Y.

This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC

Burning high-level TRU waste in fusion fission reactors

NASA Astrophysics Data System (ADS)

Shen, Yaosong

2016-09-01

Recently, the concept of actinide burning instead of a once-through fuel cycle for disposing spent nuclear fuel seems to get much more attention. A new method of burning high-level transuranic (TRU) waste combined with Thorium-Uranium (Th-U) fuel in the subcritical reactors driven by external fusion neutron sources is proposed in this paper. The thorium-based TRU fuel burns all of the long-lived actinides via a hard neutron spectrum while outputting power. A one-dimensional model of the reactor concept was built by means of the ONESN_BURN code with new data libraries. The numerical results included actinide radioactivity, biological hazard potential, and much higher burnup rate of high-level transuranic waste. The comparison of the fusion-fission reactor with the thermal reactor shows that the harder neutron spectrum is more efficient than the soft. The Th-U cycle produces less TRU, less radiotoxicity and fewer long-lived actinides. The Th-U cycle provides breeding of 233U with a long operation time (>20 years), hence significantly reducing the reactivity swing while improving safety and burnup.

Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

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

Jennifer Lyons; Wade R. Marcum; Mark D. DeHart

2014-01-01

The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by themore » Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.« less

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

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

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

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

Computer program FPIP-REV calculates fission product inventory for U-235 fission

NASA Technical Reports Server (NTRS)

Brown, W. S.; Call, D. W.

1967-01-01

Computer program calculates fission product inventories and source strengths associated with the operation of U-235 fueled nuclear power reactor. It utilizes a fission-product nuclide library of 254 nuclides, and calculates the time dependent behavior of the fission product nuclides formed by fissioning of U-235.

On fundamental quality of fission chain reaction to oppose rapid runaways of nuclear reactors

NASA Astrophysics Data System (ADS)

Kulikov, G. G.; Shmelev, A. N.; Apse, V. A.; Kulikov, E. G.

2017-01-01

It has been shown that the in-hour equation characterizes the barriers and resistibility of fission chain reaction (FCR) against rapid runaways in nuclear reactors. Traditionally, nuclear reactors are characterized by the presence of barriers based on delayed and prompt neutrons. A new barrier based on the reflector neutrons that can occur when the fast reactor core is surrounded by a weakly absorbing neutron reflector with heavy atomic weight was proposed. It has been shown that the safety of this fast reactor is substantially improved, and considerable elongation of prompt neutron lifetime "devalues" the role of delayed neutron fraction as the maximum permissible reactivity for the reactor safety.

On similarity of various reactor spectra and 235U prompt fission neutron spectrum.

PubMed

Košťál, Michal; Matěj, Zdeněk; Losa, Evžen; Huml, Ondřej; Štefánik, Milan; Cvachovec, František; Schulc, Martin; Jánský, Bohumil; Novák, Evžen; Harutyunyan, Davit; Rypar, Vojtěch

2018-05-01

A well-defined neutron spectrum is an essential tool not only for calibration and testing of neutron detectors used in dosimetry and spectroscopy but also for validation and verification of evaluated cross sections. A new evaluation of thermal-neutron induced 235 U PFNS was performed by the International Atomic Energy Agency (IAEA) in the CIELO (Collaborative International Evaluated Library Organisation Project) project; new measurements of Spectral Averaged Cross sections averaged in the evaluated spectrum are to be obtained. In general, a neutron spectrum in the core is not identical to the pure fission one because fission neutrons undergo many scattering reactions, but it can be shown that PFNS and reactor spectra become undistinguishable from a certain energy boundary. This limit is important for experiments, because when the studied reaction threshold is over this limit, the spectral averaged cross sections in PFNS can be derived from the measured reactions in the reactor core. The evaluation of the neutron spectrum measurements in three different thermal-reactor cores shows that this lower limit is around the energy of 5.5 - 6 MeV. Above this energy the reactor spectra becomes identical with the 235 U PFNS. IAEA CIELO PFNS is within 5% of the measured PFNS from 10 to 14 MeV in a LR-0 reactor, while ENDF/B-VII evaluated PFNS underestimated measured neutron spectra. Copyright © 2018 Elsevier Ltd. All rights reserved.

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

High precision measurements on fission-fragment de-excitation

NASA Astrophysics Data System (ADS)

Oberstedt, Stephan; Gatera, Angélique; Geerts, Wouter; Göök, Alf; Hambsch, Franz-Josef; Vidali, Marzio; Oberstedt, Andreas

2017-11-01

In recent years nuclear fission has gained renewed interest both from the nuclear energy community and in basic science. The first, represented by the OECD Nuclear Energy Agency, expressed the need for more accurate fission cross-section and fragment yield data for safety assessments of Generation IV reactor systems. In basic science modelling made much progress in describing the de-excitation mechanism of neutron-rich isotopes, e.g. produced in nuclear fission. Benchmarking the different models require a precise experimental data on prompt fission neutron and γ-ray emission, e.g. multiplicity, average energy per particle and total dissipated energy per fission, preferably as function of fission-fragment mass and total kinetic energy. A collaboration of scientists from JRC Geel (formerly known as JRC IRMM) and other institutes took the lead in establishing a dedicated measurement programme on prompt fission neutron and γ-ray characteristics, which has triggered even more measurement activities around the world. This contribution presents new advanced instrumentation and methodology we use to generate high-precision spectral data and will give a flavour of future data needs and opportunities.

Advanced Test Reactor Tour

ScienceCinema

Miley, Don

2017-12-21

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored.

Physics and potentials of fissioning plasmas for space power and propulsion

NASA Technical Reports Server (NTRS)

Thom, K.; Schwenk, F. C.; Schneider, R. T.

1976-01-01

Fissioning uranium plasmas are the nuclear fuel in conceptual high-temperature gaseous-core reactors for advanced rocket propulsion in space. A gaseous-core nuclear rocket would be a thermal reactor in which an enriched uranium plasma at about 10,000 K is confined in a reflector-moderator cavity where it is nuclear critical and transfers its fission power to a confining propellant flow for the production of thrust at a specific impulse up to 5000 sec. With a thrust-to-engine weight ratio approaching unity, the gaseous-core nuclear rocket could provide for propulsion capabilities needed for manned missions to the nearby planets and for economical cislunar ferry services. Fueled with enriched uranium hexafluoride and operated at temperatures lower than needed for propulsion, the gaseous-core reactor scheme also offers significant benefits in applications for space and terrestrial power. They include high-efficiency power generation at low specific mass, the burnup of certain fission products and actinides, the breeding of U-233 from thorium with short doubling times, and improved convenience of fuel handling and processing in the gaseous phase.

The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

NASA Astrophysics Data System (ADS)

Wagemans, Jan; Malambu, Edouard; Borms, Luc; Fiorito, Luca

2016-02-01

The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma) irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f) prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f) prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

A Comparison of Fast-Spectrum and Moderated Space Fission Reactors

NASA Astrophysics Data System (ADS)

Poston, David I.

2005-02-01

The reactor neutron spectrum is one of the fundamental design choices for any fission reactor, but the implications of using a moderated spectrum are vastly different for space reactors as opposed to terrestrial reactors. In addition, the pros and cons of neutron spectra are significantly different among many of the envisioned space power applications. This paper begins with a discussion of the neutronic differences between fast-spectrum and moderated space reactors. This is followed by a discussion of the pros and cons of fast-spectrum and moderated space reactors separated into three areas—technical risk, performance, and safety/safeguards. A mix of quantitative and qualitative arguments is presented, and some conclusions generally can be made regarding neutron spectrum and space power application. In most cases, a fast-spectrum system appears to be the better alternative (mostly because of simplicity and higher potential operating temperatures); however, in some cases, such as a low-power (<100-kWt) surface reactor, a moderated spectrum could provide a better approach. In all cases, the determination of which spectrum is preferred is a strong function of the metrics provided by the "customer"— i.e., if a certain level of performance is required, it could provide a different solution than if a certain level of safeguards is required (which in some cases could produce a null solution). The views expressed in this document are those of the author and do not necessarily reflect agreement by the Government.

NEET Enhanced Micro Pocket Fission Detector for High Temperature Reactors - FY15 Status Report

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

Unruh, Troy; McGregor, Douglas; Ugorowski, Phil

2015-09-01

A new project, that is a collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core parallel plate fission chambermore » and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year one of this three year project. Highlights from research accomplishments include: A joint collaboration was initiated between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. An updated HT MPFD design was developed. New high temperature-compatible materials for HT MPFD construction were procured. Construction methods to support the new design were evaluated at INL. Laboratory evaluations of HT MPFD were initiated. Electrical contact and fissile material plating has been performed at KSU. Updated detector electronics are undergoing evaluations at KSU. A

Contributions of each isotope in some fluids on neutronic performance in a fusion-fission hybrid reactor: a Monte Carlo method

NASA Astrophysics Data System (ADS)

Günay, M.; Şarer, B.; Kasap, H.

2014-08-01

In the present investigation, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa ferritic steel structural material and 99-95 % Li20Sn80-1-5 % SFG-Pu, 99-95 % Li20Sn80-1-5 % SFG-PuF4, 99-95 % Li20Sn80-1-5 % SFG-PuO2 the molten salt-heavy metal mixtures, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium zone with the width of 3 cm was used for the neutron multiplicity between liquid first wall and blanket. The contributions of each isotope in fluids on the nuclear parameters of a fusion-fission hybrid reactor such as tritium breeding ratio, energy multiplication factor, heat deposition rate were computed in liquid first wall, blanket and shield zones. Three-dimensional analyses were performed by using Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

Fission-suppressed fusion breeder on the thorium cycle and nonproliferation

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

Moir, R. W.

2012-06-19

Fusion reactors could be designed to breed fissile material while suppressing fissioning thereby enhancing safety. The produced fuel could be used to startup and makeup fuel for fission reactors. Each fusion reaction can produce typically 0.6 fissile atoms and release about 1.6 times the 14 MeV neutron's energy in the blanket in the fission-suppressed design. This production rate is 2660 kg/1000 MW of fusion power for a year. The revenues would be doubled from such a plant by selling fuel at a price of 60/g and electricity at $0.05/kWh for Q=P{sub fusion}/P{sub input}=4. Fusion reactors could be designed to destroymore » fission wastes by transmutation and fissioning but this is not a natural use of fusion whereas it is a designed use of fission reactors. Fusion could supply makeup fuel to fission reactors that were dedicated to fissioning wastes with some of their neutrons. The design for safety and heat removal and other items is already accomplished with fission reactors. Whereas fusion reactors have geometry that compromises safety with a complex and thin wall separating the fusion zone from the blanket zone where wastes could be destroyed. Nonproliferation can be enhanced by mixing {sup 233}U with {sup 238}U. Also nonproliferation is enhanced in typical fission-suppressed designs by generating up to 0.05 {sup 232}U atoms for each {sup 233}U atom produced from thorium, about twice the IAEA standards of 'reduced protection' or 'self protection.' With 2.4%{sup 232}U, high explosive material is predicted to degrade owing to ionizing radiation after a little over 1/2 year and the heat rate is 77 W just after separation and climbs to over 600 W ten years later. The fissile material can be used to fuel most any fission reactor but is especially appropriate for molten salt reactors (MSR) also called liquid fluoride thorium reactors (LFTR) because of the molten fuel does not need hands on fabrication and handling.« less

An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

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

Bucknor, M.; Farmer, M.; Grabaskas, D.

The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium-cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool, developed as part of the Integral Fast Reactor (IFR) program, was utilized in an MST trial calculation. This tool was developed by applying classical theories of aerosol scrubbing to the decontamination ofmore » gases produced as a result of postulated fuel pin failures during an SFR accident scenario. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapour condensation on aerosol scrubbing are also treated. This paper provides details of the individual scrubbing mechanisms utilized in the IFR code as well as results from a trial mechanistic source term assessment led by Argonne National Laboratory in 2016.« less

Development and Testing of Space Fission Technology at NASA-MSFC

NASA Technical Reports Server (NTRS)

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

2008-01-01

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

Investigation of Isotopically Tailored Boron in Advanced Fission and Fusion Reactor Systems.

NASA Astrophysics Data System (ADS)

Domaszek, Gerald Raymond

effectiveness of various materials as reflector control elements for a compact space reactor has shown that B^{11} is neutronically superior to graphite in these applications. Metallic boron and boron carbide isotopically enriched in B^{11} have been demonstrated to be neutronically acceptable for varied applications in advanced reactor systems. B^ {11} has been shown to be superior in performance to graphite. While only somewhat inferior to beryllium as neutron multipliers, B^ {11} and B^{11} _4C have safety, supply and cost advantage over beryllium. (Abstract shortened with permission of author.).

Fissioning uranium plasmas and nuclear-pumped lasers

NASA Technical Reports Server (NTRS)

Schneider, R. T.; Thom, K.

1975-01-01

Current research into uranium plasmas, gaseous-core (cavity) reactors, and nuclear-pumped lasers is discussed. Basic properties of fissioning uranium plasmas are summarized together with potential space and terrestrial applications of gaseous-core reactors and nuclear-pumped lasers. Conditions for criticality of a uranium plasma are outlined, and it is shown that the nonequilibrium state and the optical thinness of a fissioning plasma can be exploited for the direct conversion of fission fragment energy into coherent light (i.e., for nuclear-pumped lasers). Successful demonstrations of nuclear-pumped lasers are described together with gaseous-fuel reactor experiments using uranium hexafluoride.

Experimental Constraints on Neutrino Spectra Following Fission

NASA Astrophysics Data System (ADS)

Napolitano, Jim; Daya Bay Collaboration

2016-09-01

We discuss new initiatives to constrain predictions of fission neutrino spectra from nuclear reactors. These predictions are germane to the understanding of reactor flux anomalies; are needed to reduce systematic uncertainty in neutrino oscillation spectra; and inform searches for the diffuse supernova neutrino background. The initiatives include a search for very high- Q beta decay components to the neutrino spectrum from the Daya Bay power plant; plans for a measurement of the β- spectrum from 252Cf fission products; and precision measurements of the 235U fission neutrino spectrum from PROSPECT and other very short baseline reactor experiments.

Assessment of the high temperature fission chamber technology for the French fast reactor program

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

Jammes, C.; Filliatre, P.; Geslot, B.

2011-07-01

High temperature fission chambers are key instruments for the control and protection of the sodium-cooled fast reactor. First, the developments of those neutron detectors, which are carried out either in France or abroad are reviewed. Second, the French realizations are assessed with the use of the technology readiness levels in order to identify tracks of improvement. (authors)

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

2017-12-01

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

Space Fission Propulsion System Development Status

NASA Astrophysics Data System (ADS)

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

2001-01-01

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

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

NASA Technical Reports Server (NTRS)

Garber, Anne E.; Dickens, Ricky E.

2011-01-01

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

Advanced Demonstration and Test Reactor Options Study

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

Petti, David Andrew; Hill, R.; Gehin, J.

Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercializationmore » of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy�

Prompt fission γ-ray data from spontaneous fission and the mechanism of fission-fragment de-excitation

NASA Astrophysics Data System (ADS)

Oberstedt, Stephan; Dragic, Aleksandar; Gatera, Angelique; Göök, Alf; Hambsch, Franz-Josef; Oberstedt, Andreas

2017-09-01

The investigation of prompt γ-ray emission in nuclear fission has a great relevance for the assessment of prompt heat generation in a reactor core and for the better understanding of the de-excitation mechanism of fission fragments. Some years ago experimental data was scarce and available only from a few fission reactions, 233,235U(nth, f), 239Pu(nth, f), and 252Cf(sf). Initiated by a high priority data request published by the OECD/NEA a dedicated prompt fission γ-ray measurement program is being conducted at the Joint Research Centre Geel. In recent years we obtained new and accurate prompt fission γ-ray spectrum (PFGS) characteristics (average number of photons per fission, average total energy per fission and mean photon energy) from 252Cf(sf), 235U(nth, f) and 239,241Pu(nth, f) within 2% of uncertainty. In order to understand the dependence of prompt fission γ-ray emission on the compound nuclear mass and excitation energy, we started a first measurement campaign on spontaneously fissioning plutonium and curium isotopes. Results on PFGS characteristics from 240,242Pu(sf) show a dependence on the fragment mass distribution rather than on the average prompt neutron multiplicity, pointing to a more complex competition between prompt fission γ-ray and neutron emission.

Advanced Power Conversion Efficiency in Inventive Plasma for Hybrid Toroidal Reactor

NASA Astrophysics Data System (ADS)

Hançerlioğullari, Aybaba; Cini, Mesut; Güdal, Murat

2013-08-01

Apex hybrid reactor has a good potential to utilize uranium and thorium fuels in the future. This toroidal reactor is a type of system that facilitates the occurrence of the nuclear fusion and fission events together. The most important feature of hybrid reactor is that the first wall surrounding the plasma is liquid. The advantages of utilizing a liquid wall are high power density capacity good power transformation productivity, the magnitude of the reactor's operational duration, low failure percentage, short maintenance time and the inclusion of the system's simple technology and material. The analysis has been made using the MCNP Monte Carlo code and ENDF/B-V-VI nuclear data. Around the fusion chamber, molten salts Flibe (LI2BeF4), lead-lithium (PbLi), Li-Sn, thin-lityum (Li20Sn80) have used as cooling materials. APEX reactor has modeled in the torus form by adding nuclear materials of low significance in the specified percentages between 0 and 12 % to the molten salts. In this study, the neutronic performance of the APEX fusion reactor using various molten salts has been investigated. The nuclear parameters of Apex reactor has been searched for Flibe (LI2BeF4) and Li-Sn, for blanket layers. In case of usage of the Flibe (LI2BeF4), PbLi, and thin-lityum (Li20Sn80) salt solutions at APEX toroidal reactors, fissile material production per source neutron, tritium production speed, total fission rate, energy reproduction factor has been calculated, the results obtained for both salt solutions are compared.

NEET Micro-Pocket Fission Detector. Final Project report

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

Unruh, T.; Rempe, Joy; McGregor, Douglas

2014-09-01

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Alternative Energies and Atomic Energy Commission, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), is funded by the Nuclear Energy Enabling Technologies (NEET) program to develop and test Micro-Pocket Fission Detectors (MPFDs), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package. When deployed, these sensors will significantly advance flux detection capabilities for irradiation tests in US Material Test Reactors (MTRs). Ultimately, evaluations may lead to a more compact, more accurate, andmore » longer lifetime flux sensor for critical mock-ups, and high performance reactors, allowing several Department of Energy Office of Nuclear Energy (DOE-NE) programs to obtain higher accuracy/higher resolution data from irradiation tests of candidate new fuels and materials. Specifically, deployment of MPFDs will address several challenges faced in irradiations performed at MTRs: Current fission chamber technologies do not offer the ability to measure fast flux, thermal flux and temperature within a single compact probe; MPFDs offer this option. MPFD construction is very different than current fission chamber construction; the use of high temperature materials allow MPFDs to be specifically tailored to survive harsh conditions encountered in-core of high performance MTRs. The higher accuracy, high fidelity data available from the compact MPFD will significantly enhance efforts to validate new high-fidelity reactor physics codes and new multi-scale, multi-physics codes. MPFDs can be built with variable sensitivities to survive the lifetime of an experiment or fuel assembly in some MTRs, allowing for more efficient and cost effective power monitoring. The small size of the MPFDs allows multiple sensors to be deployed, offering the potential to

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

NASA Astrophysics Data System (ADS)

Seibert, Rachel L.; Terrani, Kurt A.; Velázquez, Daniel; Hunn, John D.; Baldwin, Charles A.; Montgomery, Fred C.; Terry, Jeff

2018-03-01

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Program and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form PdxSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. They may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.

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. Copyright © 2015 Elsevier Ltd. All rights reserved.

Fission Surface Power Technology Development Update

NASA Technical Reports Server (NTRS)

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

2011-01-01

Power is a critical consideration in planning exploration of the surfaces of the Moon, Mars, and places beyond. Nuclear power is an important option, especially for locations in the solar system where sunlight is limited or environmental conditions are challenging (e.g., extreme cold, dust storms). NASA and the Department of Energy are maintaining the option for fission surface power for the Moon and Mars by developing and demonstrating technology for a fission surface power system. The Fission Surface Power Systems project has focused on subscale component and subsystem demonstrations to address the feasibility of a low-risk, low-cost approach to space nuclear power for surface missions. Laboratory demonstrations of the liquid metal pump, reactor control drum drive, power conversion, heat rejection, and power management and distribution technologies have validated that the fundamental characteristics and performance of these components and subsystems are consistent with a Fission Surface Power preliminary reference concept. In addition, subscale versions of a non-nuclear reactor simulator, using electric resistance heating in place of the reactor fuel, have been built and operated with liquid metal sodium-potassium and helium/xenon gas heat transfer loops, demonstrating the viability of establishing system-level performance and characteristics of fission surface power technologies without requiring a nuclear reactor. While some component and subsystem testing will continue through 2011 and beyond, the results to date provide sufficient confidence to proceed with system level technology readiness demonstration. To demonstrate the system level readiness of fission surface power in an operationally relevant environment (the primary goal of the Fission Surface Power Systems project), a full scale, 1/4 power Technology Demonstration Unit (TDU) is under development. The TDU will consist of a non-nuclear reactor simulator, a sodium-potassium heat transfer loop, a power

Total absorption spectroscopy of fission fragments relevant for reactor antineutrino spectra

DOE PAGES

Fallot, M.; Porta, A.; Meur, L. Le; ...

2017-09-13

Here, the accurate determination of reactor antineutrino spectra remains a very active research topic for which new methods of study have emerged in recent years. Indeed, following the long-recognized reactor anomaly (measured antineutrino deficit in short baseline reactor experiments when compared with spectral predictions), the three international reactor neutrino experiments Double Chooz, Daya Bay and Reno have recently demonstrated the existence of spectral distortions in their measurements with respect to the same predictions. These spectral predictions were obtained through the conversion of integral beta-energy spectra obtained at the ILL research reactor. Several studies have shown that the underlying nuclear physicsmore » required for the conversion of these spectra into antineutrino spectra is not totally understood. An alternative to such converted spectra is a complementary approach that consists of determining the antineutrino spectrum by means of the measurement and processing of nuclear data. The beta properties of some key fission products suffer from the pandemonium effect which can be circumvented by the use of the Total Absorption Gamma-ray Spectroscopy technique (TAGS). The two main contributors to the Pressurized Water Reactor antineutrino spectrum in the region where the spectral distortion has been observed are 92Rb and 142Cs, which have been measured at the radioactive beam facility of the University of Jyvaskyla in two TAGS experiments. We present the results of the analysis of the TAGS measurements of the β-decay properties of 92Rb along with preliminary results on 142Cs and report on the measurements already performed.« less

Total absorption spectroscopy of fission fragments relevant for reactor antineutrino spectra

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

Fallot, M.; Porta, A.; Meur, L. Le

Here, the accurate determination of reactor antineutrino spectra remains a very active research topic for which new methods of study have emerged in recent years. Indeed, following the long-recognized reactor anomaly (measured antineutrino deficit in short baseline reactor experiments when compared with spectral predictions), the three international reactor neutrino experiments Double Chooz, Daya Bay and Reno have recently demonstrated the existence of spectral distortions in their measurements with respect to the same predictions. These spectral predictions were obtained through the conversion of integral beta-energy spectra obtained at the ILL research reactor. Several studies have shown that the underlying nuclear physicsmore » required for the conversion of these spectra into antineutrino spectra is not totally understood. An alternative to such converted spectra is a complementary approach that consists of determining the antineutrino spectrum by means of the measurement and processing of nuclear data. The beta properties of some key fission products suffer from the pandemonium effect which can be circumvented by the use of the Total Absorption Gamma-ray Spectroscopy technique (TAGS). The two main contributors to the Pressurized Water Reactor antineutrino spectrum in the region where the spectral distortion has been observed are 92Rb and 142Cs, which have been measured at the radioactive beam facility of the University of Jyvaskyla in two TAGS experiments. We present the results of the analysis of the TAGS measurements of the β-decay properties of 92Rb along with preliminary results on 142Cs and report on the measurements already performed.« less

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

PubMed

Chang, G S; Ambrosek, R G

2005-01-01

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

Heat transfer in a fissioning uranium plasma reactor cavity

NASA Technical Reports Server (NTRS)

Kascak, A. F.

1973-01-01

Two schemes are investigated by which a fission-heated uranium plasma located in the central cavity of a test reactor could be insulated to keep its temperature above condensation in a neutron flux of 10 to the 15th power neutrons/(sq cm)(sec) or less. The first scheme was to use a mirrored cavity wall to reflect the thermal radiation back into the plasma. The second scheme was to seed the transpirational cavity wall coolant so as to make it opaque to thermal radiation, thus insulating the hot plasma from the cold wall. The analysis showed that a mirrored cavity wall must have a reflectivity of over 95 percent or that seeded argon must be used as the wall coolant to give an acceptable operating margin above fuel condensation conditions.

Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

NASA Technical Reports Server (NTRS)

Butler, C.; Albright, D.

2007-01-01

Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

Investigation of applications for high-power, self-critical fissioning uranium plasma reactors

NASA Technical Reports Server (NTRS)

Rodgers, R. J.; Latham, T. S.; Krascella, N. L.

1976-01-01

Analytical studies were conducted to investigate potentially attractive applications for gaseous nuclear cavity reactors fueled by uranium hexafluoride and its decomposition products at temperatures of 2000 to 6000 K and total pressures of a few hundred atmospheres. Approximate operating conditions and performance levels for a class of nuclear reactors in which fission energy removal is accomplished principally by radiant heat transfer from the high temperature gaseous nuclear fuel to surrounding absorbing media were determined. The results show the radiant energy deposited in the absorbing media may be efficiently utilized in energy conversion system applications which include (1) a primary energy source for high thrust, high specific impulse space propulsion, (2) an energy source for highly efficient generation of electricity, and (3) a source of high intensity photon flux for heating working fluid gases for hydrogen production or MHD power extraction.

MCNP6 simulated performance of Micro-Pocket Fission Detectors (MPFDs) in the Transient REActor Test (TREAT) Facility

DOE PAGES

Reichenberger, Michael A.; Patel, Vishal K.; Roberts, Jeremy A.; ...

2017-03-03

Here, Micro-Pocket Fission Detectors (MPFDs) are under development for in-core neutron flux measurements at the Transient REActor Test facility (TREAT) and in other experiments at Idaho National Laboratory (INL). The sensitivity of MPFDs to the energy dependent neutron flux at TREAT has been determined for 0.0300-μm thick active material coatings of 242Pu, 232Th, natural uranium, and 93% enriched 235U. Self-shielding effects in the active material of the MPFD was also confirmed to be negligible. Finally, fission fragment energy deposition was found to be in conformance with previously reported results.

Comparison Of 252Cf Time Correlated Induced Fisssion With AmLi Induced Fission On Fresh MTR Research Reactor Fuel

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

Joshi, Jay Prakash

The effective application of international safeguards to research reactors requires verification of spent fuel as well as fresh fuel. To accomplish this goal various nondestructive and destructive assay techniques have been developed in the US and around the world. The Advanced Experimental Fuel Counter (AEFC) is a nondestructive assay (NDA) system developed at Los Alamos National Laboratory (LANL) combining both neutron and gamma measurement capabilities. Since spent fuel assemblies are stored in water, the system was designed to be watertight to facilitate underwater measurements by inspectors. The AEFC is comprised of six 3He detectors as well as a shielded andmore » collimated ion chamber. The 3He detectors are used for active and passive neutron coincidence counting while the ion chamber is used for gross gamma counting. Active coincidence measurement data is used to measure residual fissile mass, whereas the passive coincidence measurement data along with passive gamma measurement can provide information about burnup, cooling time, and initial enrichment. In the past, most of the active interrogation systems along with the AEFC used an AmLi neutron interrogation source. Owing to the difficulty in obtaining an AmLi source, a 252Cf spontaneous fission (SF) source was used during a 2014 field trail in Uzbekistan as an alternative. In this study, experiments were performed to calibrate the AEFC instrument and compare use of the 252Cf spontaneous fission source and the AmLi (α,n) neutron emission source. The 252Cf source spontaneously emits bursts of time-correlated prompt fission neutrons that thermalize in the water and induce fission in the fuel assembly. The induced fission (IF) neutrons are also time correlated resulting in more correlated neutron detections inside the 3He detector, which helps reduce the statistical errors in doubles when using the 252Cf interrogation source instead of the AmLi source. In this work, two MTR fuel assemblies varying both in

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

DOE PAGES

Seibert, Rachel L.; Terrani, Kurt A.; Velázquez, Daniel; ...

2018-03-01

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Programmore » and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form Pd xSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. In conclusion, they may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.« less

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

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

Seibert, Rachel L.; Terrani, Kurt A.; Velázquez, Daniel

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Programmore » and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form Pd xSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. In conclusion, they may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.« less

Thermomechanics of candidate coatings for advanced gas reactor fuels

NASA Astrophysics Data System (ADS)

Nosek, A.; Conzen, J.; Doescher, H.; Martin, C.; Blanchard, J.

2007-09-01

Candidate fuel/coating combinations for an advanced, coated-fuel particle for a gas-cooled fast reactor (GFR) have been evaluated. These all-ceramic fuel forms consist of a fuel kernel made of UC or UN, surrounded with two shells (a buffer and a coating) made of TiC, SiC, ZrC, TiN, or ZrN. These carbides and nitrides are analyzed with finite element models to determine the stresses produced in the micro fuel particles from differential thermal expansion, fission gas release, swelling, and creep during particle fabrication and reactor operation. This study will help determine the feasibility of different fuel and coating combinations and identify the critical loads. The analysis shows that differential thermal expansion of the fuel and coating dictate the amount of stress for changing temperatures (such as during fabrication), and that the coating creep is able to mitigate an otherwise overwhelming amount of stress from fuel swelling. Because fracture is a likely mode of failure, a fracture mechanics study is also included to identify the relative likelihood of catastrophic fracture of the coating and resulting gas release. Overall, the analysis predicts that UN/ZrC is the best thermomechanical fuel/coating combination for mitigating the stress within the new fuel particle, but UN/TiN and UN/ZrN could also be strong candidates if their unknown creep rates are sufficiently large.

Unit mechanisms of fission gas release: Current understanding and future needs

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

Tonks, Michael; Andersson, David; Devanathan, Ram

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel properties and, once the gas is released into the gap between the fuel and cladding, lowering gap thermal conductivity and increasing gap pressure. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are being applied to provide unprecedented understanding of the unit mechanisms that define the fission product behavior. In this article, existing research on the basic mechanisms behind the various stages of fission gas releasemore » during normal reactor operation are summarized and critical areas where experimental and simulation work is needed are identified. This basic understanding of the fission gas behavior mechanisms has the potential to revolutionize our ability to predict fission product behavior during reactor operation and to design fuels that have improved fission product retention. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.« less

Impact of Fission Neutron Energies on Reactor Antineutrino Spectra

NASA Astrophysics Data System (ADS)

Hermanek, Keith; Littlejohn, Bryce; Gustafson, Ian

2017-09-01

Recent measurements of the reactor antineutrino spectra (Double Chooz, Reno, and Daya Bay) have shown a discrepancy in the 5-7 MeV region when compared to current theoretical models (Vogel and Huber-Mueller). There are numerous theories pertaining to this antineutrino anomaly, including theories that point to new physics beyond the standard model. In the paper ``Possible Origins and Implications of the Shoulder in Reactor Neutrino Spectra'' by A. Hayes et al., explanations for this anomaly are suggested. One theory is that there are interactions from fast and epithermal incident neutrons which are significant enough to create more events in the 5-7 MeV by a noticeable amount. In our research, we used the Oklo software network created by Dan Dwyer. This generates ab initio antineutrino and beta decay spectra based on standard fission yield databases ENDF, JENDL, JEFF, and the beta decay transition database ENSDF-6. Utilizing these databases as inputs, we show with reasonable assumptions one can prove contributions of fast and epithermal neutrons is less than 3% in the 5-7 MeV region. We also discovered rare isotopes are present in beta decay chains but not well measured and have no corresponding database information, and studied its effect onto the spectrum.

Advanced Fuel Cycles for Fusion Reactors: Passive Safety and Zero-Waste Options

NASA Astrophysics Data System (ADS)

Zucchetti, Massimo; Sugiyama, Linda E.

2006-05-01

Nuclear fusion is seen as a much ''cleaner'' energy source than fission. Most of the studies and experiments on nuclear fusion are currently devoted to the Deuterium-Tritium (DT) fuel cycle, since it is the easiest way to reach ignition. The recent stress on safety by the world's community has stimulated the research on other fuel cycles than the DT one, based on 'advanced' reactions, such as the Deuterium-Helium-3 (DHe) one. These reactions pose problems, such as the availability of 3He and the attainment of the higher plasma parameters that are required for burning. However, they have many advantages, like for instance the very low neutron activation, while it is unnecessary to breed and fuel tritium. The extrapolation of Ignitor technologies towards a larger and more powerful experiment using advanced fuel cycles (Candor) has been studied. Results show that Candor does reach the passive safety and zero-waste option. A fusion power reactor based on the DHe cycle could be the ultimate response to the environmental requirements for future nuclear power plants.

Non-nuclear Testing of Reactor Systems in the Early Flight Fission Test Facilities (EFF-TF)

NASA Technical Reports Server (NTRS)

VanDyke, Melissa; Martin, James

2004-01-01

The Early Flight Fission-Test Facility (EFF-TF) can assist in the &sign and development of systems through highly effective non-nuclear testing of nuclear systems when technical issues associated with near-term space fission systems are "non-nuclear" in nature (e.g. system s nuclear operations are understood). For many systems. thermal simulators can he used to closely mimic fission heat deposition. Axial power profile, radial power profile. and fuel pin thermal conductivity can be matched. In addition to component and subsystem testing, operational and lifetime issues associated with the steady state and transient performance of the integrated reactor module can be investigated. Instrumentation at the EFF-TF allows accurate measurement of temperature, pressure, strain, and bulk core deformation (useful for accurately simulating nuclear behavior). Ongoing research at the EFF-TF is geared towards facilitating research, development, system integration, and system utilization via cooperative efforts with DOE laboratories, industry, universities, and other NASA centers. This paper describes the current efforts for the latter portion of 2003 and beginning of 2004.

THE ARMOUR DUST FUELED REACTOR (ADFR)

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

Krucoff, D.

1958-01-01

The A-DFR is based on the use of a fissionable dust carried in a gas. This fuel ferm offers promise of a major economic advance through the use of 2,000 to 3,000 F operating temperatures and a low cost fuel cycle. The development program is described that was initiated to investigate experimentally the proposed fuel and study analytically other reactor characteristics. A brief review of the reactor concept is presented. (W.D.M.)

Advanced shield development for a fission surface power system for the lunar surface

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

A. E. Craft; I. J. Silver; C. M. Clark

A nuclear reactor power system such as the affordable fission surface power system enables a potential outpostonthemoon.Aradiation shieldmustbe included in the reactor system to reduce the otherwise excessive dose to the astronauts and other vital system components. The radiation shield is typically the most massive component of a space reactor system, and thus must be optimized to reduce mass asmuchas possible while still providing the required protection.Various shield options for an on-lander reactor system are examined for outpost distances of 400m and 1 kmfromthe reactor. Also investigated is the resulting mass savings from the use of a high performance cermetmore » fuel. A thermal analysis is performed to determine the thermal behaviours of radiation shields using borated water. For an outpost located 1000m from the core, a tetramethylammonium borohydride shield is the lightest (5148.4 kg), followed by a trilayer shield (boron carbide–tungsten–borated water; 5832.3 kg), and finally a borated water shield (6020.7 kg). In all of the final design cases, the temperature of the borated water remains below 400 K.« less

Unit mechanisms of fission gas release: Current understanding and future needs

NASA Astrophysics Data System (ADS)

Tonks, Michael; Andersson, David; Devanathan, Ram; Dubourg, Roland; El-Azab, Anter; Freyss, Michel; Iglesias, Fernando; Kulacsy, Katalin; Pastore, Giovanni; Phillpot, Simon R.; Welland, Michael

2018-06-01

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel and gap properties. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are beginning to reveal new understanding of the unit mechanisms that define fission product behavior. Here, existing research on the basic mechanisms of fission gas release during normal reactor operation are summarized and critical areas where work is needed are identified. This basic understanding of the fission gas behavior mechanisms has the potential to revolutionize our ability to predict fission product behavior and to design fuels with improved performance. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.

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.

Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

NASA Technical Reports Server (NTRS)

Godfroy, Thomas; Dickens, Ricky; Houts, Michael; Pearson, Boise; Webster, Kenny; Gibson, Marc; Qualls, Lou; Poston, Dave; Werner, Jim; Radel, Ross

2011-01-01

The Nuclear Systems Team at NASA Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and Mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program, which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter for tests at MSFC. When tested at NASA Glenn Research Center (GRC) the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumentation (temperature, pressure, flow) for data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

NASA Astrophysics Data System (ADS)

Godfroy, T.; Dickens, R.; Houts, M.; Pearson, B.; Webster, K.; Gibson, M.; Qualls, L.; Poston, D.; Werner, J.; Radel, R.

The Nuclear Systems Team at Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter when being tested at MSFC. When tested at GRC the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumenta- tion (temperature, pressure, flow) data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

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

NASA Astrophysics Data System (ADS)

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

2002-01-01

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

Fission cross-sections, prompt fission neutron and γ-ray emission in request for nuclear applications

NASA Astrophysics Data System (ADS)

Hambsch, F.-J.; Salvador-Castiñeira, P.; Oberstedt, S.; Göök, A.; Billnert, R.

2016-06-01

In recent years JRC-IRMM has been investigating fission cross-sections of 240,242Pu in the fast-neutron energy range relevant for innovative reactor systems and requested in the High Priority Request List (HPRL) of the OECD/Nuclear Energy Agency (NEA). In addition to that, prompt neutron multiplicities are being investigated for the major isotopes 235U, 239Pu in the neutron-resonance region using a newly developed scintillation detector array (SCINTIA) and an innovative modification of the Frisch-grid ionisation chamber for fission-fragment detection. These data are highly relevant for improved neutron data evaluation and requested by the OECD/Working Party on Evaluation Cooperation (WPEC). Thirdly, also prompt fission γ-ray emission is investigated using highly efficient lanthanide-halide detectors with superior timing resolution. Again, those data are requested in the HPRL for major actinides to solve open questions on an under-prediction of decay heat in nuclear reactors. The information on prompt fission neutron and γ-ray emission is crucial for benchmarking nuclear models to study the de-excitation process of neutron-rich fission fragments. Information on γ-ray emission probabilities is also useful in decommissioning exercises on damaged nuclear power plants like Fukushima Daiichi to which JRC-IRMM is contributing. The results on the 240,242Pu fission cross section, 235U prompt neutron multiplicity in the resonance region and correlations with fission fragments and prompt γ-ray emission for several isotopes will be presented and put into perspective.

Unit mechanisms of fission gas release: Current understanding and future needs

DOE PAGES

Tonks, Michael; Andersson, David; Devanathan, Ram; ...

2018-03-01

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel and gap properties. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are beginning to reveal new understanding of the unit mechanisms that define fission product behavior. Here, existing research on the basic mechanisms of fission gas release during normal reactor operation are summarized and critical areas where work is needed are identified. Here, this basic understanding of the fission gas behavior mechanisms has the potentialmore » to revolutionize our ability to predict fission product behavior and to design fuels with improved performance. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.« less

Unit mechanisms of fission gas release: Current understanding and future needs

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

Tonks, Michael; Andersson, David; Devanathan, Ram

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel and gap properties. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are beginning to reveal new understanding of the unit mechanisms that define fission product behavior. Here, existing research on the basic mechanisms of fission gas release during normal reactor operation are summarized and critical areas where work is needed are identified. Here, this basic understanding of the fission gas behavior mechanisms has the potentialmore » to revolutionize our ability to predict fission product behavior and to design fuels with improved performance. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.« less

Plant maintenance and advanced reactors, 2005

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

Agnihotri, Newal

2005-09-15

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: First U.S. EPRs in 2015, by Ray Ganthner, Framatome ANP; Pursuing several opportunities, by William E. (Ed) Cummins, Westinghouse Electric Company; Vigorous plans to develop advanced reactors, by Yuliang Sun, Tsinghua University, China; Multiple designs, small and large, by Kumiaki Moriya, Hitachi Ltd., Japan; Sealed and embedded for safety and security, by Handa Norihiko, Toshiba Corporation, Japan; Scheduled online in 2010, by Johan Slabber, PMBR (Pty) Ltd., South Africa; Multi-application reactors, by Nikolay G. Kodochigov, OKBM, Russia; Six projects under budgetmore » and on schedule, by David F. Togerson, AECL, Canada; Creating a positive image, by Scott Peterson, Nuclear Energy Institute (NEI); Advanced plans for nuclear power's renaissance, by John Cleveland, International Atomic Energy Agency, Austria; and, Plant profile: last five outages in less than 20 days, by Beth Rapczynski, Exelon Nuclear.« less

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

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

Harp, Jason Michael; Stempien, John Dennis; Demkowicz, Paul Andrew

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO 2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. Thesemore » data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO 2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO 2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

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

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

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These datamore » were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

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.

Contractile-Ring Assembly in Fission Yeast Cytokinesis: Recent Advances and New Perspectives

PubMed Central

Lee, I-Ju; Coffman, Valerie C.; Wu, Jian-Qiu

2017-01-01

The fission yeast Schizosaccharomyces pombe is an excellent model organism to study cytokinesis. Here, we review recent advances on contractile-ring assembly in fission yeast. First, we summarize the assembly of cytokinesis nodes, the precursors of a normal contractile ring. IQGAP Rng2 and myosin essential light chain Cdc4 are recruited by the anillin-like protein Mid1, followed by the addition of other cytokinesis node proteins. Mid1 localization on the plasma membrane is stabilized by interphase node proteins. Second, we discuss proteins and processes that contribute to the search, capture, pull, and release mechanism of contractile-ring assembly. Actin filaments nucleated by formin Cdc12, the motor activity of myosin-II, the stiffness of the actin network, and severing of actin filaments by cofilin all play essential roles in contractile-ring assembly. Finally, we discuss the Mid1-independent pathway for ring assembly, and the possible mechanisms underlying the ring maturation and constriction. Collectively, we provide an overview of the current understanding of contractile-ring assembly and uncover future directions in studying cytokinesis in fission yeast. PMID:22887981

A Practical Approach to Starting Fission Surface Power Development

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2006-01-01

The Prometheus Power and Propulsion Program has been reformulated to address NASA needs relative to lunar and Mars exploration. Emphasis has switched from the Jupiter Icy Moons Orbiter (JIMO) flight system development to more generalized technology development addressing Fission Surface Power (FSP) and Nuclear Thermal Propulsion (NTP). Current NASA budget priorities and the deferred mission need date for nuclear systems prohibit a fully funded reactor Flight Development Program. However, a modestly funded Advanced Technology Program can and should be conducted to reduce the risk and cost of future flight systems. A potential roadmap for FSP technology development leading to possible flight applications could include three elements: 1) Conceptual Design Studies, 2) Advanced Component Technology, and 3) Non-Nuclear System Testing. The Conceptual Design Studies would expand on recent NASA and DOE analyses while increasing the depth of study in areas of greatest uncertainty such as reactor integration and human-rated shielding. The Advanced Component Technology element would address the major technology risks through development and testing of reactor fuels, structural materials, primary loop components, shielding, power conversion, heat rejection, and power management and distribution (PMAD). The Non-Nuclear System Testing would provide a modular, technology testbed to investigate and resolve system integration issues.

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

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

M. Chen; CM Regan; D. Noe

2006-01-09

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

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

NASA Astrophysics Data System (ADS)

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

2017-08-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

Comparing the new generation accelerator driven subcritical reactor system (ADS) to traditional critical reactors

NASA Astrophysics Data System (ADS)

Kemah, Elif; Akkaya, Recep; Tokgöz, Seyit Rıza

2017-02-01

In recent years, the accelerator driven subcritical reactors have taken great interest worldwide. The Accelerator Driven System (ADS) has been used to produce neutron in subcritical state by the external proton beam source. These reactors, which are hybrid systems, are important in production of clean and safe energy and conversion of radioactive waste. The ADS with the selection of reliability and robust target materials have been the new generation of fission reactors. In addition, in the ADS Reactors the problems of long-lived radioactive fission products and waste actinides encountered in the fission process of the reactor during incineration can be solved, and ADS has come to the forefront of thorium as fuel for the reactors.

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

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

Behafarid, F.; Shaver, D. R.; Bolotnov, I. A.

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 approachmore » 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)« less

Student Experiments in Spontaneous Fission.

ERIC Educational Resources Information Center

Becchetti, F. D.; Ying, J. S.

1981-01-01

Advanced undergraduate experiments utilizing a commercially available, thin spontaneous fission source are described, including studies of the energy and mass distribution of the fission fragments and their energy and angular correlation. The experiments provide a useful introduction to fission, nuclear mass equations, heavy-ion physics, and…

Cladding and duct materials for advanced nuclear recycle reactors

NASA Astrophysics Data System (ADS)

Allen, T. R.; Busby, J. T.; Klueh, R. L.; Maloy, S. A.; Toloczko, M. B.

2008-01-01

The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP’s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP’s advanced nuclear recycle reactors program.

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

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

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 transmutemore » 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.« less

Advanced Instrumentation for Transient Reactor Testing

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

Corradini, Michael L.; Anderson, Mark; Imel, George

Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and designmore » increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux

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

Contractile-ring assembly in fission yeast cytokinesis: Recent advances and new perspectives.

PubMed

Lee, I-Ju; Coffman, Valerie C; Wu, Jian-Qiu

2012-10-01

The fission yeast Schizosaccharomyces pombe is an excellent model organism to study cytokinesis. Here, we review recent advances on contractile-ring assembly in fission yeast. First, we summarize the assembly of cytokinesis nodes, the precursors of a normal contractile ring. IQGAP Rng2 and myosin essential light chain Cdc4 are recruited by the anillin-like protein Mid1, followed by the addition of other cytokinesis node proteins. Mid1 localization on the plasma membrane is stabilized by interphase node proteins. Second, we discuss proteins and processes that contribute to the search, capture, pull, and release mechanism of contractile-ring assembly. Actin filaments nucleated by formin Cdc12, the motor activity of myosin-II, the stiffness of the actin network, and severing of actin filaments by cofilin all play essential roles in contractile-ring assembly. Finally, we discuss the Mid1-independent pathway for ring assembly, and the possible mechanisms underlying the ring maturation and constriction. Collectively, we provide an overview of the current understanding of contractile-ring assembly and uncover future directions in studying cytokinesis in fission yeast. Copyright © 2012 Wiley Periodicals, Inc.

Results from the DOE Advanced Gas Reactor Fuel Development and Qualification Program

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

David Petti

2014-06-01

Modular HTGR designs were developed to provide natural safety, which prevents core damage under all design basis accidents and presently envisioned severe accidents. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude and allows potential elimination of the need for evacuation and sheltering beyond a small exclusion area. This level, however, is predicated on exceptionally high fuel fabrication quality and performance under normal operation and accident conditions. Germanymore » produced and demonstrated high quality fuel for their pebble bed HTGRs in the 1980s, but no U.S. manufactured fuel had exhibited equivalent performance prior to the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The design goal of the modular HTGRs is to allow elimination of an exclusion zone and an emergency planning zone outside the plant boundary fence, typically interpreted as being about 400 meters from the reactor. To achieve this, the reactor design concepts require a level of fuel integrity that is better than that claimed for all prior US manufactured TRISO fuel, by a few orders of magnitude. The improved performance level is about a factor of three better than qualified for German TRISO fuel in the 1980’s. At the start of the AGR program, without a reactor design concept selected, the AGR fuel program selected to qualify fuel to an operating envelope that would bound both pebble bed and prismatic options. This resulted in needing a fuel form that could survive at peak fuel temperatures of 1250°C on a time-averaged basis and high burnups in the range of 150 to 200 GWd/MTHM (metric tons of heavy metal) or 16.4 to 21.8% fissions per initial metal atom (FIMA). Although Germany has demonstrated excellent performance of TRISO

NEUTRONIC REACTOR SYSTEM

DOEpatents

Goett, J.J.

1961-01-24

A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

Use of freeze-casting in advanced burner reactor fuel design

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

Lang, A. L.; Yablinsky, C. A.; Allen, T. R.

2012-07-01

This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by thatmore » fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models

Kilopower: Small and Affordable Fission Power Systems for Space

NASA Technical Reports Server (NTRS)

Mason, Lee; Palac, Don; Gibson, Marc

2017-01-01

The Nuclear Systems Kilopower Project was initiated by NASA's Space Technology Mission Directorate Game Changing Development Program in fiscal year 2015 to demonstrate subsystem-level technology readiness of small space fission power in a relevant environment (Technology Readiness Level 5) for space science and human exploration power needs. The Nuclear Systems Kilopower Project centerpiece is the Kilopower Reactor Using Stirling Technology (KRUSTY) test, which consists of the development and testing of a fission ground technology demonstrator of a 1 kWe-class fission power system. The technologies to be developed and validated by KRUSTY are extensible to space fission power systems from 1 to 10 kWe, which can enable higher power future potential deep space science missions, as well as modular surface fission power systems for exploration. The Kilopower Project is cofounded by NASA and the Department of Energy National Nuclear Security Administration (NNSA).KRUSTY include the reactor core, heat pipes to transfer the heat from the core to the power conversion system, and the power conversion system. Los Alamos National Laboratory leads the design of the reactor, and the Y-12 National Security Complex is fabricating it. NASA Glenn Research Center (GRC) has designed, built, and demonstrated the balance of plant heat transfer and power conversion portions of the KRUSTY experiment. NASA MSFC developed an electrical reactor simulator for non-nuclear testing, and the design of the reflector and shielding for nuclear testing. In 2016, an electrically heated non-fissionable Depleted Uranium (DU) core was tested at GRC in a configuration identical to the planned nuclear test. Once the reactor core has been fabricated and shipped to the Device Assembly Facility at the NNSAs Nevada National Security Site, the KRUSTY nuclear experiment will be assembled and tested. Completion of the KRUSTY experiment will validate the readiness of 1 to 10 kWe space fission technology for NASAs

Big data mining powers fungal research: recent advances in fission yeast systems biology approaches.

PubMed

Wang, Zhe

2017-06-01

Biology research has entered into big data era. Systems biology approaches therefore become the powerful tools to obtain the whole landscape of how cell separate, grow, and resist the stresses. Fission yeast Schizosaccharomyces pombe is wonderful unicellular eukaryote model, especially studying its division and metabolism can facilitate to understanding the molecular mechanism of cancer and discovering anticancer agents. In this perspective, we discuss the recent advanced fission yeast systems biology tools, mainly focus on metabolomics profiling and metabolic modeling, protein-protein interactome and genetic interaction network, DNA sequencing and applications, and high-throughput phenotypic screening. We therefore hope this review can be useful for interested fungal researchers as well as bioformaticians.

Assessment of Sensor Technologies for Advanced Reactors

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

Korsah, Kofi; Kisner, R. A.; Britton Jr., C. L.

This paper provides an assessment of sensor technologies and a determination of measurement needs for advanced reactors (AdvRx). It is a summary of a study performed to provide the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program. The study covered two broad reactor technology categories: High Temperature Reactors and Fast Reactors. The scope of “High temperature reactors” included Gen IV reactors whose coolant exit temperatures exceed ≈650 °C and are moderated (as opposed to fast reactors). To bound the scope formore » fast reactors, this report reviewed relevant operating experience from US-operated Sodium Fast Reactor (SFR) and relevant test experience from the Fast Flux Test Facility (FFTF). For high temperature reactors the study showed that in many cases instrumentation have performed reasonably well in research and demonstration reactors. However, even in cases where the technology is “mature” (such as thermocouples), HTGRs can benefit from improved technologies. Current HTGR instrumentation is generally based on decades-old technology and adapting newer technologies could provide significant advantages. For sodium fast reactors, the study found that several key research needs arise around (1) radiation-tolerant sensor design for in-vessel or in-core applications, where possible non-invasive sensing approaches for key parameters that minimize the need to deploy sensors in-vessel, (2) approaches to exfiltrating data from in-vessel sensors while minimizing penetrations, (3) calibration of sensors in-situ, and (4) optimizing sensor placements to maximize the information content while minimizing the number of sensors needed.« less

Neutron induced fission of 237Np - status, challenges and opportunities

NASA Astrophysics Data System (ADS)

Ruskov, Ivan; Goverdovski, Andrei; Furman, Walter; Kopatch, Yury; Shcherbakov, Oleg; Hambsch, Franz-Josef; Oberstedt, Stephan; Oberstedt, Andreas

2018-03-01

Nowadays, there is an increased interest in a complete study of the neutron-induced fission of 237Np. This is due to the need of accurate and reliable nuclear data for nuclear science and technology. 237Np is generated (and accumulated) in the nuclear reactor core during reactor operation. As one of the most abundant long-lived isotopes in spent fuel ("waste"), the incineration of 237Np becomes an important issue. One scenario for burning of 237Np and other radio-toxic minor actinides suggests they are to be mixed into the fuel of future fast-neutron reactors, employing the so-called transmutation and partitioning technology. For testing present fission models, which are at the basis of new generation nuclear reactor developments, highly accurate and detailed neutron-induced nuclear reaction data is needed. However, the EXFOR nuclear database for 237Np on neutron-induced capture cross-section, σγ, and fission cross-section, σf, as well as on the characteristics of capture and fission resonance parameters (Γγ, Γf, σoΓf, fragments mass-energy yield distributions, multiplicities of neutrons vn and γ-rays vγ), has not been updated for decades.

REACTOR

DOEpatents

Christy, R.F.

1961-07-25

A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

Modification of apparent fission yields by Chemical Fractionation following Fission (CFF)

NASA Astrophysics Data System (ADS)

Hohenberg, Charles; Meshik, Alex

2008-04-01

Grain-by-grain studies of the 2 billion year old Oklo natural reactor, using laser micro-extraction^1,2, yield detailed information about Oklo, a water-moderated pulsed reactor, cycle times, total neutron fluence and duration, but it also demonstrates Chemical Fractionation following Fission. In the CFF process, members of an isobaric yield chain with long half-lives are subject to migration before decay can occur. Of particular interest is the 129 isobar where 17 million ^129I can migrate out of the host grain before decay, and iodine compounds are water soluble. This is amply demonstated by the variation of Xe spectra between micron-sized uranium-bearing minerals and adjacent uranium-free minerals. Fission 129 yields for the spontaneous fission of ^238U generally come from measured ^129Xe in pitchblend^2, ores emplaced by aqueous activity, and are incorrect due to the CFF process. ^238U yields for the 131 and 129 chains, reported in Hyde^3, as 0.455 +- .02 and < 0.012, respectively, the latter being anomalously low. ^1A Meshik, C Hohenberg and O Pravdivtesva, PRL 93, 182302 (2004); A Meshik Sci. Am. Nov (2005), 55; ^2E K Hyde, Nucl Prop of Heavy Elements III (1964).

Research on stellarator-mirror fission-fusion hybrid

NASA Astrophysics Data System (ADS)

Moiseenko, V. E.; Kotenko, V. G.; Chernitskiy, S. V.; Nemov, V. V.; Ågren, O.; Noack, K.; Kalyuzhnyi, V. N.; Hagnestål, A.; Källne, J.; Voitsenya, V. S.; Garkusha, I. E.

2014-09-01

The development of a stellarator-mirror fission-fusion hybrid concept is reviewed. The hybrid comprises of a fusion neutron source and a powerful sub-critical fast fission reactor core. The aim is the transmutation of spent nuclear fuel and safe fission energy production. In its fusion part, neutrons are generated in deuterium-tritium (D-T) plasma, confined magnetically in a stellarator-type system with an embedded magnetic mirror. Based on kinetic calculations, the energy balance for such a system is analyzed. Neutron calculations have been performed with the MCNPX code, and the principal design of the reactor part is developed. Neutron outflux at different outer parts of the reactor is calculated. Numerical simulations have been performed on the structure of a magnetic field in a model of the stellarator-mirror device, and that is achieved by switching off one or two coils of toroidal field in the Uragan-2M torsatron. The calculations predict the existence of closed magnetic surfaces under certain conditions. The confinement of fast particles in such a magnetic trap is analyzed.

The Sustainable Nuclear Future: Fission and Fusion E.M. Campbell Logos Technologies

NASA Astrophysics Data System (ADS)

Campbell, E. Michael

2010-02-01

Global industrialization, the concern over rising CO2 levels in the atmosphere and other negative environmental effects due to the burning of hydrocarbon fuels and the need to insulate the cost of energy from fuel price volatility have led to a renewed interest in nuclear power. Many of the plants under construction are similar to the existing light water reactors but incorporate modern engineering and enhanced safety features. These reactors, while mature, safe and reliable sources of electrical power have limited efficiency in converting fission power to useful work, require significant amounts of water, and must deal with the issues of nuclear waste (spent fuel), safety, and weapons proliferation. If nuclear power is to sustain its present share of the world's growing energy needs let alone displace carbon based fuels, more than 1000 reactors will be needed by mid century. For this to occur new reactors that are more efficient, versatile in their energy markets, require minimal or no water, produce less waste and more robust waste forms, are inherently safe and minimize proliferation concerns will be necessary. Graphite moderated, ceramic coated fuel, and He cooled designs are reactors that can satisfy these requirements. Along with other generation IV fast reactors that can further reduce the amounts of spent fuel and extend fuel resources, such a nuclear expansion is possible. Furthermore, facilities either in early operations or under construction should demonstrate the next step in fusion energy development in which energy gain is produced. This demonstration will catalyze fusion energy development and lead to the ultimate development of the next generation of nuclear reactors. In this presentation the role of advanced fission reactors and future fusion reactors in the expansion of nuclear power will be discussed including synergies with the existing worldwide nuclear fleet. )

Method to Reduce Long-lived Fission Products by Nuclear Transmutations with Fast Spectrum Reactors.

PubMed

Chiba, Satoshi; Wakabayashi, Toshio; Tachi, Yoshiaki; Takaki, Naoyuki; Terashima, Atsunori; Okumura, Shin; Yoshida, Tadashi

2017-10-24

Transmutation of long-lived fission products (LLFPs: 79 Se, 93 Zr, 99 Tc, 107 Pd, 129 I, and 135 Cs) into short-lived or non-radioactive nuclides by fast neutron spectrum reactors without isotope separation has been proposed as a solution to the problem of radioactive wastes disposal. Despite investigation of many methods, such transmutation remains technologically difficult. To establish an effective and efficient transmutation system, we propose a novel neutron moderator material, yttrium deuteride (YD 2 ), to soften the neutron spectrum leaking from the reactor core. Neutron energy spectra and effective half-lives of LLFPs, transmutation rates, and support ratios were evaluated with the continuous-energy Monte Carlo code MVP-II/MVP-BURN and the JENDL-4.0 cross section library. With the YD 2 moderator in the radial blanket and shield regions, effective half-lives drastically decreased from 106 to 102 years and the support ratios reached 1.0 for all six LLFPs. This successful development and implementation of a transmutation system for LLFPs without isotope separation contributes to a the ability of fast spectrum reactors to reduce radioactive waste by consuming their own LLFPs.

FCRD Advanced Reactor (Transmutation) Fuels Handbook

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

Janney, Dawn Elizabeth; Papesch, Cynthia Ann

2016-09-01

Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. U-Pu-Zr alloys are well suited for electrolytic refining, which leads to incorporation rare-earth fission products such as La, Ce, Pr, and Nd. It is, therefore, importantmore » to understand not only the properties of U-Pu-Zr alloys but also those of U-Pu-Zr alloys with concentrations of minor actinides (Np, Am) and rare-earth elements (La, Ce, Pr, and Nd) similar to those in reprocessed fuel. In addition to requiring extensive safety precautions, alloys containing U, Pu, and minor actinides (Np and Am) are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phasetransformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, rapid oxidation, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Although less toxic, rare-earth elements such as La, Ce, Pr, and Nd are also difficult to study for similar reasons. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, particularly those that also contain minor actinides and rare-earth elements. General acceptance of results commonly indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, Np, Am, La, Ce, Pr, and

Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

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

James E. O'Brien; Piyush Sabharwall; SuJong Yoon

2001-11-01

Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water.more » Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed« less

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

NASA Astrophysics Data System (ADS)

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

2017-08-01

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

Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

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

Loflin, Leonard; McRimmon, Beth

2014-12-18

This report summarizes a project by EPRI to include requirements for small modular light water reactors (smLWR) into the EPRI Utility Requirements Document (URD) for Advanced Light Water Reactors. The project was jointly funded by EPRI and the U.S. Department of Energy (DOE). The report covers the scope and content of the URD, the process used to revise the URD to include smLWR requirements, a summary of the major changes to the URD to include smLWR, and how to use the URD as revised to achieve value on new plant projects.

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.

Comparison of Fission Product Yields and Their Impact

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

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 radiologicalmore » transmutation of krypton-85 in the coolant to rubidium-85, a potentially corrosive agent to the coolant system metal piping.« less

Studies of Fission Fragment Rocket Engine Propelled Spacecraft

NASA Technical Reports Server (NTRS)

Werka, Robert O.; Clark, Rodney; Sheldon, Rob; Percy, Thomas K.

2014-01-01

The NASA Office of Chief Technologist has funded from FY11 through FY14 successive studies of the physics, design, and spacecraft integration of a Fission Fragment Rocket Engine (FFRE) that directly converts the momentum of fission fragments continuously into spacecraft momentum at a theoretical specific impulse above one million seconds. While others have promised future propulsion advances if only you have the patience, the FFRE requires no waiting, no advances in physics and no advances in manufacturing processes. Such an engine unequivocally can create a new era of space exploration that can change spacecraft operation. The NIAC (NASA Institute for Advanced Concepts) Program Phase 1 study of FY11 first investigated how the revolutionary FFRE technology could be integrated into an advanced spacecraft. The FFRE combines existent technologies of low density fissioning dust trapped electrostatically and high field strength superconducting magnets for beam management. By organizing the nuclear core material to permit sufficient mean free path for escape of the fission fragments and by collimating the beam, this study showed the FFRE could convert nuclear power to thrust directly and efficiently at a delivered specific impulse of 527,000 seconds. The FY13 study showed that, without increasing the reactor power, adding a neutral gas to the fission fragment beam significantly increased the FFRE thrust through in a manner analogous to a jet engine afterburner. This frictional interaction of gas and beam resulted in an engine that continuously produced 1000 pound force of thrust at a delivered impulse of 32,000 seconds, thereby reducing the currently studied DRM 5 round trip mission to Mars from 3 years to 260 days. By decreasing the gas addition, this same engine can be tailored for much lower thrust at much higher impulse to match missions to more distant destinations. These studies created host spacecraft concepts configured for manned round trip journeys. While the

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.

Effects of fission yield data in the calculation of antineutrino spectra for U 235 ( n , fission ) at thermal and fast neutron energies

DOE PAGES

Sonzogni, A. A.; McCutchan, E. A.; Johnson, T. D.; ...

2016-04-01

Fission yields form an integral part of the prediction of antineutrino spectra generated by nuclear reactors, but little attention has been paid to the quality and reliability of the data used in current calculations. Following a critical review of the thermal and fast ENDF/B-VII.1 235U fission yields, deficiencies are identified and improved yields are obtained, based on corrections of erroneous yields, consistency between decay and fission yield data, and updated isomeric ratios. These corrected yields are used to calculate antineutrino spectra using the summation method. An anomalous value for the thermal fission yield of 86Ge generates an excess of antineutrinosmore » at 5–7 MeV, a feature which is no longer present when the corrected yields are used. Thermal spectra calculated with two distinct fission yield libraries (corrected ENDF/B and JEFF) differ by up to 6% in the 0–7 MeV energy window, allowing for a basic estimate of the uncertainty involved in the fission yield component of summation calculations. Lastly, the fast neutron antineutrino spectrum is calculated, which at the moment can only be obtained with the summation method and may be relevant for short baseline reactor experiments using highly enriched uranium fuel.« less

Testing in Support of Space Fission System Development and Qualification

NASA Technical Reports Server (NTRS)

Houts, Mike; Bragg-Sitton, Shannon; Garber, Anne; Godfrey, Tom; Martin, Jim; Pearson, Boise; Webster, Kenny

2007-01-01

Extensive data would be required for the qualification of a fission surface power (FSP) system. The strategy for qualifying a FSP system could have a significant programmatic impact. This paper explores potential options that could be used for qualifying FSP systems, including cost-effective means for obtaining required data. three methods for obtaining qualification data are analysis, non-nuclear testing, and nuclear testing. It has been over 40 years since the US qualified a space reactor for launch. During that time, advances have been made related to all three methods. Perhaps the greatest advancement has occurred in the area of computational tools for design and analysis. Tools that have been developed, coupled with modem computers, would have a significant impact on a FSP qualification. This would be especially true for systems with materials and fuels operating well within temperature, irradiation damage, and burnup limits. The ability to perform highly realistic non-nuclear testing has also advanced throughout the past four decades. Instrumented thermal simulators were developed during the 1970s and 1980s to assist in the development, operation, and assessment of terrestrial fission systems. Instrumented thermal simulators optimized for assisting in the development, operation, and assessment of modem FSP systems have been under development (and utilized) since 1998. These thermal simulators enable heat from fission to be closely mimicked (axial power profile, radial power profile, temperature, heat flux, etc.} and extensive data to be taken from the core region. Both steady-state and transient operation can be tested. For transient testing, reactivity feedback is calculated (or measured in cold/warm criticals) based on reactor temperature and/or dimensional changes. Pin power during a transient is then calculated based on the reactivity feedback that would occur given measured values of temperature and/or dimensional change. In this way nonnuclear testing

Advanced Reactor Technology/Energy Conversion Project FY17 Accomplishments.

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

Rochau, Gary E.

The purpose of the ART Energy Conversion (EC) Project is to provide solutions to convert the heat from an advanced reactor to useful products that support commercial application of the reactor designs.

Advanced Plasma Pyrolysis Assembly (PPA) Reactor and Process Development

NASA Technical Reports Server (NTRS)

Wheeler, Richard R., Jr.; Hadley, Neal M.; Dahl, Roger W.; Abney, Morgan B.; Greenwood, Zachary; Miller, Lee; Medlen, Amber

2012-01-01

Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA's Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development.

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

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

Belles, Randy; Poore, III, Willis P.; Brown, Nicholas R.

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-basedmore » description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.« less

Johnson Noise Thermometry for Advanced Small Modular Reactors

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

Britton, C.L.,Jr.; Roberts, M.; Bull, N.D.

Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor’s physical condition. In and near the core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurementmore » due to their fundamental natures. Small Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of the current ORNL-led project, conducted under the Instrumentation, Controls, and Human-Machine Interface (ICHMI) research pathway of the U.S. Department of Energy (DOE) Advanced SMR Research and Development (R&D) program, is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.« less

System Concepts for Affordable Fission Surface Power

NASA Technical Reports Server (NTRS)

Mason, Lee; Poston, David; Qualls, Louis

2008-01-01

This paper presents an overview of an affordable Fission Surface Power (FSP) system that could be used for NASA applications on the Moon and Mars. The proposed FSP system uses a low temperature, uranium dioxide-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The concept was determined by a 12 month NASA/DOE study that examined design options and development strategies based on affordability and risk. The system is considered a low development risk based on the use of terrestrial-derived reactor technology, high efficiency power conversion, and conventional materials. The low-risk approach was selected over other options that could offer higher performance and/or lower mass.

Fission gas release during power bumping at high burnup

NASA Astrophysics Data System (ADS)

Cunningham, M. E.; Freshley, M. D.; Lanning, D. D.

1993-03-01

Research to define the behavior of Zircaloy-clad light-water reactor fuel irradiated to high burnup levels was conducted by the High Burnup Effects Program (HBEP). One activity conducted by the HBEP was to "bump" the power level of irradiated, commercial light-water reactor fuel rods to design limit linear heat generation rates at end-of-life. These bumping irradiations simulated end-of-life design limit linear heat generation rates and provided data on the effects of short-term, high power irradiations at high burnup applicable to the design and operating constraints imposed by maximum allowable fuel rod internal gas pressure limits. Based on net fission gas release during the bumping irradiations, it was observed that higher burnup rods had greater rod-average fractional fission gas release than lower burnup rods at equal bumping powers. It was also observed that a hold period of 48 hours at the peak power was insufficient to achieve equilibrium fission gas release. Finally, differences in the prebump location of fission gas, i.e., within the UO 2 matrix or at grain boundaries, affected the fission gas release during the bumping irradiations.

World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

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

Not Available

1979-06-01

Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment. (JWR)

Code qualification of structural materials for AFCI advanced recycling reactors.

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

Natesan, K.; Li, M.; Majumdar, S.

2012-05-31

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Codemore » Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP

Fission Surface Power Technology Demonstration Unit

NASA Image and Video Library

2016-11-09

NASA Glenn Technician Mark Springowski works on a 10-kilowatt Stirling Power Conversion Unit, which is part of the Fission Surface Power Technology Demonstration Unit. This is a system level demonstration of a surface power system, which could potentially be used to support manned missions to the moon or Mars. A flight system would use 180 kilowatt nuclear fission reactor and four Stirling PCU’s to produce 40 kW of electricity for manned surface missions.

Coincident measurements of prompt fission γ rays and fission fragments at DANCE

NASA Astrophysics Data System (ADS)

Walker, C. L.; Baramsai, B.; Jandel, M.; Rusev, G.; Couture, A.; Mosby, S.; Ullmann, J.; Kawano, T.; Stetcu, I.; Talou, P.

2015-10-01

Modern statistical approaches to modeling fission involve the calculation of not only average quantities but also fully correlated distributions of all fission products. Applications such as those involving the detection of special nuclear materials also rely on fully correlated data of fission products. Experimental measurements of correlated data are thus critical to the validation of theory and the development of important applications. The goal of this experiment was to measure properties of prompt fission gamma-ray emission as a function of fission fragments' total kinetic energy in the spontaneous fission of 252Cf. The measurement was carried out at the Detector for Advanced Neutron Capture Experiments (DANCE), a 4 π γ-ray calorimeter. A prototype design consisting of two silicon detectors was installed in the center of DANCE, allowing simultaneous measurement of fission fragments and γ rays. Effort has been taken to simulate fragment kinetic energy losses as well as γ-ray attenuation in DANCE using such tools as GEANT4 and SRIM. Theoretical predictions generated by the code CGMF were also incorporated as input for these simulations. Results from the experiment and simulations will be presented, along with plans for future measurements.

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

NASA Technical Reports Server (NTRS)

Palac, Donald T.

2011-01-01

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

Electroplating method for producing ultralow-mass fissionable deposits

DOEpatents

Ruddy, Francis H.

1989-01-01

A method for producing ultralow-mass fissionable deposits for nuclear reactor dosimetry is described, including the steps of holding a radioactive parent until the radioactive parent reaches secular equilibrium with a daughter isotope, chemically separating the daughter from the parent, electroplating the daughter on a suitable substrate, and holding the electroplated daughter until the daughter decays to the fissionable deposit.

Advanced reactor design study. Assessing nonbackfittable concepts for improving uranium utilization in light water reactors

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

Fleischman, R.M.; Goldsmith, S.; Newman, D.F.

1981-09-01

The objective of the Advanced Reactor Design Study (ARDS) is to identify and evaluate nonbackfittable concepts for improving uranium utilization in light water reactors (LWRs). The results of this study provide a basis for selecting and demonstrating specific nonbackfittable concepts that have good potential for implementation. Lead responsibility for managing the study was assigned to the Pacific Northwest Laboratory (PNL). Nonbackfittable concepts for improving uranium utilization in LWRs on the once-through fuel cycle were selected separately for PWRs and BWRs due to basic differences in the way specific concepts apply to those plants. Nonbackfittable concepts are those that are toomore » costly to incorporate in existing plants, and thus, could only be economically incorporated in new reactor designs or plants in very early stages of construction. Essential results of the Advanced Reactor Design Study are summarized.« less

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.

Assessment of Sensor Technologies for Advanced Reactors

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

Korsah, Kofi; Ramuhalli, Pradeep; Vlim, R.

2016-10-01

Sensors and measurement technologies provide information on processes, support operations and provide indications of component health. They are therefore crucial to plant operations and to commercialization of advanced reactors (AdvRx). This report, developed by a three-laboratory team consisting of Argonne National Laboratory (ANL), Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL), provides an assessment of sensor technologies and a determination of measurement needs for AdvRx. It provides the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program and contributesmore » to the design and implementation of AdvRx concepts.« less

Johnson Noise Thermometry for Advanced Small Modular Reactors

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

Britton Jr, Charles L; Roberts, Michael; Bull, Nora D

Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor s physical condition. In and near core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurementmore » due to their fundamental natures. Small, Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of this project is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.« less

NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

DOEpatents

Moore, W.T.

1958-09-01

This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

Eugene P. Wigner's Visionary Contributions to Generations-I through IV Fission Reactors

NASA Astrophysics Data System (ADS)

Carré, Frank

2014-09-01

Among Europe's greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechanics, Solid state physics and other topics that opened new branches of Physics.

The Munich accelerator for fission fragments MAFF

NASA Astrophysics Data System (ADS)

Habs, D.; Groß, M.; Assmann, W.; Ames, F.; Bongers, H.; Emhofer, S.; Heinz, S.; Henry, S.; Kester, O.; Neumayr, J.; Ospald, F.; Reiter, P.; Sieber, T.; Szerypo, J.; Thirolf, P. G.; Varentsov, V.; Wilfart, T.; Faestermann, T.; Krücken, R.; Maier-Komor, P.

2003-05-01

The Munich Accelerator for Fission Fragments MAFF has been designed for the new Munich research reactor FRM-II. It will deliver several intense beams (˜3×10 11 s -1) of very neutron-rich fission fragments with a final energy of 30 keV (low-energy beam) or energies between 3.7 and 5.9 MeV· A (high-energy beam). Such beams are of interest for the creation of super-heavy elements by fusion reactions, nuclear spectroscopy of exotic nuclei, but they also have a potential for applications, e.g. in medicine. Presently the Munich research reactor FRM-II is ready for operation, but authorities delay the final permission to turn the reactor critical probably till the end of 2002. Only after this final permission the financing of the major parts of MAFF can start. On the other hand all major components have been designed and special components have been tested in separate setups.

Nanocrystalline SiC and Ti 3SiC 2 Alloys for Reactor Materials: Diffusion of Fission Product Surrogates

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

Henager, Charles H.; Jiang, Weilin

2014-11-01

MAX phases, such as titanium silicon carbide (Ti 3SiC 2), have a unique combination of both metallic and ceramic properties, which make them attractive for potential nuclear applications. Ti 3SiC 2 has been suggested in the literature as a possible fuel cladding material. Prior to the application, it is necessary to investigate diffusivities of fission products in the ternary compound at elevated temperatures. This study attempts to obtain relevant data and make an initial assessment for Ti 3SiC 2. Ion implantation was used to introduce fission product surrogates (Ag and Cs) and a noble metal (Au) in Ti 3SiC 2,more » SiC, and a dual-phase nanocomposite of Ti 3SiC 2/SiC synthesized at PNNL. Thermal annealing and in-situ Rutherford backscattering spectrometry (RBS) were employed to study the diffusivity of the various implanted species in the materials. In-situ RBS study of Ti 3SiC 2 implanted with Au ions at various temperatures was also performed. The experimental results indicate that the implanted Ag in SiC is immobile up to the highest temperature (1273 K) applied in this study; in contrast, significant out-diffusion of both Ag and Au in MAX phase Ti 3SiC 2 occurs during ion implantation at 873 K. Cs in Ti 3SiC 2 is found to diffuse during post-irradiation annealing at 973 K, and noticeable Cs release from the sample is observed. This study may suggest caution in using Ti 3SiC 2 as a fuel cladding material for advanced nuclear reactors operating at very high temperatures. Further studies of the related materials are recommended.« less

ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA660, INTERIOR. REACTOR INSIDE TANK. METAL ...

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

ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA-660, INTERIOR. REACTOR INSIDE TANK. METAL WORK PLATFORM ABOVE. THE REACTOR WAS IN A SMALL WATER-FILLED POOL. INL NEGATIVE NO. 66-6373. Unknown Photographer, ca. 1966 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Advanced Small Modular Reactor Economics Status Report

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

Harrison, Thomas J.

2014-10-01

This report describes the data collection work performed for an advanced small modular reactor (AdvSMR) economics analysis activity at the Oak Ridge National Laboratory. The methodology development and analytical results are described in separate, stand-alone documents as listed in the references. The economics analysis effort for the AdvSMR program combines the technical and fuel cycle aspects of advanced (non-light water reactor [LWR]) reactors with the market and production aspects of SMRs. This requires the collection, analysis, and synthesis of multiple unrelated and potentially high-uncertainty data sets from a wide range of data sources. Further, the nature of both economic andmore » nuclear technology analysis requires at least a minor attempt at prediction and prognostication, and the far-term horizon for deployment of advanced nuclear systems introduces more uncertainty. Energy market uncertainty, especially the electricity market, is the result of the integration of commodity prices, demand fluctuation, and generation competition, as easily seen in deregulated markets. Depending on current or projected values for any of these factors, the economic attractiveness of any power plant construction project can change yearly or quarterly. For long-lead construction projects such as nuclear power plants, this uncertainty generates an implied and inherent risk for potential nuclear power plant owners and operators. The uncertainty in nuclear reactor and fuel cycle costs is in some respects better understood and quantified than the energy market uncertainty. The LWR-based fuel cycle has a long commercial history to use as its basis for cost estimation, and the current activities in LWR construction provide a reliable baseline for estimates for similar efforts. However, for advanced systems, the estimates and their associated uncertainties are based on forward-looking assumptions for performance after the system has been built and has achieved commercial

Safe Affordable Fission Engine-(SAFE-) 100a Heat Exchanger Thermal and Structural Analysis

NASA Technical Reports Server (NTRS)

Steeve, B. E.

2005-01-01

A potential fission power system for in-space missions is a heat pipe-cooled reactor coupled to a Brayton cycle. In this system, a heat exchanger (HX) transfers the heat of the reactor core to the Brayton gas. The Safe Affordable Fission Engine- (SAFE-) 100a is a test program designed to thermally and hydraulically simulate a 95 Btu/s prototypic heat pipe-cooled reactor using electrical resistance heaters on the ground. This Technical Memorandum documents the thermal and structural assessment of the HX used in the SAFE-100a program.

Design of a Mechanical NaK Pump for Fission Space Power

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Bradley, David E.; Godfroy, Thomas

2011-01-01

Alkali liquid metal cooled fission reactor concepts are under development for spaceflight power requirements. One such concept utilizes a sodium-potassium eutectic (NaK) as the primary loop working fluid, which has specific pumping requirements. Traditionally, electromagnetic linear induction pumps have been used to provide the required flow and pressure head conditions for NaK systems but they can be limited in performance, efficiency, and number of available vendors. The objective of the project was to develop a mechanical NaK centrifugal pump that takes advantages of technology advances not available in previous liquid metal mechanical pump designs. This paper details the design, build, and performance test of a mechanical NaK pump developed at NASA Marshall Space Flight Center. The pump was designed to meet reactor cooling requirements using commercially available components modified for high temperature NaK service.

Isotopic composition and neutronics of the Okelobondo natural reactor

NASA Astrophysics Data System (ADS)

Palenik, Christopher Samuel

The Oklo-Okelobondo and Bangombe uranium deposits, in Gabon, Africa host Earth's only known natural nuclear fission reactors. These 2 billion year old reactors represent a unique opportunity to study used nuclear fuel over geologic periods of time. The reactors in these deposits have been studied as a means by which to constrain the source term of fission product concentrations produced during reactor operation. The source term depends on the neutronic parameters, which include reactor operation duration, neutron flux and the neutron energy spectrum. Reactor operation has been modeled using a point-source computer simulation (Oak Ridge Isotope Generation and Depletion, ORIGEN, code) for a light water reactor. Model results have been constrained using secondary ionization mass spectroscopy (SIMS) isotopic measurements of the fission products Nd and Te, as well as U in uraninite from samples collected in the Okelobondo reactor zone. Based upon the constraints on the operating conditions, the pre-reactor concentrations of Nd (150 ppm +/- 75 ppm) and Te (<1 ppm) in uraninite were estimated. Related to the burnup measured in Okelobondo samples (0.7 to 13.8 GWd/MTU), the final fission product inventories of Nd (90 to 1200 ppm) and Te (10 to 110 ppm) were calculated. By the same means, the ranges of all other fission products and actinides produced during reactor operation were calculated as a function of burnup. These results provide a source term against which the present elemental and decay abundances at the fission reactor can be compared. Furthermore, they provide new insights into the extent to which a "fossil" nuclear reactor can be characterized on the basis of its isotopic signatures. In addition, results from the study of two other natural systems related to the radionuclide and fission product transport are included. A detailed mineralogical characterization of the uranyl mineralogy at the Bangombe uranium deposit in Gabon, Africa was completed to improve

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

NASA Astrophysics Data System (ADS)

Rest, J.; Hofman, G. L.; Kim, Yeon Soo

2009-04-01

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

Anomalies in the Charge Yields of Fission Fragments from the U ( n , f ) 238 Reaction

DOE PAGES

Wilson, J. N.; Lebois, M.; Qi, L.; ...

2017-06-01

Fast-neutron-induced fission of 238U at an energy just above the fission threshold is studied with a novel technique which involves the coupling of a high-efficiency γ-ray spectrometer (MINIBALL) to an inverse-kinematics neutron source (LICORNE) to extract charge yields of fission fragments via γ-γ coincidence spectroscopy. Experimental data and fission models are compared and found to be in reasonable agreement for many nuclei; however, significant discrepancies of up to 600% are observed, particularly for isotopes of Sn and Mo. This indicates that these models significantly overestimate the standard 1 fission mode and suggests that spherical shell effects in the nascent fissionmore » fragments are less important for low-energy fast-neutron-induced fission than for thermal neutron-induced fission. Finally, this has consequences for understanding and modeling the fission process, for experimental nuclear structure studies of the most neutron-rich nuclei, for future energy applications (e.g., Generation IV reactors which use fast-neutron spectra), and for the reactor antineutrino anomaly.« less

Advanced model for the prediction of the neutron-rich fission product yields

NASA Astrophysics Data System (ADS)

Rubchenya, V. A.; Gorelov, D.; Jokinen, A.; Penttilä, H.; Äystö, J.

2013-12-01

The consistent models for the description of the independent fission product formation cross sections in the spontaneous fission and in the neutron and proton induced fission at the energies up to 100 MeV is developed. This model is a combination of new version of the two-component exciton model and a time-dependent statistical model for fusion-fission process with inclusion of dynamical effects for accurate calculations of nucleon composition and excitation energy of the fissioning nucleus at the scission point. For each member of the compound nucleus ensemble at the scission point, the primary fission fragment characteristics: kinetic and excitation energies and their yields are calculated using the scission-point fission model with inclusion of the nuclear shell and pairing effects, and multimodal approach. The charge distribution of the primary fragment isobaric chains was considered as a result of the frozen quantal fluctuations of the isovector nuclear matter density at the scission point with the finite neck radius. Model parameters were obtained from the comparison of the predicted independent product fission yields with the experimental results and with the neutron-rich fission product data measured with a Penning trap at the Accelerator Laboratory of the University of Jyväskylä (JYFLTRAP).

A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

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

Jacques V Hugo; David I Gertman; Jeffrey C Joe

2014-08-01

This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operatingmore » experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.« less

MEANS FOR SHIELDING AND COOLING REACTORS

DOEpatents

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1959-02-10

Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

SHAPED FISSIONABLE METAL BODIES

DOEpatents

Wigner, E.P.; Williamson, R.R.; Young, G.J.

1958-10-14

A technique is presented for grooving the surface of fissionable fuel elements so that expansion can take place without damage to the interior structure of the fuel element. The fissionable body tends to develop internal stressing when it is heated internally by the operation of the nuclear reactor and at the same time is subjected to surface cooling by the circulating coolant. By producing a grooved or waffle-like surface texture, the annular lines of tension stress are disrupted at equally spaced intervals by the grooves, thereby relieving the tension stresses in the outer portions of the body while also facilitating the removal of accumulated heat from the interior portion of the fuel element.

Options For Development of Space Fission Propulsion Systems

NASA Technical Reports Server (NTRS)

Houta, Mike; VanDyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Salvail, Pat; Hrbud, Ivana; Rodgers, Stephen L. (Technical Monitor)

2001-01-01

Fission technology can enable rapid, affordable access to any point in the solar system. Potential fission-based transportation options include high specific power continuous impulse propulsion systems and bimodal nuclear thermal rockets. Despite their tremendous potential for enhancing or enabling deep space and planetary missions, to date space fission system have only been used in Earth orbit. The first step towards utilizing advanced fission propulsion systems is development of a safe, near-term, affordable fission system that can enhance or enable near-term missions of interest. An evolutionary approach for developing space fission propulsion systems is proposed.

NEUTRONIC REACTOR

DOEpatents

Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

1959-10-27

BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

Survey of Materials for Fusion Fission Hybrid Reactors Vol 1 Rev. 0

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

Farmer, Joseph Collin

2007-07-03

Materials for fusion-fission hybrid reactors fall into several broad categories, including fuels, blanket and coolant materials, cladding, structural materials, shielding, and in the specific case of inertial-confinement fusion systems, laser and optical materials. This report surveys materials in all categories of materials except for those required for lasers and optics. Preferred collants include two molten salt mixtures known as FLIBE (Li2BeF4) and FLINABE (LiNaBeF4). In the case of homogenous liquid fuels, UF4 can be dissolved in these molten salt mixtures. The transmutation of lithium in this coolant produces very corrosive hydrofluoric acid species (HF and TF), which can rapidly degrademore » structural materials. Broad ranges of high-melting radiation-tolerant structural material have been proposed for fusion-fission reactor structures. These include a wide variety of steels and refractory alloys. Ferritic steels with oxide-dispersion strengthening and graphite have been given particular attention. Refractory metals are found in Groups IVB and VB of the periodic table, and include Nb, Ta, Cr, Mo, and W, as serve as the basis of refractory alloys. Stable high-melting composites and amorphous metals may also be useful. Since amorphous metals have no lattice structure, neutron bombardment cannot dislodge atoms from lattice sites, and the materials would be immune from this specific mode of degradation. The free energy of formation of fluorides of the alloying elements found in steels and refractory alloys can be used to determine the relative stability of these materials in molten salts. The reduction of lithium transmutation products (H + and T +) drives the electrochemical corrosion process, and liberates aggressive fluoride ions that pair with ions formed from dissolved structural materials. Corrosion can be suppressed through the use of metallic Be and Li, though the molten salt becomes laden with colloidal suspensions of Be and Li corrosion products in

REACTOR FUEL SCAVENGING MEANS

DOEpatents

Coffinberry, A.S.

1962-04-10

A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

Covariance generation and uncertainty propagation for thermal and fast neutron induced fission yields

NASA Astrophysics Data System (ADS)

Terranova, Nicholas; Serot, Olivier; Archier, Pascal; De Saint Jean, Cyrille; Sumini, Marco

2017-09-01

Fission product yields (FY) are fundamental nuclear data for several applications, including decay heat, shielding, dosimetry, burn-up calculations. To be safe and sustainable, modern and future nuclear systems require accurate knowledge on reactor parameters, with reduced margins of uncertainty. Present nuclear data libraries for FY do not provide consistent and complete uncertainty information which are limited, in many cases, to only variances. In the present work we propose a methodology to evaluate covariance matrices for thermal and fast neutron induced fission yields. The semi-empirical models adopted to evaluate the JEFF-3.1.1 FY library have been used in the Generalized Least Square Method available in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation) to generate covariance matrices for several fissioning systems such as the thermal fission of U235, Pu239 and Pu241 and the fast fission of U238, Pu239 and Pu240. The impact of such covariances on nuclear applications has been estimated using deterministic and Monte Carlo uncertainty propagation techniques. We studied the effects on decay heat and reactivity loss uncertainty estimation for simplified test case geometries, such as PWR and SFR pin-cells. The impact on existing nuclear reactors, such as the Jules Horowitz Reactor under construction at CEA-Cadarache, has also been considered.

Analytical measurements of fission products during a severe nuclear accident

NASA Astrophysics Data System (ADS)

Doizi, D.; Reymond la Ruinaz, S.; Haykal, I.; Manceron, L.; Perrin, A.; Boudon, V.; Vander Auwera, J.; tchana, F. Kwabia; Faye, M.

2018-01-01

The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d'Investissement d'Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements) is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium) outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

A summary of the results from the DOE advanced gas reactor (AGR) fuel development and qualification program

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

Petti, David Andrew

2017-04-01

Modular high temperature gas-cooled reactor (HTGR) designs were developed to provide natural safety, which prevents core damage under all licensing basis events. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. The required level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude relative to source terms for other reactor types and allows a graded approach to emergency planning and the potential elimination of the need for evacuation and sheltering beyond a small exclusion area. Achieving this level, however,more » is predicated on exceptionally high coated-particle fuel fabrication quality and excellent performance under normal operation and accident conditions. The design goal of modular HTGRs is to meet the Environmental Protection Agency (EPA) Protective Action Guides (PAGs) for offsite dose at the Exclusion Area Boundary (EAB). To achieve this, the reactor design concepts require a level of fuel integrity that is far better than that achieved for all prior U.S.-manufactured tristructural isotropic (TRISO) coated particle fuel.« less

Cermet-fueled reactors for advanced space applications

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

Cowan, C.L.; Palmer, R.S.; Taylor, I.N.

Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel weremore » carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper.« less

D-He-3 spherical torus fusion reactor system study

NASA Astrophysics Data System (ADS)

Macon, William A., Jr.

1992-04-01

This system study extrapolates present physics knowledge and technology to predict the anticipated characteristics of D-He3 spherical torus fusion reactors and their sensitivity to uncertainties in important parameters. Reference cases for steady-state 1000 MWe reactors operating in H-mode in both the 1st stability regime and the 2nd stability regime were developed and assessed quantitatively. These devices would a very small aspect ratio (A=1,2), a major radius of about 2.0 m, an on-axis magnetic field less than 2 T, a large plasma current (80-120 MA) dominated by the bootstrap effect, and high plasma beta (greater than O.6). The estimated cost of electricity is in the range of 60-90 mills/kW-hr, assuming the use of a direct energy conversion system. The inherent safety and environmental advantages of D-He3 fusion indicate that this reactor concept could be competitive with advanced fission breeder reactors and large-scale solar electric plants by the end of the 21st century if research and development can produce the anticipated physics and technology advances.

Fission control system for nuclear reactor

DOEpatents

Conley, G.H.; Estes, G.P.

Control system for nuclear reactor comprises a first set of reactivity modifying rods fixed in a reactor core with their upper ends stepped in height across the core, and a second set of reactivity modifying rods movable vertically within the reactor core and having their lower ends stepped to correspond with the stepped arrangement of the first set of rods, pairs of the rods of the first and second sets being in coaxial alignment.

Advanced In-Pile Instrumentation for Materials Testing Reactors

NASA Astrophysics Data System (ADS)

Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T. C.; Chase, B. M.; Davis, K. L.; Palmer, A. J.; Schley, R. S.

2014-08-01

The U.S. Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified, and the progress of other development efforts is summarized. As reported in this paper, INL researchers are currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating `advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers, are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors.

A retrospective analysis of funding and focus in US advanced fission innovation

NASA Astrophysics Data System (ADS)

Abdulla, A.; Ford, M. J.; Morgan, M. G.; Victor, D. G.

2017-08-01

Deep decarbonization of the global energy system will require large investments in energy innovation and the deployment of new technologies. While many studies have focused on the expenditure that will be needed, here we focus on how government has spent public sector resources on innovation for a key carbon-free technology: advanced nuclear. We focus on nuclear power because it has been contributing almost 20% of total US electric generation, and because the US program in this area has historically been the world’s leading effort. Using extensive data acquired through the Freedom of Information Act, we reconstruct the budget history of the Department of Energy’s program to develop advanced, non-light water nuclear reactors. Our analysis shows that—despite spending 2 billion since the late 1990s—no advanced design is ready for deployment. Even if the program had been well designed, it still would have been insufficient to demonstrate even one non-light water technology. It has violated much of the wisdom about the effective execution of innovative programs: annual funding varies fourfold, priorities are ephemeral, incumbent technologies and fuels are prized over innovation, and infrastructure spending consumes half the budget. Absent substantial changes, the possibility of US-designed advanced reactors playing a role in decarbonization by mid-century is low.

Advanced Reactor Technologies - Regulatory Technology Development Plan (RTDP)

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

Moe, Wayne L.

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However,more » it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated

Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

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

Moe, Wayne Leland

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However,more » it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated

One-group fission cross sections for plutonium and minor actinides inserted in calculated neutron spectra of fast reactor cooled with lead-208 or lead-bismuth eutectic

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

Khorasanov, G. L.; Blokhin, A. I.

The paper is dedicated to one-group fission cross sections of Pu and MA in LFRs spectra with the aim to increase these values by choosing a coolant which hardens neutron spectra. It is shown that replacement of coolant from Pb-Bi with Pb-208 in the fast reactor RBEC-M, designed in Russia, leads to increasing the core mean neutron energy. As concerns fuel Pu isotopes, their one-group fission cross sections become slightly changed, while more dramatically Am-241 one-group fission cross section is changed. Another situation occurs in the lateral blanket containing small quantities of minor actinides. It is shown that as amore » result of lateral blanket mean neutron energy hardening the one-group fission cross sections of Np-237, Am-241 and Am-243 increases up to 8-11%. This result allows reducing the time of minor actinides burning in FRs. (authors)« less

Advanced propulsion engine assessment based on a cermet reactor

NASA Technical Reports Server (NTRS)

Parsley, Randy C.

1993-01-01

A preferred Pratt & Whitney conceptual Nuclear Thermal Rocket Engine (NTRE) has been designed based on the fundamental NASA priorities of safety, reliability, cost, and performance. The basic philosophy underlying the design of the XNR2000 is the utilization of the most reliable form of ultrahigh temperature nuclear fuel and development of a core configuration which is optimized for uniform power distribution, operational flexibility, power maneuverability, weight, and robustness. The P&W NTRE system employs a fast spectrum, cermet fueled reactor configured in an expander cycle to ensure maximum operational safety. The cermet fuel form provides retention of fuel and fission products as well as high strength. A high level of confidence is provided by benchmark analysis and independent evaluations.

Recent Advances in Power Conversion and Heat Rejection Technology for Fission Surface Power

NASA Technical Reports Server (NTRS)

Mason, Lee

2010-01-01

Under the Exploration Technology Development Program, the National Aeronautics and Space Administration (NASA) and the Department of Energy (DOE) are jointly developing Fission Surface Power (FSP) technology for possible use in human missions to the Moon and Mars. A preliminary reference concept was generated to guide FSP technology development. The concept consists of a liquid-metal-cooled reactor, Stirling power conversion, and water heat rejection, with Brayton power conversion as a backup option. The FSP project has begun risk reduction activities on some key components with the eventual goal of conducting an end-to-end, non-nuclear, integrated system test. Several power conversion and heat rejection hardware prototypes have been built and tested. These include multi-kilowatt Stirling and Brayton power conversion units, titanium-water heat pipes, and composite radiator panels.

Recent MELCOR and VICTORIA Fission Product Research at the NRC

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

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

1999-01-21

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

77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-12

... for Passive Advanced Light Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... Passive Advanced Light Water Reactors.'' The current SRP does not contain guidance on the proposed RTNSS for Passive Advance Light Water Reactors. DATES: Submit comments by November 13, 2012. Comments...

Magnetic Materials Suitable for Fission Power Conversion in Space Missions

NASA Technical Reports Server (NTRS)

Bowman, Cheryl L.

2012-01-01

Terrestrial fission reactors use combinations of shielding and distance to protect power conversion components from elevated temperature and radiation. Space mission systems are necessarily compact and must minimize shielding and distance to enhance system level efficiencies. Technology development efforts to support fission power generation scenarios for future space missions include studying the radiation tolerance of component materials. The fundamental principles of material magnetism are reviewed and used to interpret existing material radiation effects data for expected fission power conversion components for target space missions. Suitable materials for the Fission Power System (FPS) Project are available and guidelines are presented for bounding the elevated temperature/radiation tolerance envelope for candidate magnetic materials.

METHOD OF OPERATING A NEUTRONIC REACTOR

DOEpatents

Turkevich, A.

1963-01-22

This patent relates to one step in a method of operating a neutronic reactor consisting of a slurry of fissionable material in heavy water. Deuterium gas is passed through the slurry to sweep gaseous fission products therefrom and the deuterium is then separated from the gaseous fission products. (AEC)

Plant maintenance and advanced reactors issue, 2008

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

Agnihotri, Newal

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Technologies of national importance, by Tsutomu Ohkubo, Japan Atomic Energy Agency, Japan; Modeling and simulation advances brighten future nuclear power, by Hussein Khalil, Argonne National Laboratory, Energy and desalination projects, by Ratan Kumar Sinha, Bhabha Atomic Research Centre, India; A plant with simplified design, by John Higgins, GE Hitachi Nuclear Energy; A forward thinking design, by Ray Ganthner, AREVA; A passively safe design, by Ed Cummins, Westinghouse Electric Company; A market-ready design, by Ken Petrunik, Atomic Energy of Canada Limited, Canada;more » Generation IV Advanced Nuclear Energy Systems, by Jacques Bouchard, French Commissariat a l'Energie Atomique, France, and Ralph Bennett, Idaho National Laboratory; Innovative reactor designs, a report by IAEA, Vienna, Austria; Guidance for new vendors, by John Nakoski, U.S. Nuclear Regulatory Commission; Road map for future energy, by John Cleveland, International Atomic Energy Agency, Vienna, Austria; and, Vermont's largest source of electricity, by Tyler Lamberts, Entergy Nuclear Operations, Inc. The Industry Innovation article is titled Intelligent monitoring technology, by Chris Demars, Exelon Nuclear.« less

FAST NEUTRON REACTOR

DOEpatents

Soodak, H.; Wigner, E.P.

1961-07-25

A reactor comprising fissionable material in concentration sufficiently high so that the average neutron enengy within the reactor is at least 25,000 ev is described. A natural uranium blanket surrounds the reactor, and a moderating reflector surrounds the blanket. The blanket is thick enough to substantially eliminate flow of neutrons from the reflector.

A NEUTRONIC REACTOR

DOEpatents

Luebke, E.A.; Vandenberg, L.B.

1959-09-01

A nuclear reactor for producing thermoelectric power is described. The reactor core comprises a series of thermoelectric assemblies, each assembly including fissionable fuel as an active element to form a hot junction and a thermocouple. The assemblies are disposed parallel to each other to form spaces and means are included for Introducing an electrically conductive coolant between the assemblies to form cold junctions of the thermocouples. An electromotive force is developed across the entire series of the thermoelectric assemblies due to fission heat generated in the fuel causing a current to flow perpendicular to the flow of coolant and is distributed to a load outside of the reactor by means of bus bars electrically connected to the outermost thermoelectric assembly.

NEUTRONIC REACTOR

DOEpatents

Daniels, F.

1959-10-27

A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

How to Produce a Reactor Neutron Spectrum Using a Proton Accelerator

DOE PAGES

Burns, Kimberly A.; Wootan, David W.; Gates, Robert O.; ...

2015-06-18

A method for reproducing the neutron energy spectrum present in the core of an operating nuclear reactor using an engineered target in an accelerator proton beam is proposed. The protons interact with a target to create neutrons through various (p,n) type reactions. Spectral tailoring of the emitted neutrons can be used to modify the energy of the generated neutron spectrum to represent various reactor spectra. Through the use of moderators and reflectors, the neutron spectrum can be modified to reproduce many different spectra of interest including spectra in small thermal test reactors, large pressurized water reactors, and fast reactors. Themore » particular application of this methodology is the design of an experimental approach for using an accelerator to measure the betas produced during fission to be used to reduce uncertainties in the interpretation of reactor antineutrino measurements. This approach involves using a proton accelerator to produce a neutron field representative of a power reactor, and using this neutron field to irradiate fission foils of the primary isotopes contributing to fission in the reactor, creating unstable, neutron rich fission products that subsequently beta decay and emit electron antineutrinos. A major advantage of an accelerator neutron source over a neutron beam from a thermal reactor is that the fast neutrons can be slowed down or tailored to approximate various power reactor spectra. An accelerator based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra affect parameters such as the resulting fission product beta spectrum.« less

NUCLEAR REACTOR

DOEpatents

Christy, R.F.

1958-07-15

A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

Advanced applications of cosmic-ray muon radiography

NASA Astrophysics Data System (ADS)

Perry, John

The passage of cosmic-ray muons through matter is dominated by the Coulomb interaction with electrons and atomic nuclei. The muon's interaction with electrons leads to continuous energy loss and stopping through the process of ionization. The muon's interaction with nuclei leads to angular diffusion. If a muon stops in matter, other processes unfold, as discussed in more detail below. These interactions provide the basis for advanced applications of cosmic-ray muon radiography discussed here, specifically: 1) imaging a nuclear reactor with near horizontal muons, and 2) identifying materials through the analysis of radiation lengths weighted by density and secondary signals that are induced by cosmic-ray muon trajectories. We have imaged a nuclear reactor, type AGN-201m, at the University of New Mexico, using data measured with a particle tracker built from a set of sealed drift tubes, the Mini Muon Tracker (MMT). Geant4 simulations were compared to the data for verification and validation. In both the data and simulation, we can identify regions of interest in the reactor including the core, moderator, and shield. This study reinforces our claims for using muon tomography to image reactors following an accident. Warhead and special nuclear materials (SNM) imaging is an important thrust for treaty verification and national security purposes. The differentiation of SNM from other materials, such as iron and aluminum, is useful for these applications. Several techniques were developed for material identification using cosmic-ray muons. These techniques include: 1) identifying the radiation length weighted by density of an object and 2) measuring the signals that can indicate the presence of fission and chain reactions. By combining the radiographic images created by tracking muons through a target plane with the additional fission neutron and gamma signature, we are able to locate regions that are fissionable from a single side. The following materials were imaged

Dependence of the prompt fission γ-ray spectrum on the entrance channel of compound nucleus: Spontaneous vs. neutron-induced fission

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

Chyzh, A.; Jaffke, P.; Wu, C. Y.

Prompt γ-ray spectra were measured for the spontaneous fission of 240,242Pu and the neutron-induced fission of 239,241Pu with incident neutron energies ranging from thermal to about 100 keV. Measurements were made using the Detector for Advanced Neutron Capture Experiments (DANCE) array in coincidence with the detection of fission fragments using a parallel-plate avalanche counter. The unfolded prompt fission γ-ray energy spectra can be reproduced reasonably well by Monte Carlo Hauser–Feshbach statistical model for the neutron-induced fission channel but not for the spontaneous fission channel. However, this entrance-channel dependence of the prompt fission γ-ray emission can be described qualitatively by themore » model due to the very different fission-fragment mass distributions and a lower average fragment spin for spontaneous fission. The description of measurements and the discussion of results under the framework of a Monte Carlo Hauser–Feshbach statistical approach are presented.« less

Dependence of the prompt fission γ-ray spectrum on the entrance channel of compound nucleus: Spontaneous vs. neutron-induced fission

DOE PAGES

Chyzh, A.; Jaffke, P.; Wu, C. Y.; ...

2018-06-07

Prompt γ-ray spectra were measured for the spontaneous fission of 240,242Pu and the neutron-induced fission of 239,241Pu with incident neutron energies ranging from thermal to about 100 keV. Measurements were made using the Detector for Advanced Neutron Capture Experiments (DANCE) array in coincidence with the detection of fission fragments using a parallel-plate avalanche counter. The unfolded prompt fission γ-ray energy spectra can be reproduced reasonably well by Monte Carlo Hauser–Feshbach statistical model for the neutron-induced fission channel but not for the spontaneous fission channel. However, this entrance-channel dependence of the prompt fission γ-ray emission can be described qualitatively by themore » model due to the very different fission-fragment mass distributions and a lower average fragment spin for spontaneous fission. The description of measurements and the discussion of results under the framework of a Monte Carlo Hauser–Feshbach statistical approach are presented.« less

Three-dimensional neutronics optimization of helium-cooled blanket for multi-functional experimental fusion-fission hybrid reactor (FDS-MFX)

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

Jiang, J.; Yuan, B.; Jin, M.

2012-07-01

Three-dimensional neutronics optimization calculations were performed to analyse the parameters of Tritium Breeding Ratio (TBR) and maximum average Power Density (PDmax) in a helium-cooled multi-functional experimental fusion-fission hybrid reactor named FDS (Fusion-Driven hybrid System)-MFX (Multi-Functional experimental) blanket. Three-stage tests will be carried out successively, in which the tritium breeding blanket, uranium-fueled blanket and spent-fuel-fueled blanket will be utilized respectively. In this contribution, the most significant and main goal of the FDS-MFX blanket is to achieve the PDmax of about 100 MW/m3 with self-sustaining tritium (TBR {>=} 1.05) based on the second-stage test with uranium-fueled blanket to check and validate themore » demonstrator reactor blanket relevant technologies based on the viable fusion and fission technologies. Four different enriched uranium materials were taken into account to evaluate PDmax in subcritical blanket: (i) natural uranium, (ii) 3.2% enriched uranium, (iii) 19.75% enriched uranium, and (iv) 64.4% enriched uranium carbide. These calculations and analyses were performed using a home-developed code VisualBUS and Hybrid Evaluated Nuclear Data Library (HENDL). The results showed that the performance of the blanket loaded with 64.4% enriched uranium was the most attractive and it could be promising to effectively obtain tritium self-sufficiency (TBR-1.05) and a high maximum average power density ({approx}100 MW/m{sup 3}) when the blanket was loaded with the mass of {sup 235}U about 1 ton. (authors)« less

Fission Product Yields of 233U, 235U, 238U and 239Pu in Fields of Thermal Neutrons, Fission Neutrons and 14.7-MeV Neutrons

NASA Astrophysics Data System (ADS)

Laurec, J.; Adam, A.; de Bruyne, T.; Bauge, E.; Granier, T.; Aupiais, J.; Bersillon, O.; Le Petit, G.; Authier, N.; Casoli, P.

2010-12-01

The yields of more than fifteen fission products have been carefully measured using radiochemical techniques, for 235U(n,f), 239Pu(n,f) in a thermal spectrum, for 233U(n,f), 235U(n,f), and 239Pu(n,f) reactions in a fission neutron spectrum, and for 233U(n,f), 235U(n,f), 238U(n,f), and 239Pu(n,f) for 14.7 MeV monoenergetic neutrons. Irradiations were performed at the EL3 reactor, at the Caliban and Prospero critical assemblies, and at the Lancelot electrostatic accelerator in CEA-Valduc. Fissions were counted in thin deposits using fission ionization chambers. The number of fission products of each species were measured by gamma spectrometry of co-located thick deposits.

Metal fires and their implications for advanced reactors.

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

Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean

This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in thesemore » areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the

Fission Technology for Exploring and Utilizing the Solar System

NASA Technical Reports Server (NTRS)

Houts, Mike; VanDyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Salvail, Pat; Hrbub, Ivana; Schmidt, George R. (Technical Monitor)

2000-01-01

Fission technology can enable rapid, affordable access to any point in the solar system. Potential fission-based transportation options include bimodal nuclear thermal rockets, high specific energy propulsion systems, and pulsed fission propulsion systems. In-space propellant re-supply enhances the effective performance of all systems, but requires significant infrastructure development. Safe, timely, affordable utilization of first-generation space fission propulsion systems will enable the development of more advanced systems. First generation space systems will build on over 45 years of US and international space fission system technology development to minimize cost,

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

NASA Astrophysics Data System (ADS)

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

2005-02-01

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

Enhanced In-Pile Instrumentation at the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Rempe, Joy L.; Knudson, Darrell L.; Daw, Joshua E.; Unruh, Troy; Chase, Benjamin M.; Palmer, Joe; Condie, Keith G.; Davis, Kurt L.

2012-08-01

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

78 FR 20959 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-08

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... Committee on Reactor Safeguards. [FR Doc. 2013-08131 Filed 4-5-13; 8:45 am] BILLING CODE 7590-01-P ...

Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

PubMed

Hill, R N; Nutt, W M; Laidler, J J

2011-01-01

The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. Copyright © 2010 Health Physics Society

Investigation of materials for fusion power reactors

NASA Astrophysics Data System (ADS)

Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

2014-06-01

The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

Nuclear propulsion apparatus with alternate reactor segments

DOEpatents

Szekely, Thomas

1979-04-03

1. Nuclear propulsion apparatus comprising: A. means for compressing incoming air; B. nuclear fission reactor means for heating said air; C. means for expanding a portion of the heated air to drive said compressing means; D. said nuclear fission reactor means being divided into a plurality of radially extending segments; E. means for directing a portion of the compressed air for heating through alternate segments of said reactor means and another portion of the compressed air for heating through the remaining segments of said reactor means; and F. means for further expanding the heated air from said drive means and the remaining heated air from said reactor means through nozzle means to effect reactive thrust on said apparatus.

Gaseous fuel reactors for power systems

NASA Technical Reports Server (NTRS)

Kendall, J. S.; Rodgers, R. J.

1977-01-01

Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

NEUTRONIC REACTOR

DOEpatents

Fraas, A.P.; Mills, C.B.

1961-11-21

A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

FALCON reactor-pumped laser description and program overview

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

None

1989-12-01

The FALCON (Fission Activated Laser CONcept) reactor-pumped laser program at Sandia National Laboratories is examining the feasibility of high-power systems pumped directly by the energy from a nuclear reactor. In this concept we use the highly energetic fission fragments from neutron induced fission to excite a large volume laser medium. This technology has the potential to scale to extremely large optical power outputs in a primarily self-powered device. A laser system of this type could also be relatively compact and capable of long run times without refueling.

A thrust-sheet propulsion concept using fissionable elements

NASA Technical Reports Server (NTRS)

Moeckel, W. E.

1976-01-01

A space propulsion concept is proposed and analyzed which consists of a thin sheet coated on one side with fissionable material, so that nuclear power is converted directly into propulsive power. Thrust is available both from ejected fission fragments and from thermal radiation. Optimum thicknesses are determined for the active and substrate layers. This concept is shown to have potential mission capability (in terms of velocity increments) superior to that of all other advanced propulsion concepts for which performance estimates are available. A suitable spontaneously fissioning material such as Cf254 could provide an extremely high-performance first stage beyond earth orbit. In contrast with some other advanced nuclear propulsion concepts, there is no minimum size below which this concept is infeasible.

A thrust-sheet propulsion concept using fissionable elements

NASA Technical Reports Server (NTRS)

Moeckel, W. E.

1976-01-01

A space propulsion concept is proposed and analyzed which consists of a thin sheet coated on one side with fissionable material, so that nuclear power is converted directly into propulsive power. Thrust is available both from ejected fission fragments and from thermal radiation. Optimum thicknesses are determined for the active and substrate layers. This concept is shown to have potential mission capability (in terms of velocity increments) superior to that of all other advanced propulsion concepts for which performance estimates are available. A suitable spontaneously fissioning material such as Cf-254 could provide an extremely high-performance first stage beyond earth orbit. In contrast with some other advanced nuclear propulsion concepts, there is no minimum size below which this concept is infeasible.

Comparison of actinide production in traveling wave and pressurized water reactors

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

Osborne, A.G.; Smith, T.A.; Deinert, M.R.

The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactormore » cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)« less

78 FR 46621 - Status of the Office of New Reactors' Implementation of Electronic Distribution of Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-01

... of Electronic Distribution of Advanced Reactor Correspondence AGENCY: Nuclear Regulatory Commission. ACTION: Implementation of electronic distribution of advanced reactor correspondence; issuance. SUMMARY... public that, in the future, publicly available correspondence originating from the Division of Advanced...

Strengthening the fission reactor nuclear science and engineering program at UCLA. Final technical report

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

Okrent, D.

1997-06-23

This is the final report on DOE Award No. DE-FG03-92ER75838 A000, a three year matching grant program with Pacific Gas and Electric Company (PG and E) to support strengthening of the fission reactor nuclear science and engineering program at UCLA. The program began on September 30, 1992. The program has enabled UCLA to use its strong existing background to train students in technological problems which simultaneously are of interest to the industry and of specific interest to PG and E. The program included undergraduate scholarships, graduate traineeships and distinguished lecturers. Four topics were selected for research the first year, withmore » the benefit of active collaboration with personnel from PG and E. These topics remained the same during the second year of this program. During the third year, two topics ended with the departure o the students involved (reflux cooling in a PWR during a shutdown and erosion/corrosion of carbon steel piping). Two new topics (long-term risk and fuel relocation within the reactor vessel) were added; hence, the topics during the third year award were the following: reflux condensation and the effect of non-condensable gases; erosion/corrosion of carbon steel piping; use of artificial intelligence in severe accident diagnosis for PWRs (diagnosis of plant status during a PWR station blackout scenario); the influence on risk of organization and management quality; considerations of long term risk from the disposal of hazardous wastes; and a probabilistic treatment of fuel motion and fuel relocation within the reactor vessel during a severe core damage accident.« less

Fission Fragment Mass Distributions and Total Kinetic Energy Release of 235-Uranium and 238-Uranium in Neutron-Induced Fission at Intermediate and Fast Neutron Energies

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

Duke, Dana Lynn

2015-11-12

This Ph.D. dissertation describes a measurement of the change in mass distributions and average total kinetic energy (TKE) release with increasing incident neutron energy for fission of 235U and 238U. Although fission was discovered over seventy-five years ago, open questions remain about the physics of the fission process. The energy of the incident neutron, En, changes the division of energy release in the resulting fission fragments, however, the details of energy partitioning remain ambiguous because the nucleus is a many-body quantum system. Creating a full theoretical model is difficult and experimental data to validate existing models are lacking. Additional fissionmore » measurements will lead to higher-quality models of the fission process, therefore improving applications such as the development of next-generation nuclear reactors and defense. This work also paves the way for precision experiments such as the Time Projection Chamber (TPC) for fission cross section measurements and the Spectrometer for Ion Determination in Fission (SPIDER) for precision mass yields.« less

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

PubMed

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

1994-10-01

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

Recent Advances in Pd-Based Membranes for Membrane Reactors.

PubMed

Arratibel Plazaola, Alba; Pacheco Tanaka, David Alfredo; Van Sint Annaland, Martin; Gallucci, Fausto

2017-01-01

Palladium-based membranes for hydrogen separation have been studied by several research groups during the last 40 years. Much effort has been dedicated to improving the hydrogen flux of these membranes employing different alloys, supports, deposition/production techniques, etc. High flux and cheap membranes, yet stable at different operating conditions are required for their exploitation at industrial scale. The integration of membranes in multifunctional reactors (membrane reactors) poses additional demands on the membranes as interactions at different levels between the catalyst and the membrane surface can occur. Particularly, when employing the membranes in fluidized bed reactors, the selective layer should be resistant to or protected against erosion. In this review we will also describe a novel kind of membranes, the pore-filled type membranes prepared by Pacheco Tanaka and coworkers that represent a possible solution to integrate thin selective membranes into membrane reactors while protecting the selective layer. This work is focused on recent advances on metallic supports, materials used as an intermetallic diffusion layer when metallic supports are used and the most recent advances on Pd-based composite membranes. Particular attention is paid to improvements on sulfur resistance of Pd based membranes, resistance to hydrogen embrittlement and stability at high temperature.

REACTOR

DOEpatents

Szilard, L.

1963-09-10

A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

Health physics aspects of advanced reactor licensing reviews

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

Hinson, C.S.

1995-03-01

The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on {open_quotes}next-generation{close_quotes} reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovativemore » design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs currently being reviewed by the NRC.« less

76 FR 5218 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-28

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on February 8, 2011, 11545 Rockville Pike, Rockville, MD...

The U.S. Geological Survey's TRIGA® reactor

USGS Publications Warehouse

DeBey, Timothy M.; Roy, Brycen R.; Brady, Sally R.

2012-01-01

The U.S. Geological Survey (USGS) operates a low-enriched uranium-fueled, pool-type reactor located at the Federal Center in Denver, Colorado. The mission of the Geological Survey TRIGA® Reactor (GSTR) is to support USGS science by providing information on geologic, plant, and animal specimens to advance methods and techniques unique to nuclear reactors. The reactor facility is supported by programs across the USGS and is organizationally under the Associate Director for Energy and Minerals, and Environmental Health. The GSTR is the only facility in the United States capable of performing automated delayed neutron analyses for detecting fissile and fissionable isotopes. Samples from around the world are submitted to the USGS for analysis using the reactor facility. Qualitative and quantitative elemental analyses, spatial elemental analyses, and geochronology are performed. Few research reactor facilities in the United States are equipped to handle the large number of samples processed at the GSTR. Historically, more than 450,000 sample irradiations have been performed at the USGS facility. Providing impartial scientific information to resource managers, planners, and other interested parties throughout the world is an integral part of the research effort of the USGS.

Solar vs. Fission Surface Power for Mars

NASA Technical Reports Server (NTRS)

Rucker, Michelle A.; Oleson, Steve; George, Pat; Landis, Geoffrey A.; Fincannon, James; Bogner, Amee; Jones, Robert E.; Turnbull, Elizabeth; Martini, Michael C.; Gyekenyesi, John Z.;

Prompt fission gamma-ray studies at DANCE

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

Jandel, M.; Rusev, G.; Bond, E. M.

2014-11-26

Measurements of correlated data on prompt-fission γ-rays (PFG) have been carried out for various actinide isotopes in recent years using the Detector for Advanced Neutron Capture Experiments (DANCE) at Los Alamos National Laboratory (LANL). We have developed a model that conveniently parametrizes the correlated data of γ-ray multiplicity and energy. New results on two- dimensional prompt-fission γ-ray multiplicity versus energy distributions from spontaneous fission on ²⁵²Cf and neutron-induced fission on 242mAm are presented together with previously obtained results on 233,235U and ²³⁹Pu. Correlated PFG data from ²⁵²Cf are also compared to results of the detailed theoretical model developed at LANL,more » for different thresholds of PFG energies. Future plans to measure correlated data on fission fragments, prompt fission neutrons and γ-rays at DANCE are presented.« less

Measurement of the prompt fissionγ-ray spectrum of 242Pu

NASA Astrophysics Data System (ADS)

Urlass, Sebastian; Beyer, Roland; Junghans, Arnd Rudolf; Kögler, Toni; Schwengner, Ronald; Wagner, Andreas

2018-03-01

The prompt γ-ray spectrum of fission fragments is important in understanding the dynamics of the fission process, as well as for nuclear engineering in terms of predicting the γ-ray heating in nuclear reactors. The γ-ray spectrum measured from the fission fragments of the spontaneous fission of 242Pu will be presented here. A fission chamber containing in total 37mg of 242Pu was used as active sample. The γ-quanta were detected with high time- and energy-resolution using LaBr3 and HPGe detectors, respectively, in coincidence with spontaneous fission events detected by the fission chamber. The acquired γ-ray spectra were corrected for the detector response using the spectrum stripping method. About 70 million fission events were detected which results in a very low statistical uncertainty and a wider energy range covered compared to previous measurements. The prompt fission γ-ray spectrum measured with the HPGe detectors shows structures that allow conclusions about the nature of γ-ray transitions in the fission fragments. The average photon multiplicity of 8.2 and the average total energy release by prompt photons per fission event of about 6.8 MeV were determined for both detector types.

77 FR 76089 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-26

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on January 16, 2013, Room T-2B3, 11545 Rockville Pike...

NEUTRONIC REACTOR POWER PLANT

DOEpatents

Metcalf, H.E.

1962-12-25

This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

77 FR 59678 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-28

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on October 2, 2012, Room T-2B1, 11545 Rockville Pike, Rockville...

76 FR 34276 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-13

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on June 21, 2011, Room T-2B1, 11545 Rockville Pike, Rockville...

Thorium fueled reactor

NASA Astrophysics Data System (ADS)

Sipaun, S.

2017-01-01

Current development in thorium fueled reactors shows that they can be designed to operate in the fast or thermal spectrum. The thorium/uranium fuel cycle converts fertile thorium-232 into fissile uranium-233, which fissions and releases energy. This paper analyses the characteristics of thorium fueled reactors and discusses the thermal reactor option. It is found that thorium fuel can be utilized in molten salt reactors through many configurations and designs. A balanced assessment on the feasibility of adopting one reactor technology versus another could lead to optimized benefits of having thorium resource.

A Blueprint for GNEP Advanced Burner Reactor Startup Fuel Fabrication Facility

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

S. Khericha

2010-12-01

The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn thesemore » actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the startup fuel for 10-15 years and would take 3 to 5 years to construct.« less

ANALYTICAL CHEMISTRY IN NUCLEAR REACTOR TECHNOLOGY. Analysis of Reactor Fuels, Fission-Product Mixtures and Related Materials: Analytical Chemistry of Plutonium and the Transplutonic Elements. Third Conference, Gatlinburg, Tennessee, October 26-29, 1959

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

None

1960-01-01

Thirty-one papers and 10 summaries of papers presented at the Third Conference on Analytical Chemistry in Nuclear Reactor Technology held at Gatlinburg, Tennessee, October 26 to 29, 1959, are given. The papers are grouped into four sections: general, analytical chemistry of fuels, analytical chemistry of plutonium and the transplutonic elements, and the analysis of fission-product mixtures. Twenty-seven of the papers are covered by separate abstracts. Four were previously abstracted for NSA. (M.C.G.)

COMPOSITE NEUTRONIC REACTOR

DOEpatents

Menke, J.R.

1963-06-11

This patent relates to a reactor having a core which comprises an inner active region and an outer active region, each region separately having a k effective less than one and a k infinity greater than one. The inner and outer regions in combination have a k effective at least equal to one and each region contributes substantially to the k effective of the reactor core. The inner region has a low moderator to fuel ratio such that the majority of fissions occurring therein are induced by neutrons having energies greater than thermal. The outer region has a high moderator to fuel ratio such that the majority of fissions occurring therein are induced by thermal neutrons. (AEC)

NEUTRONIC REACTORS

DOEpatents

Wigner, E.P.

1960-11-22

A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

Breeder Reactors, Understanding the Atom Series.

ERIC Educational Resources Information Center

Mitchell, Walter, III; Turner, Stanley E.

The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

Parameter study of dual-mode space nuclear fission solid core power and propulsion systems, NUROC3A. AMS report No. 1239c

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

Smith, W.W.; Layton, J.P.

1976-09-13

The three-volume report describes a dual-mode nuclear space power and propulsion system concept that employs an advanced solid-core nuclear fission reactor coupled via heat pipes to one of several electric power conversion systems. The NUROC3A systems analysis code was designed to provide the user with performance characteristics of the dual-mode system. Volume 3 describes utilization of the NUROC3A code to produce a detailed parameter study of the system.

Nuclear reactor fuel element

DOEpatents

Johnson, Carl E.; Crouthamel, Carl E.

1980-01-01

A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

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

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

McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel

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 suchmore » 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.« less

78 FR 66968 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-07

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on November 22, 2013, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. The meeting will be...

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

NASA Astrophysics Data System (ADS)

Bundgaard, Jeremy; Niffte Collaboration

2015-10-01

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

78 FR 37595 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-21

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on July 9, 2013, Room T-2B3, 11545 Rockville Pike, Rockville, Maryland. The meeting will be open to...

An advanced carbon reactor subsystem for carbon dioxide reduction

NASA Technical Reports Server (NTRS)

Noyes, Gary P.; Cusick, Robert J.

1986-01-01

An evaluation is presented of the development status of an advanced carbon-reactor subsystem (ACRS) for the production of water and dense, solid carbon from CO2 and hydrogen, as required in physiochemical air revitalization systems for long-duration manned space missions. The ACRS consists of a Sabatier Methanation Reactor (SMR) that reduces CO2 with hydrogen to form methane and water, a gas-liquid separator to remove product water from the methane, and a Carbon Formation Reactor (CFR) to pyrolize methane to carbon and hydrogen; the carbon is recycled to the SMR, while the produce carbon is periodically removed from the CFR. A preprototype ACRS under development for the NASA Space Station is described.

EXPERIMENTAL LIQUID METAL FUEL REACTOR

DOEpatents

Happell, J.J.; Thomas, G.R.; Denise, R.P.; Bunts, J.L. Jr.

1962-01-23

A liquid metal fuel nuclear fission reactor is designed in which the fissionable material is dissolved or suspended in a liquid metal moderator and coolant. The liquid suspension flows into a chamber in which a critical amount of fissionable material is obtained. The fluid leaves the chamber and the heat of fission is extracted for power or other utilization. The improvement is in the support arrangement for a segrnented graphite core to permit dif ferential thermal expansion, effective sealing between main and blanket liquid metal flows, and avoidance of excessive stress development in the graphite segments. (AEC)

Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

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

Cormon, S.; Fallot, M., E-mail: fallot@subatech.in2p3.fr; Bui, V.-M.

This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (ν{sup ¯}{sub e}) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of {sup 235}U, {supmore » 239}Pu and {sup 241}Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.« less

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

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

Abrecht, David G.; Schwantes, Jon M.

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 (ΔG rxn°(T C))/(RT C)+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(T C). These models allowedmore » an estimate of the upper bound for the reactor temperatures of T C 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.« less

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

PubMed

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 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.

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

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

Not Available

1992-04-01

This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets.

PubMed

Abedi, Ebrahim; Ebrahimkhani, Marzieh; Davari, Amin; Mirvakili, Seyed Mohammad; Tabasi, Mohsen; Maragheh, Mohammad Ghannadi

2016-12-01

Efficient and safe production of molybdenum-99 ( 99 Mo) radiopharmaceutical at Tehran Research Reactor (TRR) via fission of LEU targets is studied. Neutronic calculations are performed to evaluate produced 99 Mo activity, core neutronic safety parameters and also the power deposition values in target plates during a 7 days irradiation interval. Thermal-hydraulic analysis has been also carried out to obtain thermal behavior of these plates. Using Thermal-hydraulic analysis, it can be concluded that the safety parameters are satisfied in the current study. Consequently, the present neutronic and thermal-hydraulic calculations show efficient 99 Mo production is accessible at significant activity values in TRR current core configuration. Copyright © 2016 Elsevier Ltd. All rights reserved.

VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

DOEpatents

Furgerson, W.T.

1963-12-17

A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

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

NASA Astrophysics Data System (ADS)

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

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

NEUTRONIC REACTOR DESIGN TO REDUCE NEUTRON LOSS

DOEpatents

Mills, F.T.

1961-05-01

A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall which is surrounded by successive layers of pure fertile material and fertile material having moderator. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. As the steel has a smaller capture cross-section for the fast neutrons, then greater numbers of the neutrons will pass into the blanket thereby increasing the over-all efficiency of the reactor.

Neutronic Reactor Design to Reduce Neutron Loss

DOEpatents

Miles, F. T.

1961-05-01

A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall. The wall is surrounded by successive layers of pure fertile material and moderator containing fertile material. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. Since the steel has a smaller capture cross section for the fast neutrons, greater nunnbers of neutrons will pass into the blanket, thereby increasing the over-all efficiency of the reactor. (AEC)

Recent Results from Lohengrin on Fission Yields and Related Decay Properties

NASA Astrophysics Data System (ADS)

Serot, O.; Amouroux, C.; Bidaud, A.; Capellan, N.; Chabod, S.; Ebran, A.; Faust, H.; Kessedjian, G.; Köester, U.; Letourneau, A.; Litaize, O.; Martin, F.; Materna, T.; Mathieu, L.; Panebianco, S.; Regis, J.-M.; Rudigier, M.; Sage, C.; Urban, W.

2014-05-01

The Lohengrin mass spectrometer is one of the 40 instruments built around the reactor of the Institute Laue-Langevin (France) which delivers a very intense thermal neutron flux. Usually, Lohengrin was combined with a high-resolution ionization chamber in order to obtain good nuclear charge discrimination within a mass line, yielding an accurate isotopic yield determination. Unfortunately, this experimental procedure can only be applied for fission products with a nuclear charge less than about 42, i.e. in the light fission fragment region. Since 2008, a large collaboration has started with the aim of studying various fission aspects, mainly in the heavy fragment region. For that, a new experimental setup which allows isotopic identification by γ-ray spectrometry has been developed and validated. This technique was applied on the 239Pu(nth,f) reaction where about 65 fission product yields were measured with an uncertainty that has been reduced on average by a factor of 2 compared with what was that previously available in nuclear data libraries. The same γ-ray spectrometric technique is currently being applied to the study of the 233U(nth,f) reaction. Our aim is to deduce charge and mass distributions of the fission products and to complete the experimental data that exist mainly for light fission fragments. The measurement of 41 mass yields from the 241Am(2nth,f) reaction has been also performed. In addition to these activities on fission yield measurements, various new nanosecond isomers were discovered. Their presence can be revealed from a strong deformed ionic charge distribution compared to a 'normal' Gaussian shape. Finally, a new neutron long-counter detector designed to have a detection efficiency independent of the detected neutron energy has been built. Combining this neutron device with a Germanium detector and a beta-ray detector array allowed us to measure the beta-delayed neutron emission probability Pn of some important fission products for reactor

a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

2009-08-01

This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

NEUTRONIC REACTOR COUNTER METHOD AND SYSTEM

DOEpatents

Graham, C.B.; Spiewak, I.

1960-05-31

An improved method is given for controlling the rate of fission in circulating-fuel neutronic reactors in which the fuel is a homogeneous liquid containing fissionable material and a neutron moderator. A change in the rate of flssion is effected by preferentially retaining apart from the circulating fuel a variable amount of either fissionable material or moderator, thereby varying the concentration of fissionable material in the fuel. In the case of an aqueous fuel solution a portion of the water may be continuously vaporized from the circulating solution and the amount of condensate, or condensate plus make-up water, returned to the solution is varied to control the fission rate.

NEUTRON REACTOR HAVING A Xe$sup 135$ SHIELD

DOEpatents

Stanton, H.E.

1957-10-29

Shielding for reactors of the type in which the fuel is a chain reacting liquid composition comprised essentially of a slurry of fissionable and fertile material suspended in a liquid moderator is discussed. The neutron reflector comprises a tank containing heavy water surrounding the reactor, a shield tank surrounding the reflector, a gamma ray shield surrounding said shield tank, and a means for conveying gaseous fission products, particularly Xe/sup 135/, from the reactor chamber to the shield tank, thereby serving as a neutron shield by capturing the thermalized neutrons that leak outwardly from the shield tank.

Fission Surface Power for the Exploration and Colonization of Mars

NASA Technical Reports Server (NTRS)

Houts, Mike; Porter, Ron; Gaddis, Steve; Van Dyke, Melissa; Martin, Jim; Godfroy, Tom; Bragg-Sitton, Shannon; Garber, Anne; Pearson, Boise

2006-01-01

The colonization of Mars will require abundant energy. One potential energy source is nuclear fission. Terrestrial fission systems are highly developed and have the demonstrated ability to safely produce tremendous amounts of energy. In space, fission systems not only have the potential to safely generate tremendous amounts of energy, but could also potentially be used on missions where alternatives are not practical. Programmatic risks such as cost and schedule are potential concerns with fission surface power (FSP) systems. To be mission enabling, FSP systems must be affordable and programmatic risk must be kept acceptably low to avoid jeopardizing exploration efforts that may rely on FSP. Initial FSP systems on Mars could be "workhorse" units sized to enable the establishment of a Mars base and the early growth of a colony. These systems could be nearly identical to FSP systems used on the moon. The systems could be designed to be safe, reliable, and have low development and recurring costs. Systems could also be designed to fit on relatively small landers. One potential option for an early Mars FSP system would be a 100 kWt class, NaK cooled system analogous to space reactors developed and flown under the U.S. "SNAP" program or those developed and flown by the former Soviet Union ("BUK" reactor). The systems could use highly developed fuel and materials. Water and Martian soil could be used to provide shielding. A modern, high-efficiency power conversion subsystem could be used to reduce required reactor thermal power. This, in turn, would reduce fuel burnup and radiation damage .effects by reducing "nuclear" fuels and materials development costs. A realistic, non-nuclear heated and fully integrated technology demonstration unit (TDU) could be used to reduce cost and programmatic uncertainties prior to initiating a flight program.

Integrated intelligent systems in advanced reactor control rooms

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

Beckmeyer, R.R.

1989-01-01

An intelligent, reactor control room, information system is designed to be an integral part of an advanced control room and will assist the reactor operator's decision making process by continuously monitoring the current plant state and providing recommended operator actions to improve that state. This intelligent system is an integral part of, as well as an extension to, the plant protection and control systems. This paper describes the interaction of several functional components (intelligent information data display, technical specifications monitoring, and dynamic procedures) of the overall system and the artificial intelligence laboratory environment assembled for testing the prototype. 10 refs.,more » 5 figs.« less

Analysis of C/E results of fission rate ratio measurements in several fast lead VENUS-F cores

NASA Astrophysics Data System (ADS)

Kochetkov, Anatoly; Krása, Antonín; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente; Bianchini, Giancarlo; Fabrizio, Valentina; Carta, Mario; Firpo, Gabriele; Fridman, Emil; Sarotto, Massimo

2017-09-01

During the GUINEVERE FP6 European project (2006-2011), the zero-power VENUS water-moderated reactor was modified into VENUS-F, a mock-up of a lead cooled fast spectrum system with solid components that can be operated in both critical and subcritical mode. The Fast Reactor Experiments for hybrid Applications (FREYA) FP7 project was launched in 2011 to support the designs of the MYRRHA Accelerator Driven System (ADS) and the ALFRED Lead Fast Reactor (LFR). Three VENUS-F critical core configurations, simulating the complex MYRRHA core design and one configuration devoted to the LFR ALFRED core conditions were investigated in 2015. The MYRRHA related cores simulated step by step design peculiarities like the BeO reflector and in pile sections. For all of these cores the fuel assemblies were of a simple design consisting of 30% enriched metallic uranium, lead rodlets to simulate the coolant and Al2O3 rodlets to simulate the oxide fuel. Fission rate ratios of minor actinides such as Np-237, Am-241 as well as Pu-239, Pu-240, Pu-242 and U-238 to U-235 were measured in these VENUS-F critical assemblies with small fission chambers in specially designed locations, to determine the spectral indices in the different neutron spectrum conditions. The measurements have been analyzed using advanced computational tools including deterministic and stochastic codes and different nuclear data sets like JEFF-3.1, JEFF-3.2, ENDF/B7.1 and JENDL-4.0. The analysis of the C/E discrepancies will help to improve the nuclear data in the specific energy region of fast neutron reactor spectra.

Correlated Production and Analog Transport of Fission Neutrons and Photons using Fission Models FREYA, FIFRELIN and the Monte Carlo Code TRIPOLI-4® .

NASA Astrophysics Data System (ADS)

Verbeke, Jérôme M.; Petit, Odile; Chebboubi, Abdelhazize; Litaize, Olivier

2018-01-01

Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: FREYA and FIFRELIN. TRIPOLI-4 simulations can now be performed, either by connecting via an API to the LLNL fission library including FREYA, or by reading external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, broad underlying principles of the two fission models are recalled. We then present experimental measurements of neutron angular correlations for 252Cf(sf) and 240Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of the paper, simulation results are compared to experimental data. Spontaneous fissions in 252Cf and 240Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using

Research on fission fragment excitation of gases and nuclear pumping of lasers

NASA Technical Reports Server (NTRS)

Schneider, R. T.; Davie, R. N.; Davis, J. F.; Fuller, J. L.; Paternoster, R. R.; Shipman, G. R.; Sterritt, D. E.; Helmick, H. H.

1974-01-01

Experimental investigations of fission fragment excited gases are reported along with a theoretical analysis of population inversions in fission fragment excited helium. Other studies reported include: nuclear augmentation of gas lasers, direct nuclear pumping of a helium-xenon laser, measurements of a repetitively pulsed high-power CO2 laser, thermodynamic properties of UF6 and UF6/He mixtures, and nuclear waste disposal utilizing a gaseous core reactor.

From pure fusion to fusion-fission Demo tokamaks

NASA Astrophysics Data System (ADS)

Mirnov, S. V.

2013-04-01

The major requirements for pure fusion tokamak reactors and tokamak-based fusion neutron sources (FNS) are analyzed together with possible paths from the present-day tokamak towards the FNS tokamak. The FNS are of interest for traditional fission reactors as a method of waste management by burning of long-lived transuranic radionuclides (minorities) and fission fuel breeding. The Russian fission community places several hard requirements on the quality of FNS suitable for the first step of the investigation program of minority burning and breeding. They are (a) a steady-state regime of neutron production (more than 80% of the operational time), (b) a neutron power flux density greater than >0.2 MW m-2, (c) a total surface integrated neutron power >10 MW. Among the different FNS projects, based on magnetically confined plasmas, only ‘classical tokamak’ is most likely to fulfill these requirements in the nearest future. Some of the most important improvements of the ‘classical tokamak’ needed for successful realization of the FNS are (1) decrease in Zeff (probably, by making use of lithium as a part of plasma-facing components), (2) He removal and closed loop DT fuel circulation, (3) increase in the energy of stationary injected neutral tritium beams up to 150-170 keV and (4) control of impurity contamination at the plasma center (probably, by local RF heating). These key issues are discussed.

Thermos reactors

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

Labrousse, M.; Lerouge, B.; Dupuy, G.

1978-04-01

THERMOS is a water reactor designed to provide hot water up to 120/sup 0/C for district heating or for desalination applications. It is a 100-MW reactor based on proven technology: oxide fuel plate elements, integrated primary circuit, and reactor vessel located in the bottom of a pool. As in swimming pool reactors, the pool is used for biological shielding, emergency core cooling, and fission product filtering (in case of an accident). Before economics, safety is the main characteristic of the concept: no fuel failure admitted, core under water in any accidental configuration, inspection of every ''nuclear'' component, and double-wall containment.

ASME Material Challenges for Advanced Reactor Concepts

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

Piyush Sabharwall; Ali Siahpush

2013-07-01

This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at highermore » temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.« less

Beta decay heat following U-235, U-238 and Pu-239 neutron fission

NASA Astrophysics Data System (ADS)

Li, Shengjie

1997-09-01

This is an experimental study of beta-particle decay heat from 235U, 239Pu and 238U aggregate fission products over delay times 0.4-40,000 seconds. The experimental results below 2s for 235U and 239Pu, and below 20s for 238U, are the first such results reported. The experiments were conducted at the UMASS Lowell 5.5-MV Van de Graaff accelerator and 1-MW swimming-pool research reactor. Thermalized neutrons from the 7Li(p,n)7Be reaction induced fission in 238U and 239Pu, and fast neutrons produced in the reactor initiated fission in 238U. A helium-jet/tape-transport system rapidly transferred fission fragments from a fission chamber to a low background counting area. Delay times after fission were selected by varying the tape speed or the position of the spray point relative to the beta spectrometer that employed a thin-scintillator-disk gating technique to separate beta-particles from accompanying gamma-rays. Beta and gamma sources were both used in energy calibration. Based on low-energy(<1 MeV) internal-conversion electron studies, a set of trial responses for the spectrometer was established and spanned electron energies 0-10 MeV. Measured beta spectra were unfolded for their energy distributions by the program FERD, and then compared to other measurements and summation calculations based on ENDF/B-VI fission-product data performed on the LANL Cray computer. Measurements of the beta activity as a function of decay time furnished a relative normalization. Results for the beta decay heat are presented and compared with other experimental data and the summation calculations.

The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

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

T. R. Allen; J. B. Benson; J. A. Foster

2009-05-01

To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities ismore » granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty

Testing in Support of Fission Surface Power System Qualification

NASA Technical Reports Server (NTRS)

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

2007-01-01

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

Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE-100 Core

NASA Technical Reports Server (NTRS)

Martin, James J.; Reid, Robert S.

2004-01-01

A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat pipe reactor option. A hardware-based program was initiated to establish the infrastructure necessary to build heat pipe modules. This effort, initiated by Los Alamos National Laboratory and referred to as the Safe Affordable Fission Engine (SAFE) project, set out to fabricate and perform non-nuclear testing on a modular heat pipe reactor prototype that can provide 100 kilowatt from the core to an energy conversion system at 700 C. Prototypic heat pipe hardware was designed, fabricated, filled, closed-out and acceptance tested.

Thermodynamics of soluble fission products cesium and iodine in the Molten Salt Reactor

NASA Astrophysics Data System (ADS)

Capelli, E.; Beneš, O.; Konings, R. J. M.

2018-04-01

The present study describes the full thermodynamic assessment of the Li,Cs,Th//F,I system. The existing database for the relevant fluoride salts considered as fuel for the Molten Salt Reactor (MSR) has been extended with two key fission products, cesium and iodine. A complete evaluation of all the common-ion binary and ternary sub-systems of the LiF-ThF4-CsF-LiI-ThI4-CsI system has been performed and the optimized parameters are presented in this work. New equilibrium data have been measured using Differential Scanning Calorimetry and were used to assess the reciprocal ternary systems and confirm the extrapolated phase diagrams. The developed database significantly contributes to the understanding of the behaviour of cesium and iodine in the MSR, which strongly depends on their concentration and chemical form. Cesium bonded with fluorine is well retained in the fuel mixture while in the form of CsI the solubility of these elements is very limited. Finally, the influence of CsI and CsF on the physico-chemical properties of the fuel mixture was calculated as function of composition.

10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

Code of Federal Regulations, 2010 CFR

2010-01-01

... Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES... Rule for the U.S. Advanced Boiling Water Reactor I. Introduction Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR...

10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

Code of Federal Regulations, 2011 CFR

2011-01-01

... Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES... Rule for the U.S. Advanced Boiling Water Reactor I. Introduction Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR...

JACKETED FUEL ELEMENTS FOR GRAPHITE MODERATED REACTORS

DOEpatents

Szilard, L.; Wigner, E.P.; Creutz, E.C.

1959-05-12

Fuel elements for a heterogeneous, fluid cooled, graphite moderated reactor are described. The fuel elements are comprised of a body of natural uranium hermetically sealed in a jacket of corrosion resistant material. The jacket, which may be aluminum or some other material which is non-fissionable and of a type having a low neutron capture cross-section, acts as a barrier between the fissioning isotope and the coolant or moderator or both. The jacket minimizes the tendency of the moderator and coolant to become radioactive and/or contaminated by fission fragments from the fissioning isotope.

Spectral structure of electron antineutrinos from nuclear reactors.

PubMed

Dwyer, D A; Langford, T J

2015-01-09

Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principles calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructures in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of these substructures can elucidate the nuclear processes occurring within reactors. These substructures can be a systematic issue for measurements utilizing the detailed spectral shape.

Structural materials challenges for advanced reactor systems

NASA Astrophysics Data System (ADS)

Yvon, P.; Carré, F.

2009-03-01

Key technologies for advanced nuclear systems encompass high temperature structural materials, fast neutron resistant core materials, and specific reactor and power conversion technologies (intermediate heat exchanger, turbo-machinery, high temperature electrolytic or thermo-chemical water splitting processes, etc.). The main requirements for the materials to be used in these reactor systems are dimensional stability under irradiation, whether under stress (irradiation creep or relaxation) or without stress (swelling, growth), an acceptable evolution under ageing of the mechanical properties (tensile strength, ductility, creep resistance, fracture toughness, resilience) and a good behavior in corrosive environments (reactor coolant or process fluid). Other criteria for the materials are their cost to fabricate and to assemble, and their composition could be optimized in order for instance to present low-activation (or rapid desactivation) features which facilitate maintenance and disposal. These requirements have to be met under normal operating conditions, as well as in incidental and accidental conditions. These challenging requirements imply that in most cases, the use of conventional nuclear materials is excluded, even after optimization and a new range of materials has to be developed and qualified for nuclear use. This paper gives a brief overview of various materials that are essential to establish advanced systems feasibility and performance for in pile and out of pile applications, such as ferritic/martensitic steels (9-12% Cr), nickel based alloys (Haynes 230, Inconel 617, etc.), oxide dispersion strengthened ferritic/martensitic steels, and ceramics (SiC, TiC, etc.). This article gives also an insight into the various natures of R&D needed on advanced materials, including fundamental research to investigate basic physical and chemical phenomena occurring in normal and accidental operating conditions, lab-scale tests to characterize candidate materials

Radiogenic lead as coolant, reflector and moderator in advanced fast reactors

NASA Astrophysics Data System (ADS)

Kulikov, E. G.

2017-01-01

Main purpose of the study is assessing reasonability for recovery, production and application of radiogenic lead as a coolant, neutron moderator and neutron reflector in advanced fast reactors. When performing the study, thermal, physical and neutron-physical properties of natural and radiogenic lead were analyzed. The following results were obtained: 1. Radiogenic lead with high content of isotope 208Pb can be extracted from thorium or mixed thorium-uranium ores because 208Pb is a final product of 232Th natural decay chain. 2. The use of radiogenic lead with high 208Pb content in advanced fast reactors and accelerator-driven systems (ADS) makes it possible to improve significantly their neutron-physical and thermal-hydraulic parameters. 3. The use of radiogenic lead with high 208Pb content in advanced fast reactors as a coolant opens the possibilities for more intense fuel breeding and for application of well-known oxide fuel instead of the promising but not tested enough nitride fuel under the same safety parameters. 4. The use of radiogenic lead with high 208Pb content in ADS as a coolant can upgrade substantially the level of neutron flux in the ADS blanket, which enables effective transmutation of radioactive wastes with low cross-sections of radiative neutron capture.

NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR

DOEpatents

Rasor, N.S.; Hirsch, R.L.

1963-12-01

The patent relates to the direct conversion of fission heat to electricity by use of thermionic plasma diodes having fissionable material cathodes, said diodes arranged to form a critical mass in a nuclear reactor. The patent describes a fuel element comprising a plurality of diodes each having a fissionable material cathode, an anode around said cathode, and an ionizable gas therebetween. Provision is made for flowing the gas and current serially through the diodes. (AEC)

Development of advanced strain diagnostic techniques for reactor environments.

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

Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.

2013-02-01

The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding.more » During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.« less

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

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

Pierson, Bruce D.; Finn, Erin C.; Friese, Judah I.

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 datamore » collected from each detector. Ultimately, these three portable detectors were able to identify a large range of fission products showing potential for material discrimination.« less

Radio-toxicity of spent fuel of the advanced heavy water reactor.

PubMed

Anand, S; Singh, K D S; Sharma, V K

2010-01-01

The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR.

Propellant actuated nuclear reactor steam depressurization valve

DOEpatents

Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

1992-01-01

A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

NEUTRONIC REACTORS

DOEpatents

Vernon, H.C.

1959-01-13

A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

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

NONE

1993-09-15

This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

Legal and Regulatroy Obstacles to Nuclear Fission Technology in Space

NASA Astrophysics Data System (ADS)

Force, Melissa K.

2013-09-01

In forecasting the prospective use of small nuclear reactors for spacecraft and space-based power stations, the U.S. Air Force describes space as "the ultimate high ground," providing access to every part of the globe. But is it? A report titled "Energy Horizons: United States Air Force Energy Science &Technology Vision 2011-2026," focuses on core Air Force missions in space energy generation, operations and propulsion and recognizes that investments into small modular nuclear fission reactors can be leveraged for space-based systems. However, the report mentions, as an aside, that "potential catastrophic outcomes" are an element to be weighed and provides no insight into the monumental political and legal will required to overcome the mere stigma of nuclear energy, even when referring only to the most benign nuclear power generation systems - RTGs. On the heels of that report, a joint Department of Energy and NASA team published positive results from the demonstration of a uranium- powered fission reactor. The experiment was perhaps most notable for exemplifying just how effective the powerful anti-nuclear lobby has been in the United States: It was the first such demonstration of its kind in nearly fifty years. Space visionaries must anticipate a difficult war, consisting of multiple battles that must be waged in order to obtain a license to fly any but the feeblest of nuclear power sources in space. This paper aims to guide the reader through the obstacles to be overcome before nuclear fission technology can be put to use in space.

Molten salt considerations for accelerator-driven subcritical fission to close the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Sooby, Elizabeth; Adams, Marvin; Baty, Austin; Gerity, James; McIntyre, Peter; Melconian, Karie; Phongikaroon, Supathorn; Pogue, Nathaniel; Sattarov, Akhdiyor; Simpson, Michael; Tripathy, Prabhat; Tsevkov, Pavel

2013-04-01

The host salt selection, molecular modeling, physical chemistry, and processing chemistry are presented here for an accelerator-driven subcritical fission in a molten salt core (ADSMS). The core is fueled solely with the transuranics (TRU) and long-lived fission products (LFP) from used nuclear fuel. The neutronics and salt composition are optimized to destroy the transuranics by fission and the long-lived fission products by transmutation. The cores are driven by proton beams from a strong-focusing cyclotron stack. One such ADSMS system can destroy the transuranics in the used nuclear fuel produced by a 1GWe conventional reactor. It uniquely provides a method to close the nuclear fuel cycle for green nuclear energy.

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

DOE PAGES

An, F. P.; Balantekin, A. B.; Band, H. R.; ...

2017-06-19

Here, the Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW th reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F 239 from 0.25 to 0.35, Daya Bay measures an average IBD yield ¯σf of (5.90±0.13)×10 –43 cm 2/fission and a fuel-dependent variation in the IBDmore » yield, dσ f/dF 239, of (–1.86±0.18)×10 –43 cm 2/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10 –43 cm 2/fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.« less

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay.

PubMed

An, F P; Balantekin, A B; Band, H R; Bishai, M; Blyth, S; Cao, D; Cao, G F; Cao, J; Chan, Y L; Chang, J F; Chang, Y; Chen, H S; Chen, Q Y; Chen, S M; Chen, Y X; Chen, Y; Cheng, J; Cheng, Z K; Cherwinka, J J; Chu, M C; Chukanov, A; Cummings, J P; Ding, Y Y; Diwan, M V; Dolgareva, M; Dove, J; Dwyer, D A; Edwards, W R; Gill, R; Gonchar, M; Gong, G H; Gong, H; Grassi, M; Gu, W Q; Guo, L; Guo, X H; Guo, Y H; Guo, Z; Hackenburg, R W; Hans, S; He, M; Heeger, K M; Heng, Y K; Higuera, A; Hsiung, Y B; Hu, B Z; Hu, T; Huang, E C; Huang, H X; Huang, X T; Huang, Y B; Huber, P; Huo, W; Hussain, G; Jaffe, D E; Jen, K L; Ji, X P; Ji, X L; Jiao, J B; Johnson, R A; Jones, D; Kang, L; Kettell, S H; Khan, A; Kohn, S; Kramer, M; Kwan, K K; Kwok, M W; Langford, T J; Lau, K; Lebanowski, L; Lee, J; Lee, J H C; Lei, R T; Leitner, R; Leung, J K C; Li, C; Li, D J; Li, F; Li, G S; Li, Q J; Li, S; Li, S C; Li, W D; Li, X N; Li, X Q; Li, Y F; Li, Z B; Liang, H; Lin, C J; Lin, G L; Lin, S; Lin, S K; Lin, Y-C; Ling, J J; Link, J M; Littenberg, L; Littlejohn, B R; Liu, J L; Liu, J C; Loh, C W; Lu, C; Lu, H Q; Lu, J S; Luk, K B; Ma, X Y; Ma, X B; Ma, Y Q; Malyshkin, Y; Martinez Caicedo, D A; McDonald, K T; McKeown, R D; Mitchell, I; Nakajima, Y; Napolitano, J; Naumov, D; Naumova, E; Ngai, H Y; Ochoa-Ricoux, J P; Olshevskiy, A; Pan, H-R; Park, J; Patton, S; Pec, V; Peng, J C; Pinsky, L; Pun, C S J; Qi, F Z; Qi, M; Qian, X; Qiu, R M; Raper, N; Ren, J; Rosero, R; Roskovec, B; Ruan, X C; Steiner, H; Stoler, P; Sun, J L; Tang, W; Taychenachev, D; Treskov, K; Tsang, K V; Tull, C E; Viaux, N; Viren, B; Vorobel, V; Wang, C H; Wang, M; Wang, N Y; Wang, R G; Wang, W; Wang, X; Wang, Y F; Wang, Z; Wang, Z; Wang, Z M; Wei, H Y; Wen, L J; Whisnant, K; White, C G; Whitehead, L; Wise, T; Wong, H L H; Wong, S C F; Worcester, E; Wu, C-H; Wu, Q; Wu, W J; Xia, D M; Xia, J K; Xing, Z Z; Xu, J L; Xu, Y; Xue, T; Yang, C G; Yang, H; Yang, L; Yang, M S; Yang, M T; Yang, Y Z; Ye, M; Ye, Z; Yeh, M; Young, B L; Yu, Z Y; Zeng, S; Zhan, L; Zhang, C; Zhang, C C; Zhang, H H; Zhang, J W; Zhang, Q M; Zhang, R; Zhang, X T; Zhang, Y M; Zhang, Y X; Zhang, Y M; Zhang, Z J; Zhang, Z Y; Zhang, Z P; Zhao, J; Zhou, L; Zhuang, H L; Zou, J H

2017-06-23

The Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW_{th} reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective ^{239}Pu fission fractions F_{239} from 0.25 to 0.35, Daya Bay measures an average IBD yield σ[over ¯]_{f} of (5.90±0.13)×10^{-43}  cm^{2}/fission and a fuel-dependent variation in the IBD yield, dσ_{f}/dF_{239}, of (-1.86±0.18)×10^{-43}  cm^{2}/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the ^{239}Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes ^{235}U, ^{239}Pu, ^{238}U, and ^{241}Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10^{-43}  cm^{2}/fission have been determined for the two dominant fission parent isotopes ^{235}U and ^{239}Pu. A 7.8% discrepancy between the observed and predicted ^{235}U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

NASA Astrophysics Data System (ADS)

An, F. P.; Balantekin, A. B.; Band, H. R.; Bishai, M.; Blyth, S.; Cao, D.; Cao, G. F.; Cao, J.; Chan, Y. L.; Chang, J. F.; Chang, Y.; Chen, H. S.; Chen, Q. Y.; Chen, S. M.; Chen, Y. X.; Chen, Y.; Cheng, J.; Cheng, Z. K.; Cherwinka, J. J.; Chu, M. C.; Chukanov, A.; Cummings, J. P.; Ding, Y. Y.; Diwan, M. V.; Dolgareva, M.; Dove, J.; Dwyer, D. A.; Edwards, W. R.; Gill, R.; Gonchar, M.; Gong, G. H.; Gong, H.; Grassi, M.; Gu, W. Q.; Guo, L.; Guo, X. H.; Guo, Y. H.; Guo, Z.; Hackenburg, R. W.; Hans, S.; He, M.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hsiung, Y. B.; Hu, B. Z.; Hu, T.; Huang, E. C.; Huang, H. X.; Huang, X. T.; Huang, Y. B.; Huber, P.; Huo, W.; Hussain, G.; Jaffe, D. E.; Jen, K. L.; Ji, X. P.; Ji, X. L.; Jiao, J. B.; Johnson, R. A.; Jones, D.; Kang, L.; Kettell, S. H.; Khan, A.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, J. H. C.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, S.; Li, S. C.; Li, W. D.; Li, X. N.; Li, X. Q.; Li, Y. F.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, S.; Lin, S. K.; Lin, Y.-C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, J. L.; Liu, J. C.; Loh, C. W.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, K. B.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Malyshkin, Y.; Martinez Caicedo, D. A.; McDonald, K. T.; McKeown, R. D.; Mitchell, I.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Ngai, H. Y.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pan, H.-R.; Park, J.; Patton, S.; Pec, V.; Peng, J. C.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Qiu, R. M.; Raper, N.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Steiner, H.; Stoler, P.; Sun, J. L.; Tang, W.; Taychenachev, D.; Treskov, K.; Tsang, K. V.; Tull, C. E.; Viaux, N.; Viren, B.; Vorobel, V.; Wang, C. H.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Wei, H. Y.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, C.-H.; Wu, Q.; Wu, W. J.; Xia, D. M.; Xia, J. K.; Xing, Z. Z.; Xu, J. L.; Xu, Y.; Xue, T.; Yang, C. G.; Yang, H.; Yang, L.; Yang, M. S.; Yang, M. T.; Yang, Y. Z.; Ye, M.; Ye, Z.; Yeh, M.; Young, B. L.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, C. C.; Zhang, H. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, R.; Zhang, X. T.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhou, L.; Zhuang, H. L.; Zou, J. H.; Daya Bay Collaboration

2017-06-01

The Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 G Wth reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F239 from 0.25 to 0.35, Daya Bay measures an average IBD yield σ¯f of (5.90 ±0.13 )×10-43 cm2/fission and a fuel-dependent variation in the IBD yield, d σf/d F239, of (-1.86 ±0.18 )×10-43 cm2/fission . This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1 σ . This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17 ±0.17 ) and (4.27 ±0.26 )×10-43 cm2 /fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

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

An, F. P.; Balantekin, A. B.; Band, H. R.

Here, the Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW th reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F 239 from 0.25 to 0.35, Daya Bay measures an average IBD yield ¯σf of (5.90±0.13)×10 –43 cm 2/fission and a fuel-dependent variation in the IBDmore » yield, dσ f/dF 239, of (–1.86±0.18)×10 –43 cm 2/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10 –43 cm 2/fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.« less

Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

NASA Astrophysics Data System (ADS)

Shi, Xue-Ming; Peng, Xian-Jue

2016-09-01

Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

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

Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

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

Bucknor, Matthew D.; Grabaskas, David; Brunett, Acacia J.

2016-01-01

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologiesmore » for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Centering on an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive reactor cavity cooling system following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. While this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability for the reactor cavity cooling system (and the reactor system in general) to the postulated transient event.« less

Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

DOE PAGES

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; ...

2017-01-24

We report that many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has beenmore » examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Lastly, although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.« less

Modeling Fission Product Sorption in Graphite Structures

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

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 distributionsmore » 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

Modeling of Fission Gas Release in UO2

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

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 Mathcadmore » [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].« less

Total prompt γ-ray emission in fission

NASA Astrophysics Data System (ADS)

Wu, C. Y.; Chyzh, A.; Kwan, E.; Henderson, R. A.; Bredeweg, T. A.; Haight, R. C.; Hayes-Sterbenz, A. C.; Lee, H. Y.; O'Donnell, J. M.; Ullmann, J. L.

2017-09-01

The total prompt γ-ray energy distributions were measured for the neutron-induced fission of 235U, 239,241Pu at incident neutron energy of 0.025 eV-100 keV, and the spontaneous fission of 252Cf using the Detector for Advanced Neutron Capture Experiments (DANCE) array in coincidence with the detection of fission fragments by a parallel-plate avalanche counter. Corrections were made to the measured distribution by unfolding the two-dimension spectrum of total prompt γ-ray energy vs multiplicity using a simulated DANCE response matrix. A summary of this work is presented with the emphasis on the comparison of total prompt fission γ-ray energy between our results and previous ones. The mean values of the total prompt γ-ray energy ⟨Eγ,tot⟩, determined from the unfolded distributions, are ˜20% higher than those derived from measurements using single γ-ray detector for all the fissile nuclei studied.

Research Activities in Fission Chamber Modeling in Support of the Nuclear Energy Industry

NASA Astrophysics Data System (ADS)

Jammes, C.; Filliatre, P.; Geslot, B.; Oriol, L.; Berhouet, F.; Villard, J.-F.; Vermeeren, L.

2010-12-01

Fission chambers are widely used in the nuclear industry. As an example, they play a major role in the control of any fission reactor and are thus regarded as a key component for ensuring their safety. They are also employed in the material testing reactors for monitoring irradiations. We have recently started a research program, the objective of which is to improve the performance of those neutron detectors in terms of lifetime, calibration, and online diagnosis. In this paper, we present several studies carried out in order to model the signal delivered by a fission chamber. First, the simulation of the deposit evolution allowed us to select the most appropriate fissile material for a given spectrum and fluence. Second, we studied the impact of the bias voltage and filling gas characteristics on the charge collection time. Finally, the simulation of a pulse signal prior to amplification showed how it is important to have a satisfactory knowledge of the energy for creating ion pairs to accurately assess the signal in current or Campbelling mode.

Feasibility study of a magnetic fusion production reactor

NASA Astrophysics Data System (ADS)

Moir, R. W.

1986-12-01

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

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Advance notification of shipment of irradiated reactor... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... required under this section for shipments of irradiated reactor fuel in quantities less than that subject...

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Advance notification of shipment of irradiated reactor... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... required under this section for shipments of irradiated reactor fuel in quantities less than that subject...

FALSTAFF: A new tool for fission studies

NASA Astrophysics Data System (ADS)

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

2013-12-01

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

Development and Analysis of Cold Trap for Use in Fission Surface Power-Primary Test Circuit

NASA Technical Reports Server (NTRS)

Wolfe, T. M.; Dervan, C. A.; Pearson, J. B.; Godfroy, T. J.

2012-01-01

The design and analysis of a cold trap proposed for use in the purification of circulated eutectic sodium potassium (NaK-78) loops is presented. The cold trap is designed to be incorporated into the Fission Surface Power-Primary Test Circuit (FSP-PTC), which incorporates a pumped NaK loop to simulate in-space nuclear reactor-based technology using non-nuclear test methodology as developed by the Early Flight Fission-Test Facility. The FSP-PTC provides a test circuit for the development of fission surface power technology. This system operates at temperatures that would be similar to those found in a reactor (500-800 K). By dropping the operating temperature of a specified percentage of NaK flow through a bypass containing a forced circulation cold trap, the NaK purity level can be increased by precipitating oxides from the NaK and capturing them within the cold trap. This would prevent recirculation of these oxides back through the system, which may help prevent corrosion.

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.

Regulatory Risk Reduction for Advanced Reactor Technologies – FY2016 Status and Work Plan Summary

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

Moe, Wayne Leland

2016-08-01

Millions of public and private sector dollars have been invested over recent decades to realize greater efficiency, reliability, and the inherent and passive safety offered by advanced nuclear reactor technologies. However, a major challenge in experiencing those benefits resides in the existing U.S. regulatory framework. This framework governs all commercial nuclear plant construction, operations, and safety issues and is highly large light water reactor (LWR) technology centric. The framework must be modernized to effectively deal with non-LWR advanced designs if those designs are to become part of the U.S energy supply. The U.S. Department of Energy’s (DOE) Advanced Reactor Technologiesmore » (ART) Regulatory Risk Reduction (RRR) initiative, managed by the Regulatory Affairs Department at the Idaho National Laboratory (INL), is establishing a capability that can systematically retire extraneous licensing risks associated with regulatory framework incompatibilities. This capability proposes to rely heavily on the perspectives of the affected regulated community (i.e., commercial advanced reactor designers/vendors and prospective owner/operators) yet remain tuned to assuring public safety and acceptability by regulators responsible for license issuance. The extent to which broad industry perspectives are being incorporated into the proposed framework makes this initiative unique and of potential benefit to all future domestic non-LWR applicants« less

Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

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

Honma, George

The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will bemore » used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).« less

Energy from nuclear fission()

NASA Astrophysics Data System (ADS)

Ripani, M.

2015-08-01

The main features of nuclear fission as physical phenomenon will be revisited, emphasizing its peculiarities with respect to other nuclear reactions. Some basic concepts underlying the operation of nuclear reactors and the main types of reactors will be illustrated, including fast reactors, showing the most important differences among them. The nuclear cycle and radioactive-nuclear-waste production will be also discussed, along with the perspectives offered by next generation nuclear assemblies being proposed. The current situation of nuclear power in the world, its role in reducing carbon emission and the available resources will be briefly illustrated.

Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

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

Tan, Lizhen; Yang, Ying; Sridharan, K.

2015-12-01

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computationalmore » tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe 2+) irradiation.« less

Data summary report for fission product release test VI-5

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

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

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

75 FR 10840 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-09

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on ABWR will... would result in major inconvenience. Dated: March 3, 2010. Antonio F. Dias, Chief, Reactor Safety Branch...

Plant maintenance and advanced reactors issue, 2004

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

Agnihotri, Newal

2004-09-15

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Optimism about the future of nuclear power, by Ruth G. Shaw, Duke Power Company; Licensed in three countries, by GE Energy; Enhancing public acceptance, by Westinghouse Electric Company; Standardized MOV program, by Ted Neckowicz, Exelon; Inservice testing, by Steven Unikewicz, U.S. Nuclear Regulatory Commission; Asian network for education, Fatimah Mohd Amin, Malaysian Institute for Nuclear Technology Research; and, Cooling water intake optimization, by Jeffrey M. Jones and Bert Mayer, P.E., Framatome ANP.

Kilowatt-Class Fission Power Systems for Science and Human Precursor Missions

NASA Technical Reports Server (NTRS)

Mason, Lee S.; Gibson, Marc Andrew; Poston, Dave

2013-01-01

Nuclear power provides an enabling capability for NASA missions that might otherwise be constrained by power availability, mission duration, or operational robustness. NASA and the Department of Energy (DOE) are developing fission power technology to serve a wide range of future space uses. Advantages include lower mass, longer life, and greater mission flexibility than competing power system options. Kilowatt-class fission systems, designated "Kilopower," were conceived to address the need for systems to fill the gap above the current 100-W-class radioisotope power systems being developed for science missions and below the typical 100-k We-class reactor power systems being developed for human exploration missions. This paper reviews the current fission technology project and examines some Kilopower concepts that could be used to support future science missions or human precursors.

Kilowatt-Class Fission Power Systems for Science and Human Precursor Missions

NASA Technical Reports Server (NTRS)

Mason, Lee; Gibson, Marc; Poston, Dave

2013-01-01

Nuclear power provides an enabling capability for NASA missions that might otherwise be constrained by power availability, mission duration, or operational robustness. NASA and the Department of Energy (DOE) are developing fission power technology to serve a wide range of future space uses. Advantages include lower mass, longer life, and greater mission flexibility than competing power system options. Kilowatt-class fission systems, designated "Kilopower," were conceived to address the need for systems to fill the gap above the current 100-Wclass radioisotope power systems being developed for science missions and below the typical 100-kWe-class reactor power systems being developed for human exploration missions. This paper reviews the current fission technology project and examines some Kilopower concepts that could be used to support future science missions or human precursors.

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

NASA Technical Reports Server (NTRS)

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

2010-01-01

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

Advanced ceramic materials for next-generation nuclear applications

NASA Astrophysics Data System (ADS)

Marra, John

2011-10-01

The nuclear industry is at the eye of a 'perfect storm' with fuel oil and natural gas prices near record highs, worldwide energy demands increasing at an alarming rate, and increased concerns about greenhouse gas (GHG) emissions that have caused many to look negatively at long-term use of fossil fuels. This convergence of factors has led to a growing interest in revitalization of the nuclear power industry within the United States and across the globe. Many are surprised to learn that nuclear power provides approximately 20% of the electrical power in the US and approximately 16% of the world-wide electric power. With the above factors in mind, world-wide over 130 new reactor projects are being considered with approximately 25 new permit applications in the US. Materials have long played a very important role in the nuclear industry with applications throughout the entire fuel cycle; from fuel fabrication to waste stabilization. As the international community begins to look at advanced reactor systems and fuel cycles that minimize waste and increase proliferation resistance, materials will play an even larger role. Many of the advanced reactor concepts being evaluated operate at high-temperature requiring the use of durable, heat-resistant materials. Advanced metallic and ceramic fuels are being investigated for a variety of Generation IV reactor concepts. These include the traditional TRISO-coated particles, advanced alloy fuels for 'deep-burn' applications, as well as advanced inert-matrix fuels. In order to minimize wastes and legacy materials, a number of fuel reprocessing operations are being investigated. Advanced materials continue to provide a vital contribution in 'closing the fuel cycle' by stabilization of associated low-level and high-level wastes in highly durable cements, ceramics, and glasses. Beyond this fission energy application, fusion energy will demand advanced materials capable of withstanding the extreme environments of high

Thermomechanical analysis of fast-burst reactors

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

Miller, J.D.

1994-08-01

Fast-burst reactors are designed to provide intense, short-duration pulses of neutrons. The fission reaction also produces extreme time-dependent heating of the nuclear fuel. An existing transient-dynamic finite element code was modified specifically to compute the time-dependent stresses and displacements due to thermal shock loads of reactors. Thermomechanical analysis was then applied to determine structural feasibility of various concepts for an EDNA-type reactor and to optimize the mechanical design of the new SPR III-M reactor.

FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Abbott, W.E.; Balent, R.

1958-09-16

A fuel element design to facilitate breeding reactor fuel is described. The fuel element is comprised of a coatainer, a central core of fertile material in the container, a first bonding material surrounding the core, a sheet of fissionable material immediately surrounding the first bonding material, and a second bonding material surrounding the fissionable material and being in coniact with said container.

Experimental programme on absolute fission fragment yields with the lohengrin spectrometer: New optical and statistical methodologies

NASA Astrophysics Data System (ADS)

Abdelaziz, Chebboubi; Grégoire, Kessedjian; Olivier, Serot; Sylvain, Julien-Laferriere; Christophe, Sage; Florence, Martin; Olivier, Méplan; David, Bernard; Olivier, Litaize; Aurélien, Blanc; Herbert, Faust; Paolo, Mutti; Ulli, Köster; Alain, Letourneau; Thomas, Materna; Michal, Rapala

2017-09-01

The study of fission yields has a major impact on the characterization and understanding of the fission process and is mandatory for reactor applications. In the past with the LOHENGRIN spectrometer of the ILL, priority has been given for the studies in the light fission fragment mass range. The LPSC in collaboration with ILL and CEA has developed a measurement program on symmetric and heavy mass fission fragment distributions. The combination of measurements with ionisation chamber and Ge detectors is necessary to describe precisely the heavy fission fragment region in mass and charge. Recently, new measurements of fission yields and kinetic energy distributions are has been made on the 233U(nth,f) reaction. The focus of this work has been on the new optical and statistical methodology and the self-normalization of the data to provide new absolute measurements, independently of any libraries, and the associated experimental covariance matrix.

The performance of hafnium and gadolinium self powered neutron detectors in the TREAT reactor

NASA Astrophysics Data System (ADS)

Imel, G. R.; Hart, P. R.

1996-05-01

The use of gadolinium and hafnium self powered neutron detectors in a transient reactor is described in this paper. The detectors were calibrated to the fission rate of U-235 using calibrated fission chambers; the calibration factors were tested in two reactors in steady state and found to be consistent. Calibration of the detectors in transient reactor conditions was done by using uranium wires that were analyzed by radiochemistry techniques to determine total fissions during the transient. This was correlated to the time-integrated current of the detectors during the transient. A temperature correction factor was derived to account for self-shielding effects in the hafnium and gadolinium detectors. The dynamic response of the detectors under transient conditions was studied, and found to be excellent.

ENDF/B-IV fission-product files: summary of major nuclide data

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

England, T.R.; Schenter, R.E.

1975-09-01

The major fission-product parameters [sigma/sub th/, RI, tau/sub 1/2/, E- bar/sub $beta$/, E-bar/sub $gamma$/, E-bar/sub $alpha$/, decay and (n,$gamma$) branching, Q, and AWR] abstracted from ENDF/B-IV files for 824 nuclides are summarized. These data are most often requested by users concerned with reactor design, reactor safety, dose, and other sundry studies. The few known file errors are corrected to date. Tabular data are listed by increasing mass number. (auth)

Compact fission counter for DANCE

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

Wu, C Y; Chyzh, A; Kwan, E

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, additionalmore » 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

SUPERHEATING IN A BOILING WATER REACTOR

DOEpatents

Treshow, M.

1960-05-31

A boiling-water reactor is described in which the steam developed in the reactor is superheated in the reactor. This is accomplished by providing means for separating the steam from the water and passing the steam over a surface of the fissionable material which is not in contact with the water. Specifically water is boiled on the outside of tubular fuel elements and the steam is superheated on the inside of the fuel elements.

PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP

DOEpatents

Puechl, K.H.

1963-09-24

A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

Towards the high-accuracy determination of the 238U fission cross section at the threshold region at CERN - n_TOF

NASA Astrophysics Data System (ADS)

Diakaki, M.; Audouin, L.; Berthoumieux, E.; Calviani, M.; Colonna, N.; Dupont, E.; Duran, I.; Gunsing, F.; Leal-Cidoncha, E.; Le Naour, C.; Leong, L. S.; Mastromarco, M.; Paradela, C.; Tarrio, D.; Tassan-Got, L.; Aerts, G.; Altstadt, S.; Alvarez, H.; Alvarez-Velarde, F.; Andriamonje, S.; Andrzejewski, J.; Badurek, G.; Barbagallo, M.; Baumann, P.; Becares, V.; Becvar, F.; Belloni, F.; Berthier, B.; Billowes, J.; Boccone, V.; Bosnar, D.; Brugger, M.; Calvino, F.; Cano-Ott, D.; Capote, R.; Carrapiço, C.; Cennini, P.; Cerutti, F.; Chiaveri, E.; Chin, M.; Cortes, G.; Cortes-Giraldo, M. A.; Cosentino, L.; Couture, A.; Cox, J.; David, S.; Dillmann, I.; Domingo-Pardo, C.; Dressler, R.; Dridi, W.; Eleftheriadis, C.; Embid-Segura, M.; Ferrant, L.; Ferrari, A.; Finocchiaro, P.; Fraval, K.; Fujii, K.; Furman, W.; Ganesan, S.; Garcia, A. R.; Giubrone, G.; Gomez-Hornillos, M. B.; Goncalves, I. F.; Gonzalez-Romero, E.; Goverdovski, A.; Gramegna, F.; Griesmayer, E.; Guerrero, C.; Gurusamy, P.; Haight, R.; Heil, M.; Heinitz, S.; Igashira, M.; Isaev, S.; Jenkins, D. G.; Jericha, E.; Kadi, Y.; Kaeppeler, F.; Karadimos, D.; Karamanis, D.; Kerveno, M.; Ketlerov, V.; Kivel, N.; Kokkoris, M.; Konovalov, V.; Krticka, M.; Kroll, J.; Lampoudis, C.; Langer, C.; Lederer, C.; Leeb, H.; Lo Meo, S.; Losito, R.; Lozano, M.; Manousos, A.; Marganiec, J.; Martinez, T.; Marrone, S.; Massimi, C.; Mastinu, P.; Mendoza, E.; Mengoni, A.; Milazzo, P. M.; Mingrone, F.; Mirea, M.; Mondelaers, W.; Moreau, C.; Mosconi, M.; Musumarra, A.; O'Brien, S.; Pancin, J.; Patronis, N.; Pavlik, A.; Pavlopoulos, P.; Perkowski, J.; Perrot, L.; Pigni, M. T.; Plag, R.; Plompen, A.; Plukis, L.; Poch, A.; Pretel, C.; Praena, J.; Quesada, J.; Rauscher, T.; Reifarth, R.; Riego, A.; Roman, F.; Rudolf, G.; Rubbia, C.; Rullhusen, P.; Salgado, J.; Santos, C.; Sarchiapone, L.; Sarmento, R.; Saxena, A.; Schillebeeckx, P.; Schmidt, S.; Schumann, D.; Stephan, C.; Tagliente, G.; Tain, J. L.; Tavora, L.; Terlizzi, R.; Tsinganis, A.; Valenta, S.; Vannini, G.; Variale, V.; Vaz, P.; Ventura, A.; Versaci, R.; Vermeulen, M. J.; Villamarin, D.; Vincente, M. C.; Vlachoudis, V.; Vlastou, R.; Voss, F.; Wallner, A.; Walter, S.; Ware, T.; Weigand, M.; Weiß, C.; Wiesher, M.; Wisshak, K.; Wright, T.; Zugec, P.

2016-03-01

The 238U fission cross section is an international standard beyond 2 MeV where the fission plateau starts. However, due to its importance in fission reactors, this cross-section should be very accurately known also in the threshold region below 2 MeV. The 238U fission cross section has been measured relative to the 235U fission cross section at CERN - n_TOF with different detection systems. These datasets have been collected and suitably combined to increase the counting statistics in the threshold region from about 300 keV up to 3 MeV. The results are compared with other experimental data, evaluated libraries, and the IAEA standards.

Basic requirements for a 1000-MW(electric) class tokamak fusion-fission hybrid reactor and its blanket concept

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

Hatayama, Ariyoshi; Ogasawara, Masatada; Yamauchi, Michinori

1994-08-01

Plasma size and other basic performance parameters for 1000-MW(electric) power production are calculated with the blanket energy multiplication factor, the M value, as a parameter. The calculational model is base don the International Thermonuclear Experimental Reactor (ITER) physics design guidelines and includes overall plant power flow. Plasma size decreases as the M value increases. However, the improvement in the plasma compactness and other basic performance parameters, such as the total plant power efficiency, becomes saturated above the M = 5 to 7 range. THus, a value in the M = 5 to 7 range is a reasonable choice for 1000-MW(electric)more » hybrids. Typical plasma parameters for 1000-MW(electric) hybrids with a value of M = 7 are a major radius of R = 5.2 m, minor radius of a = 1.7 m, plasma current of I{sub p} = 15 MA, and toroidal field on the axis of B{sub o} = 5 T. The concept of a thermal fission blanket that uses light water as a coolant is selected as an attractive candidate for electricity-producing hybrids. An optimization study is carried out for this blanket concept. The result shows that a compact, simple structure with a uniform fuel composition for the fissile region is sufficient to obtain optimal conditions for suppressing the thermal power increase caused by fuel burnup. The maximum increase in the thermal power is +3.2%. The M value estimated from the neutronics calculations is {approximately}7.0, which is confirmed to be compatible with the plasma requirement. These studies show that it is possible to use a tokamak fusion core with design requirements similar to those of ITER for a 1000-MW(electric) power reactor that uses existing thermal reactor technology for the blanket. 30 refs., 22 figs., 4 tabs.« less

Advanced Space Nuclear Reactors from Fiction to Reality

NASA Astrophysics Data System (ADS)

Popa-Simil, L.

The advanced nuclear power sources are used in a large variety of science fiction movies and novels, but their practical development is, still, in its early conceptual stages, some of the ideas being confirmed by collateral experiments. The novel reactor concept uses the direct conversion of nuclear energy into electricity, has electronic control of reactivity, being surrounded by a transmutation blanket and very thin shielding being small and light that at its very limit may be suitable to power an autonomously flying car. It also provides an improved fuel cycle producing minimal negative impact to environment. The key elements started to lose the fiction attributes, becoming viable actual concepts and goals for the developments to come, and on the possibility to achieve these objectives started to become more real because the theory shows that using the novel nano-technologies this novel reactor might be achievable in less than a century.

Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

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

Fromm, Bradley; Hauch, Benjamin; Sridharan, Kumar

2016-12-01

A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulationmore » tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.« less

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.

Interim waste storage for the Integral Fast Reactor

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

Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

1991-01-01

The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes thatmore » are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig.« less

Heterogeneous sodium fast reactor designed for transmuting minor actinide waste isotopes into plutonium fuel

NASA Astrophysics Data System (ADS)

Bays, Samuel Eugene

2008-10-01

In the past several years there has been a renewed interest in sodium fast reactor (SFR) technology for the purpose of destroying transuranic waste (TRU) produced by light water reactors (LWR). The utility of SFRs as waste burners is due to the fact that higher neutron energies allow all of the actinides, including the minor actinides (MA), to contribute to fission. It is well understood that many of the design issues of LWR spent nuclear fuel (SNF) disposal in a geologic repository are linked to MAs. Because the probability of fission for essentially all the "non-fissile" MAs is nearly zero at low neutron energies, these isotopes act as a neutron capture sink in most thermal reactor systems. Furthermore, because most of the isotopes produced by these capture reactions are also non-fissile, they too are neutron sinks in most thermal reactor systems. Conversely, with high neutron energies, the MAs can produce neutrons by fast fission. Additionally, capture reactions transmute the MAs into mostly plutonium isotopes, which can fission more readily at any energy. The transmutation of non-fissile into fissile atoms is the premise of the plutonium breeder reactor. In a breeder reactor, not only does the non-fissile "fertile" U-238 atom contribute fast fission neutrons, but also transmutes into fissile Pu-239. The fissile value of the plutonium produced by MA transmutation can only be realized in fast neutron spectra. This is due to the fact that the predominate isotope produced by MA transmutation, Pu-238, is itself not fissile. However, the Pu-238 fission cross section is significantly larger than the original transmutation parent, predominately: Np-237 and Am-241, in the fast energy range. Also, Pu-238's fission cross section and fission-to-capture ratio is almost as high as that of fissile Pu-239 in the fast neutron spectrum. It is also important to note that a neutron absorption in Pu-238, that does not cause fission, will instead produce fissile Pu-239. Given this

Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and

Neutronic reactor

DOEpatents

Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

1983-01-01

A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

Developing the Pulsed Fission-Fusion (PuFF) Engine

NASA Technical Reports Server (NTRS)

Adams, Robert B.; Cassibry, Jason; Bradley, David; Fabisinski, Leo; Statham, Geoffrey

2014-01-01

In September 2013 the NASA Innovative Advanced Concept (NIAC) organization awarded a phase I contract to the PuFF team. Our phase 1 proposal researched a pulsed fission-fusion propulsion system that compressed a target of deuterium (D) and tritium (T) as a mixture in a column, surrounded concentrically by Uranium. The target is surrounded by liquid lithium. A high power current would flow down the liquid lithium and the resulting Lorentz force would compress the column by roughly a factor of 10. The compressed column would reach criticality and a combination of fission and fusion reactions would occur. Our Phase I results, summarized herein, review our estimates of engine and vehicle performance, our work to date to model the fission-fusion reaction, and our initial efforts in experimental analysis.

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.; ...

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

NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.

1957-09-17

A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.

Method of fission heat flux determination from experimental data

DOEpatents

Paxton, Frank A.

1999-01-01

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

Comparative evaluation of solar, fission, fusion, and fossil energy resources, part 3

NASA Technical Reports Server (NTRS)

Clement, J. D.; Reupke, W. A.

1974-01-01

The role of nuclear fission reactors in becoming an important power source in the world is discussed. The supply of fissile nuclear fuel will be severely depleted by the year 2000. With breeder reactors the world supply of uranium could last thousands of years. However, breeder reactors have problems of a large radioactive inventory and an accident potential which could present an unacceptable hazard. Although breeder reactors afford a possible solution to the energy shortage, their ultimate role will depend on demonstrated safety and acceptable risks and environmental effects. Fusion power would also be a long range, essentially permanent, solution to the world's energy problem. Fusion appears to compare favorably with breeders in safety and environmental effects. Research comparing a controlled fusion reactor with the breeder reactor in solving our long range energy needs is discussed.

Acoustic emission signal processing technique to characterize reactor in-pile phenomena

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

Agarwal, Vivek, E-mail: vivek.agarwal@inl.gov; Tawfik, Magdy S., E-mail: magdy.tawfik@inl.gov; Smith, James A., E-mail: james.smith@inl.gov

2015-03-31

Existing and developing advanced sensor technologies and instrumentation will allow non-intrusive in-pile measurement of temperature, extension, and fission gases when coupled with advanced signal processing algorithms. The transmitted measured sensor signals from inside to the outside of containment structure are corrupted by noise and are attenuated, thereby reducing the signal strength and the signal-to-noise ratio. Identification and extraction of actual signal (representative of an in-pile phenomenon) is a challenging and complicated process. In the paper, empirical mode decomposition technique is utilized to reconstruct actual sensor signal by partially combining intrinsic mode functions. Reconstructed signal will correspond to phenomena and/or failuremore » modes occurring inside the reactor. In addition, it allows accurate non-intrusive monitoring and trending of in-pile phenomena.« less

ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

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

E. Blanford; E. Keldrauk; M. Laufer

2010-09-20

Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement,more » and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction

A ``NEW'' Solid-Core Reactor Fuel Form that Maximizes the Performance of Nuclear Thermal and Electric Rockets

NASA Astrophysics Data System (ADS)

Rom, Frank E.; Finnegan, Patrick M.

1994-07-01

The ``NEW'' solid-core fuel form is the old Vapor Transport (VT) fuel pin investigated at NASA about 30 years ago. It is simply a tube sealed at both ends partially filled with UO2. During operation the UO2 forms an annular layer on the inside of the tube by vaporization and condensation. This form is an ideal structure for overall strength and retention of fission products. All of the structural material lies between the fuel (including fission products) and the reactor coolant. The isothermal inside fuel surface temperature that results from the vaporization and condensation of fuel during operation eliminates hotspots, significantly increasing the design fuel pin surface temperature. For NTP, W-UO2 fuel pins yield higher operating temperatures than for other fuel forms, because W has about a ten-fold lower vaporization rate compared to any other known material. The use of perigee propulsion using W-UO2 fuel pins can result in a more than ten-fold reduction in reactor power. Lower reactor power, together with zero fission product release potential, and the simplicity of fabrication of VT fuel pins should greatly simplify and reduce the cost of development of NTP. For NEP, VT fuel pins can increase fast neutron spectrum reactor life with no fission product release. Thermal spectrum NEP reactors using W184 or Mo VT fuel pins, with only small amounts of high neutron absorbing additives, offer benefits because of much lower fissionable fuel requirements. The VT fuel pin has application to commercial power reactors with similar benefits.

Advanced fuels modeling: Evaluating the steady-state performance of carbide fuel in helium-cooled reactors using FRAPCON 3.4

NASA Astrophysics Data System (ADS)

Hallman, Luther, Jr.

Uranium carbide (UC) has long been considered a potential alternative to uranium dioxide (UO2) fuel, especially in the context of Gen IV gas-cooled reactors. It has shown promise because of its high uranium density, good irradiation stability, and especially high thermal conductivity. Despite its many benefits, UC is known to swell at a rate twice that of UO2. However, the swelling phenomenon is not well understood, and we are limited to a weak empirical understanding of the swelling mechanism. One suggested cladding for UC is silicon carbide (SiC), a ceramic that demonstrates a number of desirable properties. Among them are an increased corrosion resistance, high mechanical strength, and irradiation stability. However, with increased temperatures, SiC exhibits an extremely brittle nature. The brittle behavior of SiC is not fully understood and thus it is unknown how SiC would respond to the added stress of a swelling UC fuel. To better understand the interaction between these advanced materials, each has been implemented into FRAPCON, the preferred fuel performance code of the Nuclear Regulatory Commission (NRC); additionally, the material properties for a helium coolant have been incorporated. The implementation of UC within FRAPCON required the development of material models that described not only the thermophysical properties of UC, such as thermal conductivity and thermal expansion, but also models for the swelling, densification, and fission gas release associated with the fuel's irradiation behavior. This research is intended to supplement ongoing analysis of the performance and behavior of uranium carbide and silicon carbide in a helium-cooled reactor.

Utilizing Fission Technology to Enable Rapid and Affordable Access to any Point in the Solar System

NASA Technical Reports Server (NTRS)

Houts, Mike; Bonometti, Joe; Morton, Jeff; Hrbud, Ivana; Bitteker, Leo; VanDyke, Melissa; Godfroy, T.; Pedersen, K.; Dobson, C.; Patton, B.;

COMPARTMENTED REACTOR FUEL ELEMENT

DOEpatents

Cain, F.M. Jr.

1962-09-11

A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

FUEL ELEMENTS FOR NEUTRONIC REACTORS

DOEpatents

Foote, F.G.; Jette, E.R.

1963-05-01

A fuel element for a nuclear reactor is described that consists of a jacket containing a unitary core of fissionable material and a filling of a metal of the group consisting of sodium and sodium-potassium alloys. (AEC)

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.

The SPIDER fission fragment spectrometer for fission product yield measurements

DOE PAGES

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

2015-04-01

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

Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor

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

Daniel M. Wachs; Richard G. Ambrosek; Gray Chang

2006-10-01

Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progressmore » toward element testing will be reviewed.« less

Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors

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

William Richins; Stephen Novascone; Cheryl O'Brien

Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors William Richins1, Stephen Novascone1, and Cheryl O’Brien1 1Idaho National Laboratory, US Dept. of Energy, Idaho Falls, Idaho, USA, e-mail: William.Richins@inl.gov The Idaho National Laboratory (INL, USA) and IASMiRT sponsored an international forum Nov 5-6, 2008 in Porvoo, Finland for nuclear industry, academic, and regulatory representatives to identify structural issues in current and future advanced reactor design, especially for extreme conditions and external threats. The purpose of this Topical Workshop was to articulate research, engineering, and regulatory Code development needs. The topics addressed by the Workshop were selectedmore » to address critical industry needs specific to advanced reactor structures that have long lead times and can be the subject of future SMiRT technical sessions. The topics were; 1) structural/materials needs for extreme conditions and external threats in contemporary (Gen. III) and future (Gen. IV and NGNP) advanced reactors and 2) calibrating simulation software and methods that address topic 1 The workshop discussions and research needs identified are presented. The Workshop successfully produced interactive discussion on the two topics resulting in a list of research and technology needs. It is recommended that IASMiRT communicate the results of the discussion to industry and researchers to encourage new ideas and projects. In addition, opportunities exist to retrieve research reports and information that currently exists, and encourage more international cooperation and collaboration. It is recommended that IASMiRT continue with an off-year workshop series on select topics.« less

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

NASA Astrophysics Data System (ADS)

Van Dyke, Melissa; Martin, James

2005-02-01

The NASA Marshall Space Flight Center's Early Flight Fission Test Facility (EFF-TF), provides a facility to experimentally evaluate nuclear reactor related thermal hydraulic issues through the use of non-nuclear testing. This facility provides a cost effective method to evaluate concepts/designs and support mitigation of developmental risk. Electrical resistance thermal simulators can be used to closely mimic the heat deposition of the fission process, providing axial and radial profiles. A number of experimental and design programs were underway in 2004 which include the following. Initial evaluation of the Department of Energy Los Alamos National Laboratory 19 module stainless steel/sodium heat pipe reactor with integral gas heat exchanger was operated at up to 17.5 kW of input power at core temperatures of 1000 K. A stainless steel sodium heat pipe module was placed through repeated freeze/thaw cyclic testing accumulating over 200 restarts to a temperature of 1000 K. Additionally, the design of a 37- pin stainless steel pumped sodium/potassium (NaK) loop was finalized and components procured. Ongoing testing at the EFF-TF is geared towards facilitating both research and development necessary to support future decisions regarding potential use of space nuclear systems for space exploration. All efforts are coordinated with DOE laboratories, industry, universities, and other NASA centers. This paper describes some of the 2004 efforts.

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

NASA Technical Reports Server (NTRS)

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

2013-01-01

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

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Fielding, R.; Gan, J.

2007-09-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

Future research program on prompt γ-ray emission in nuclear fission

NASA Astrophysics Data System (ADS)

Oberstedt, S.; Billnert, R.; Hambsch, F.-J.; Lebois, M.; Oberstedt, A.; Wilson, J. N.

2015-12-01

In recent years the measurement of prompt fission γ-ray spectra (PFGS) has gained renewed interest, after about forty years since the first comprehensive studies of the reactions 235U(n th , f), 239Pu(n th ,f) and 252Cf(sf). The renaissance was initiated by requests for new values especially for γ-ray multiplicity and average total energy release per fission in neutron-induced fission of 235U and 239Pu. Both isotopes are considered the most important ones with respect to the modeling of innovative cores required for the Generation-IV reactors, the majority working with fast neutrons. During the last 5 years we have conducted a systematic study of spectral data for thermal-neutron-induced fission on 235U and 241Pu as well as for the spontaneous fission of 252Cf with unprecedented accuracy. From the new data we conclude that those reactions do not considerably contribute to the observed heat excess and suspect other reactions playing a significant role. Possible contributions may originate from fast-neutron-induced reactions on 238U, which is largely present in the fuel, or from γ-induced fission from neutron capture in the construction material. A first experiment campaign on prompt γ-ray emission from fast-neutron-induced fission on 235,238U was successfully performed in order to test our assumptions. In the following we attempt to summarize, what has been done in the field to date, and to motivate future measurement campaigns exploiting dedicated neutron and photon beams as well as upcoming highly efficient detector assemblies.

Fission cross section uncertainties with the NIFFTE TPC

NASA Astrophysics Data System (ADS)

Sangiorgio, Samuele; Niffte Collaboration

2014-09-01

Nuclear data such as neutron-induced fission cross sections play a fundamental role in nuclear energy and defense applications. In recent years, understanding of these systems has become increasingly dependent upon advanced simulation and modeling, where uncertainties in nuclear data propagate in the expected performances of existing and future systems. It is important therefore that uncertainties in nuclear data are minimized and fully understood. For this reason, the Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) uses a Time Projection Chamber (TPC) to measure energy-differential (n,f) cross sections with unprecedented precision. The presentation will discuss how the capabilities of the NIFFTE TPC allow to directly measures systematic uncertainties in fission cross sections, in particular for what concerns fission-fragment identification, and target and beam uniformity. Preliminary results from recent analysis of 238U/235U and 239Pu/235U data collected with the TPC will be presented. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

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

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

Fission products detection in irradiated TRIGA fuel by means of gamma spectroscopy and MCNP calculation.

PubMed

Cagnazzo, M; Borio di Tigliole, A; Böck, H; Villa, M

2018-05-01

Aim of this work was the detection of fission products activity distribution along the axial dimension of irradiated fuel elements (FEs) at the TRIGA Mark II research reactor of the Technische Universität (TU) Wien. The activity distribution was measured by means of a customized fuel gamma scanning device, which includes a vertical lifting system to move the fuel rod along its vertical axis. For each investigated FE, a gamma spectrum measurement was performed along the vertical axis, with steps of 1 cm, in order to determine the axial distribution of the fission products. After the fuel elements underwent a relatively short cooling down period, different fission products were detected. The activity concentration was determined by calibrating the gamma detector with a standard calibration source of known activity and by MCNP6 simulations for the evaluation of self-absorption and geometric effects. Given the specific TRIGA fuel composition, a correction procedure is developed and used in this work for the measurement of the fission product Zr 95 . This measurement campaign is part of a more extended project aiming at the modelling of the TU Wien TRIGA reactor by means of different calculation codes (MCNP6, Serpent): the experimental results presented in this paper will be subsequently used for the benchmark of the models developed with the calculation codes. Copyright © 2018 Elsevier Ltd. All rights reserved.

Needs of Accurate Prompt and Delayed γ-spectrum and Multiplicity for Nuclear Reactor Designs

NASA Astrophysics Data System (ADS)

Rimpault, G.; Bernard, D.; Blanchet, D.; Vaglio-Gaudard, C.; Ravaux, S.; Santamarina, A.

The local energy photon deposit must be accounted accurately for Gen-IV fast reactors, advanced light-water nuclear reactors (Gen-III+) and the new experimental Jules Horowitz Reactor (JHR). The γ energy accounts for about 10% of the total energy released in the core of a thermal or fast reactor. The γ-energy release is much greater in the core of the reactor than in its structural sub-assemblies (such as reflector, control rod followers, dummy sub-assemblies). However, because of the propagation of γ from the core regions to the neighboring fuel-free assemblies, the contribution of γ energy to the total heating can be dominant. For reasons related to their performance, power reactors require a 7.5% (1σ) uncertainty for the energy deposition in non-fuelled zones. For the JHR material-testing reactor, a 5% (1 s) uncertainty is required in experimental positions. In order to verify the adequacy of the calculation of γ-heating, TLD and γ-fission chambers were used to derive the experimental heating values. Experimental programs were and are still conducted in different Cadarache facilities such as MASURCA (for SFR), MINERVE and EOLE (for JHR and Gen-III+ reactors). The comparison of calculated and measured γ-heating values shows an underestimation in all experimental programs indicating that for the most γ-production data from 239Pu in current nuclear-data libraries is highly suspicious.The first evaluation priority is for prompt γ-multiplicity for U and Pu fission but similar values for otheractinides such as Pu and U are also required. The nuclear data library JEFF3.1.1 contains most of the photon production data. However, there are some nuclei for which there are missing or erroneous data which need to be completed or modified. A review of the data available shows a lack of measurements for conducting serious evaluation efforts. New measurements are needed to guide new evaluation efforts which benefit from consolidated modeling techniques.

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

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

Gan, J.; Miller, B. D.; Keiser, D. D.

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (< 20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. This paper discusses the TEM results of the U-10Mo/Zr/Al6061 monolithic fuel plate (Plate ID: L1P09T, ~ 59% U-235 enrichment) irradiated in Advancedmore » Test Reactor at Idaho National Laboratory as part of RERTR-9B irradiation campaign with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 C, respectively. A total of 5 TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (> 1 µm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ~ 30 at% and ~ 7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.« less

Gaseous-fuel nuclear reactor research for multimegawatt power in space

NASA Technical Reports Server (NTRS)

Thom, K.; Schneider, R. T.; Helmick, H. H.

1977-01-01

In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

Nuclear reactor fuel element with vanadium getter on cladding

DOEpatents

Johnson, Carl E.; Carroll, Kenneth G.

1977-01-01

A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

Investigation of Liquid Metal Heat Exchanger Designs for Fission Surface Power

NASA Technical Reports Server (NTRS)

Dyson, Rodger W.; Penswick, Barry; Robbie, Malcolm; Geng, Steven M.

2009-01-01

Fission surface power is an option for future Moon and Mars surface missions. High power nuclear reactor heated Stirling convertors are an option to provide reliable power for long duration outpost operations. This report investigates various design approaches for the liquid metal to acceptor heat exchange and clarifies the details used in the analysis.

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

NASA Technical Reports Server (NTRS)

Garber, Anne E.

2007-01-01

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

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-11-21

A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY

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

S. T. Khericha

2007-04-01

The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed tomore » achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.« less

New prompt fission gamma-ray spectral data from 239Pu(nth, f) in response to a high priority request from OECD Nuclear Energy Agency

NASA Astrophysics Data System (ADS)

Gatera, Angélique; Belgya, Tamás; Geerts, Wouter; Göök, Alf; Hambsch, Franz-Josef; Lebois, Matthieu; Maróti, Boglárka; Oberstedt, Stephan; Oberstedt, Andreas; Postelt, Frederik; Qi, Liqiang; Szentmiklósi, Laszló; Vidali, Marzio; Zeiser, Fabio

2017-09-01

Benchmark reactor calculations have revealed an underestimation of γ-heat following fission of up to 28%. To improve the modelling of new nuclear reactors, the OECD/NEA initiated a nuclear data High Priority Request List (HPRL) entry for the major isotopes (235U, 239Pu). In response to that HPRL entry, we executed a dedicated measurement program on prompt fission γ-rays employing state-of-the-art lanthanum bromide (LaBr3) detectors with superior timing and good energy resolution. Our new results from 252Cf(sf), 235U(nth,f) and 241Pu(nth,f) provide prompt fission γ-ray spectra characteristics : average number of photons per fission, average total energy per fission and mean photon energy; all within 2% of uncertainty. We present preliminary results on 239Pu(nth,f), recently measured at the Budapest Neutron Centre and supported by the CHANDA Trans-national Access Activity, as well as discussing our different published results in comparison to the historical data and what it says about the discrepancy observed in the benchmark calculations.

Update on ORNL TRANSFORM Tool: Simulating Multi-Module Advanced Reactor with End-to-End I&C

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

Hale, Richard Edward; Fugate, David L.; Cetiner, Sacit M.

2015-05-01

The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the fourth year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled reactor) concepts, including the use of multiple coupled reactors at a single site. The focus of this report is the development of a steam generator and drum system model that includes the complex dynamics of typical steam drum systems, the development of instrumentation and controls for the steam generator with drum system model, and the development of multi-reactor module models that reflect the full power reactormore » innovative small module design concept. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor models; ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface technical area; and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the TRANSFORM tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the Advanced Reactors Technology program; (2) developing a library of baseline component modules that can be assembled into full plant models using available geometry, design, and thermal-hydraulic data; (3) defining modeling conventions for interconnecting component models; and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.« less

Spectroscopy of neutron rich nuclei using cold neutron induced fission of actinide targets at the ILL: The EXILL campaign

NASA Astrophysics Data System (ADS)

Blanc, A.; de France, G.; Drouet, F.; Jentschel, M.; Köster, U.; Mancuso, C.; Mutti, P.; Régis, J. M.; Simpson, G.; Soldner, T.; Ur, C. A.; Urban, W.; Vancraeyenest, A.

2013-12-01

One way to explore exotic nuclei is to study their structure by performing γ-ray spectroscopy. At the ILL, we exploit a high neutron flux reactor to induce the cold fission of actinide targets. In this process, fission products that cannot be accessed using standard spontaneous fission sources are produced with a yield allowing their detailed study using high resolution γ-ray spectroscopy. This is what was pursued at the ILL with the EXILL (for EXOGAM at the ILL) campaign. In the present work, the EXILL setup and performance will be presented.

Improved Fission Neutron Data Base for Active Interrogation of Actinides

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

Pozzi, Sara; Czirr, J. Bart; Haight, Robert

2013-11-06

This project will develop an innovative neutron detection system for active interrogation measurements. Many active interrogation methods to detect fissionable material are based on the detection of neutrons from fission induced by fast neutrons or high-energy gamma rays. The energy spectrum of the fission neutrons provides data to identify the fissionable isotopes and materials such as shielding between the fissionable material and the detector. The proposed path for the project is as follows. First, the team will develop new neutron detection systems and algorithms by Monte Carlo simulations and bench-top experiments. Next, They will characterize and calibrate detection systems bothmore » with monoenergetic and white neutron sources. Finally, high-fidelity measurements of neutron emission from fissions induced by fast neutrons will be performed. Several existing fission chambers containing U-235, Pu-239, U-238, or Th-232 will be used to measure the neutron-induced fission neutron emission spectra. The challenge for making confident measurements is the detection of neutrons in the energy ranges of 0.01 – 1 MeV and above 8 MeV, regions where the basic data on the neutron energy spectrum emitted from fission is least well known. In addition, improvements in the specificity of neutron detectors are required throughout the complete energy range: they must be able to clearly distinguish neutrons from other radiations, in particular gamma rays and cosmic rays. The team believes that all of these challenges can be addressed successfully with emerging technologies under development by this collaboration. In particular, the collaboration will address the area of fission neutron emission spectra for isotopes of interest in the advanced fuel cycle initiative (AFCI).« less

Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

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

A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

2010-06-01

2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has beenmore » restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since

Development of a Polysilicon Process Based on Chemical Vapor Deposition of Dichlorosilane in an Advanced Siemen's Reactor

NASA Technical Reports Server (NTRS)

Arevidson, A. N.; Sawyer, D. H.; Muller, D. M.

1983-01-01

Dichlorosilane (DCS) was used as the feedstock for an advanced decomposition reactor for silicon production. The advanced reactor had a cool bell jar wall temperature, 300 C, when compared to Siemen's reactors previously used for DCS decomposition. Previous reactors had bell jar wall temperatures of approximately 750 C. The cooler wall temperature allows higher DCS flow rates and concentrations. A silicon deposition rate of 2.28 gm/hr-cm was achieved with power consumption of 59 kWh/kg. Interpretation of data suggests that a 2.8 gm/hr-cm deposition rate is possible. Screening of lower cost materials of construction was done as a separate program segment. Stainless Steel (304 and 316), Hastalloy B, Monel 400 and 1010-Carbon Steel were placed individually in an experimental scale reactor. Silicon was deposited from trichlorosilane feedstock. The resultant silicon was analyzed for electrically active and metallic impurities as well as carbon. No material contributed significant amounts of electrically active or metallic impurities, but all contributed carbon.

Measurement of the ROT effect in the neutron induced fission of 235U in the 0.3 eV resonance at a hot source of polarized neutrons

NASA Astrophysics Data System (ADS)

Kopatch, Yuri; Novitsky, Vadim; Ahmadov, Gadir; Gagarsky, Alexei; Berikov, Daniyar; Danilyan, Gevorg; Hutanu, Vladimir; Klenke, Jens; Masalovich, Sergey

2018-03-01

The TRI and ROT asymmetries in fission of heavy nuclei have been extensively studied during more than a decade. The effects were first discovered in the ternary fission in a series of experiments performed at the ILL reactor (Grenoble) by a collaboration of Russian and European institutes, and were carefully measured for a number of fissioning nuclei. Later on, the ROT effect has been observed in the emission of prompt gamma rays and neutrons in fission of 235U and 233U, although its value was an order of magnitude smaller than in the α-particle emission from ternary fission. All experiments performed so far are done with cold polarized neutrons, what assumes a mixture of several spin states, the weights of these states being not well known. The present paper describes the first attempt to get "clean" data by performing the measurement of gamma and neutron asymmetries in an isolated resonance of 235U at the POLI instrument of the FRM2 reactor in Garching.

Summary of the Advanced Reactor Design Criteria (ARDC) Phase 2 Activities

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

Holbrook, Mark Raymond

This report provides an end-of-year summary reflecting the progress and status of proposed regulatory design criteria for advanced non-LWR designs in accordance with the Level 3 milestone in M3AT-15IN2001017 in work package AT-15IN200101. These criteria have been designated as ARDC, and they provide guidance to future applicants for addressing the GDC that are currently applied specifically to LWR designs. The report provides a summary of Phase 2 activities related to the various tasks associated with ARDC development and the subsequent development of example adaptations of ARDC for Sodium Fast Reactor (SFR) and modular High Temperature Gas-cooled Reactor (HTGR) designs.

Advanced Reactors-Intermediate Heat Exchanger (IHX) Coupling: Theoretical Modeling and Experimental Validation

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

Utgikar, Vivek; Sun, Xiaodong; Christensen, Richard

2016-12-29

The overall goal of the research project was to model the behavior of the advanced reactorintermediate heat exchange system and to develop advanced control techniques for off-normal conditions. The specific objectives defined for the project were: 1. To develop the steady-state thermal hydraulic design of the intermediate heat exchanger (IHX); 2. To develop mathematical models to describe the advanced nuclear reactor-IHX-chemical process/power generation coupling during normal and off-normal operations, and to simulate models using multiphysics software; 3. To develop control strategies using genetic algorithm or neural network techniques and couple these techniques with the multiphysics software; 4. To validate themore » models experimentally The project objectives were accomplished by defining and executing four different tasks corresponding to these specific objectives. The first task involved selection of IHX candidates and developing steady state designs for those. The second task involved modeling of the transient and offnormal operation of the reactor-IHX system. The subsequent task dealt with the development of control strategies and involved algorithm development and simulation. The last task involved experimental validation of the thermal hydraulic performances of the two prototype heat exchangers designed and fabricated for the project at steady state and transient conditions to simulate the coupling of the reactor- IHX-process plant system. The experimental work utilized the two test facilities at The Ohio State University (OSU) including one existing High-Temperature Helium Test Facility (HTHF) and the newly developed high-temperature molten salt facility.« less

Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

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

S. Bragg-Sitton; J. Bess; J. Werner

2011-09-01

Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al.,more » 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).« less

Reactor Neutronics: Impact of Fissile Material

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

Heidet, F.; Hill, R. N.

Here, given a wide variety of reactor designs and fuel types, it can be difficult to identify the underlying cause of basic performance differences such as flux level and enrichment requirement. In this paper, using solely the definitions of the core multiplication factor and core power, simple relations have been derived allowing estimates of the flux ratio and fissile material concentration ratio for any reactor concept when 235U is replaced with 239Pu or vice-versa. These relations are functions of the neutron non-leakage probability, and one only needs to know number of neutrons emitted per fission, and the fission cross-section ratiomore » between the 235U system and the 239Pu system. It is found that for a reactor concept having significant leakage, the achievable flux level when using 239Pu as fissile material can be up to 45% larger than when using 235U as fissile material, and the required fissile concentration of 239Pu is up to 48% lower than that of 235U to achieve criticality.« less

Reactor Neutronics: Impact of Fissile Material

DOE PAGES

Heidet, F.; Hill, R. N.

2017-06-09

Here, given a wide variety of reactor designs and fuel types, it can be difficult to identify the underlying cause of basic performance differences such as flux level and enrichment requirement. In this paper, using solely the definitions of the core multiplication factor and core power, simple relations have been derived allowing estimates of the flux ratio and fissile material concentration ratio for any reactor concept when 235U is replaced with 239Pu or vice-versa. These relations are functions of the neutron non-leakage probability, and one only needs to know number of neutrons emitted per fission, and the fission cross-section ratiomore » between the 235U system and the 239Pu system. It is found that for a reactor concept having significant leakage, the achievable flux level when using 239Pu as fissile material can be up to 45% larger than when using 235U as fissile material, and the required fissile concentration of 239Pu is up to 48% lower than that of 235U to achieve criticality.« less

Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

NASA Astrophysics Data System (ADS)

Mella, R.; Wenman, M. R.

2013-06-01

Thermo-mechanical contributions to pellet-clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS's well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used. The usability of a FE based fuel performance code would be an enhancement over past codes. Pre- and post-processors have lowered the entry barrier for the development of a fuel performance model to permit the ability to model complicated systems. Typical runtimes for a 5 year axisymmetric model takes less than one hour on a single core workstation. The current model has implemented: Non-linear fuel thermal behaviour, including a complex description of heat flow in the fuel. Coupled with a variety of

Design of a Mechanical NaK Pump for Fission Space Power Systems

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Bradley, David; Godfroy, Thomas

2010-01-01

Alkali liquid metal cooled fission reactor concepts are under development for mid-range spaceflight power requirements. One such concept utilizes a sodium-potassium eutectic (NaK) as the primary loop working fluid. Traditionally, linear induction pumps have been used to provide the required flow and head conditions for liquid metal systems but can be limited in performance. This paper details the design, build, and check-out test of a mechanical NaK pump. The pump was designed to meet reactor cooling requirements using commercially available components modified for high temperature NaK service.

FALSTAFF: A New Tool for Fission Fragment Characterization

NASA Astrophysics Data System (ADS)

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

2014-05-01

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

Advance High Temperature Inspection Capabilities for Small Modular Reactors: Part 1 - Ultrasonics

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

Bond, Leonard J.; Bowler, John R.

The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-servicemore » inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO 3-xPbTiO 3) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.« less

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

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

Tovesson, Fredrik; Mayorov, Dmitriy; Duke, Dana

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

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

DOE PAGES

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

2017-09-13

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

Advanced Small Modular Reactor Economics Model Development

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

Harrison, Thomas J.

2014-10-01

The US Department of Energy Office of Nuclear Energy’s Advanced Small Modular Reactor (SMR) research and development activities focus on four key areas: Developing assessment methods for evaluating advanced SMR technologies and characteristics; and Developing and testing of materials, fuels and fabrication techniques; and Resolving key regulatory issues identified by US Nuclear Regulatory Commission and industry; and Developing advanced instrumentation and controls and human-machine interfaces. This report focuses on development of assessment methods to evaluate advanced SMR technologies and characteristics. Specifically, this report describes the expansion and application of the economic modeling effort at Oak Ridge National Laboratory. Analysis ofmore » the current modeling methods shows that one of the primary concerns for the modeling effort is the handling of uncertainty in cost estimates. Monte Carlo–based methods are commonly used to handle uncertainty, especially when implemented by a stand-alone script within a program such as Python or MATLAB. However, a script-based model requires each potential user to have access to a compiler and an executable capable of handling the script. Making the model accessible to multiple independent analysts is best accomplished by implementing the model in a common computing tool such as Microsoft Excel. Excel is readily available and accessible to most system analysts, but it is not designed for straightforward implementation of a Monte Carlo–based method. Using a Monte Carlo algorithm requires in-spreadsheet scripting and statistical analyses or the use of add-ons such as Crystal Ball. An alternative method uses propagation of error calculations in the existing Excel-based system to estimate system cost uncertainty. This method has the advantage of using Microsoft Excel as is, but it requires the use of simplifying assumptions. These assumptions do not necessarily bring into question the analytical results. In

Phase 1 Space Fission Propulsion System Design Considerations

NASA Technical Reports Server (NTRS)

Houts, Mike; VanDyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Carter, Robert; Dickens, Ricky; Salvail, Pat; Hrbud, Ivana; Rodgers, Stephen L. (Technical Monitor)

2001-01-01

Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and operated. Studies conducted in fiscal year 2001 (IISTP, 2001) show that fission electric propulsion (FEP) systems operating at 80 kWe or above could enhance or enable numerous robotic outer solar system missions of interest. At these power levels it is possible to develop safe, affordable systems that meet mission performance requirements. In selecting the system design to pursue, seven evaluation criteria were identified: safety, reliability, testability, specific mass, cost, schedule, and programmatic risk. A top-level comparison of three potential concepts was performed: an SP-100 based pumped liquid lithium system, a direct gas cooled system, and a heatpipe cooled system. For power levels up to at least 500 kWt (enabling electric power levels of 125-175 kWe, given 25-35% power conversion efficiency) the heatpipe system has advantages related to several criteria and is competitive with respect to all. Hardware-based research and development has further increased confidence in the heatpipe approach. Successful development and utilization of a "Phase 1" fission electric propulsion system will enable advanced Phase 2 and Phase 3 systems capable of providing rapid, affordable access to any point in the solar system.

Phase 1 space fission propulsion system design considerations

NASA Astrophysics Data System (ADS)

Houts, Mike; van Dyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Salvail, Pat; Hrbud, Ivana; Carter, Robert

2002-01-01

Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and operated. Studies conducted in fiscal year 2001 (IISTP, 2001) show that fission electric propulsion (FEP) systems operating at 80 kWe or above could enhance or enable numerous robotic outer solar system missions of interest. At these power levels it is possible to develop safe, affordable systems that meet mission performance requirements. In selecting the system design to pursue, seven evaluation criteria were identified: safety, reliability, testability, specific mass, cost, schedule, and programmatic risk. A top-level comparison of three potential concepts was performed: an SP-100 based pumped liquid lithium system, a direct gas cooled system, and a heatpipe cooled system. For power levels up to at least 500 kWt (enabling electric power levels of 125-175 kWe, given 25-35% power conversion efficiency) the heatpipe system has advantages related to several criteria and is competitive with respect to all. Hardware-based research and development has further increased confidence in the heatpipe approach. Successful development and utilization of a ``Phase 1'' fission electric propulsion system will enable advanced Phase 2 and Phase 3 systems capable of providing rapid, affordable access to any point in the solar system. .

JEN-1 Reactor Control System; SISTEMA DE CONTROL DEL REACTOR JEN-1

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

Cantillo, M.F.; Nuno, C.M.; Andreu, J.L.M.

1963-01-01

ABS>The JEN-1 3Mw power swimming pool reactor electrical control circuits are described. Start-up, power generation in the core, and shutdown are controlled by the reactor control system. This control system guarantees in each moment the safety conditions during reactor operation. Each circuit was represented by a scheme, complemented with a description of its function, components, and operation theory. Components described include: scram circuit; fission counter control circuit; servo control circuit; control circuit of safety sheets; control circuits of primary, secondary, and clean-up pump motors and tower fan motor; primary valve motor circuit; center cubicle alarm circuit; and process alarm circuit.more » (auth)« less

Using Genetic Buffering Relationships Identified in Fission Yeast to Elucidate the Molecular Pathology of Tuberous Sclerosis

DTIC Science & Technology

2015-07-01

AWARD NUMBER: W81XWH-14-1-0169 TITLE: Using Genetic Buffering Relationships Identified in Fission Yeast to Elucidate the Molecular Pathology of...DATES COVERED 1 July 2014 - 30 June 2015 4. TITLE AND SUBTITLE 5a. CONTRACT NUMBER Using Genetic Buffering Relationships Identified in Fission Yeast ...SUPPLEMENTARY NOTES 14. ABSTRACT Using the genetically tractable fission yeast as a model, we sought to exploit recent advances in gene interaction

REACTOR CONTROL DEVICE

DOEpatents

Graham, R.H.

1962-09-01

A wholly mechanical compact control device is designed for automatically rendering the core of a fission reactor subcritical in response to core temperatures in excess of the design operating temperature limit. The control device comprises an expansible bellows interposed between the base of a channel in a reactor core and the inner end of a fuel cylinder therein which is normally resiliently urged inwardly. The bellows contains a working fluid which undergoes a liquid to vapor phase change at a temperature substantially equal to the design temperature limit. Hence, the bellows abruptiy expands at this limiting temperature to force the fuel cylinder outward and render the core subcritical. The control device is particularly applicable to aircraft propulsion reactor service. (AEC)

NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.

1962-08-14

A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

Determining Reactor Fuel Type from Continuous Antineutrino Monitoring

NASA Astrophysics Data System (ADS)

Jaffke, Patrick; Huber, Patrick

2017-09-01

We investigate the ability of an antineutrino detector to determine the fuel type of a reactor. A hypothetical 5-ton antineutrino detector is placed 25 m from the core and measures the spectral shape and rate of antineutrinos emitted by fission fragments in the core for a number of 90-d periods. Our results indicate that four major fuel types can be differentiated from the variation of fission fractions over the irradiation time with a true positive probability of detection at approximately 95%. In addition, we demonstrate that antineutrinos can identify the burnup at which weapons-grade mixed-oxide (MOX) fuel would be reduced to reactor-grade MOX, on average, providing assurance that plutonium-disposition goals are met. We also investigate removal scenarios where plutonium is purposefully diverted from a mixture of MOX and low-enriched uranium fuel. Finally, we discuss how our analysis is impacted by a spectral distortion around 6 MeV observed in the antineutrino spectrum measured from commercial power reactors.

NUCLEAR REACTOR

DOEpatents

Grebe, J.J.

1961-01-24

A core structure for neutronic reactors adapted for the propulsion of aircraft and rockets is offered. The core is designed for cooling by gaseous media, and comprises a plurality of hollow tapered tubular segments of a porous moderating material impregniated with fissionable fuel nested about a common axis. Alternate ends of the segments are joined. In operation a coolant gas passes through the porous structure and is heated.

Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

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

Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna

This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t 1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

DOE PAGES

Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; ...

2015-10-26

This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t 1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

Supervisory Control System Architecture for Advanced Small Modular Reactors

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

Cetiner, Sacit M; Cole, Daniel L; Fugate, David L

2013-08-01

This technical report was generated as a product of the Supervisory Control for Multi-Modular SMR Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (SMR) Research and Development Program of the U.S. Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular advanced SMR (AdvSMR) plants. This research activity advances the state-of-the art by incorporating decision making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides a brief history ofmore » hierarchical functional architectures and the current state-of-the-art, describes a reference AdvSMR to show the dependencies between systems, presents a hierarchical structure for supervisory control, indicates the importance of understanding trip setpoints, applies a new theoretic approach for comparing architectures, identifies cyber security controls that should be addressed early in system design, and describes ongoing work to develop system requirements and hardware/software configurations.« less

Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

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

O'Brien, James E.; Sabharwall, Piyush; Yoon, Su -Jong

2014-09-01

This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs)more » at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and

Modelling of the anti-neutrino production and spectra from a Magnox reactor

NASA Astrophysics Data System (ADS)

Mills, Robert W.; Mountford, David J.; Coleman, Jonathon P.; Metelko, Carl; Murdoch, Matthew; Schnellbach, Yan-Jie

2018-01-01

The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152) during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of monitored reactor

NEUTRONIC REACTOR FUEL PUMP

DOEpatents

Cobb, W.G.

1959-06-01

A reactor fuel pump is described which offers long life, low susceptibility to radiation damage, and gaseous fission product removal. An inert-gas lubricated bearing supports a journal on one end of the drive shsft. The other end has an impeller and expansion chamber which effect pumping and gas- liquid separation. (T.R.H.)

Apollo - An advanced fuel fusion power reactor for the 21st century

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

Kulcinski, G.L.; Emmert, G.A.; Blanchard, J.P.

1989-03-01

A preconceptual design of a tokamak reactor fueled by a D-He-3 plasma is presented. A low aspect ratio (A=2-4) device is studied here but high aspect ratio devices (A > 6) may also be quite attractive. The Apollo D-He-3 tokamak capitalizes on recent advances in high field magnets (20 T) and utilizes rectennas to convert the synchrotron radiation directly to electricity. The overall efficiency ranges from 37 to 52% depending on whether the bremsstrahlung energy is utilized. The low neutron wall loading (0.1 MW/m/sup 2/) allows a permanent first wall to be designed and the low nuclear decay heat enablesmore » the reactor to be classed as inherently safe. The cost of electricity from Apollo is > 40% lower than electricity from a similar sized DT reactor.« less

Estimated inventory of radionuclides in former Soviet Union naval reactors dumped in the Kara Sea

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

Mount, M.E.; Sheaffer, M.K.; Abbott, D.T.

1993-07-01

Radionuclide inventories have been estimated for the reactor cores, reactor components, and primary system corrosion products in the former Soviet Union naval reactors dumped at the Abrosimov Inlet, Tsivolka Inlet, Stepovoy Inlet, Techeniye Inlet, and Novaya Zemlya Depression sites in the Kara Sea between 1965 and 1988. For the time of disposal, the inventories are estimated at 69 to 111 kCi of actinides plus daughters and 3,053 to 7,472 kCi of fission products in the reactor cores, 917 to 1,127 kCi of activation products in the reactor components, and 1.4 to 1.6 kCi of activation products in the primary systemmore » corrosion products. At the present time, the inventories are estimated to have decreased to 23 to 38 kCi of actinides plus daughters and 674 to 708 kCi of fission products in the reactor cores, 124 to 126 kCi of activation products in the reactor components, and 0.16 to 0.17 kCi of activation products in the primary system corrosion products. Twenty years from now, the inventories are projected to be 11 to 18 kCi of actinides plus daughters and 415 to 437 kCi of fission products in the reactor cores, 63.5 to 64 kCi of activation products in the reactor components, and 0.014 to 0.015 kCi of activation products in the primary system corrosion products. All actinide activities are estimated to be within a factor of two.« less

A molten Salt Am242M Production Reactor for Space Applications

NASA Technical Reports Server (NTRS)

Emrich, William

2005-01-01

The use of Am242m holds great promise for increasing the efficiency nuclear thermal rocket engines. Because Am242m has the highest fission cross section of any known isotope (1000's of barns), its extremely high reactivity may be used to directly heat a propellant gas with fission fragments. Since this isotope does not occur naturally, it must be bred in special production reactors designed for that purpose. The primary advantage to using molten salt reactors for breeding Am242m is that the reactors can be reprocessed continually yielding a constant rate of production of the isotope. Once built and initially fueled, the reactor will continually breed the additional fuel it needs to remain critical. The only feedstock required is a salt of U238. No enriched fuel is required during normal operation and all fissile material, except the Am242m, is maintained in a closed loop. For a reactor operating at 200 MW several kilograms of Am242m may be bred each year.

Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

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

R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag

2012-04-01

The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather thanmore » graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in

Fission and activation of uranium by fusion-plasma neutrons

NASA Technical Reports Server (NTRS)

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

1978-01-01

Fusion-fission hybrid reactors are discussed in terms of two main purposes: to breed fissile materials (Pu 233 and Th 233 from U 238 or Th 232) for use in low-reactivity breeders, and to produce tritium from lithium to refuel fusion plasma cores. Neutron flux generation is critical for both processes. Various methods for generating the flux are described, with attention to new geometries for multiple plasma focus arrays, e.g., hypocycloidal pinch and staged plasma focus devices. These methods are evaluated with reference to their applicability to D-D fusion reactors, which will ensure a virtually unlimited energy supply. Accurate observations of the neutron flux from such schemes are obtained by using different target materials in the plasma focus.

Development work for a borax internal core-catcher for a gas-cooled fast reactor

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

Donne, M.D.; Dorner, S.; Schumacher, G.

1978-07-01

Preliminary thermal calculations show that a corecatcher, which is able to cope with the complete meltdown of the core and blankets of a 1000-MW(electric) gas-cooled fast reactor, appears to be feasible. This core-catcher is based on borax (Na/sub 2/B/sub 4/O/sub 7/) dissolving the oxide fuel and the fission products occurring in oxide form. The borax is contained in steel boxes forming a 2.2-m-thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel (PCRV), just underneath the reactor core. After a complete meltdown accident, the fission products, in oxide form, are dispersed in the pool formedmore » by the liquid borax. The metallic fission products are contained in the steel lying below the borax pool and in contact with the water-cooled PCRV liner. The volumetric power density of the molten core is conveniently reduced as it is dissolved in the borax, and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system.« less

FLUID MODERATED REACTOR

DOEpatents

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1957-10-22

A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

1976-01-01

The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

The Politics of Forgetting: Otto Hahn and the German Nuclear-Fission Project in World War II

NASA Astrophysics Data System (ADS)

Sime, Ruth Lewin

2012-03-01

As the co-discoverer of nuclear fission and director of the Kaiser Wilhelm Institute for Chemistry, Otto Hahn (1879-1968) took part in Germany`s nuclear-fission project throughout the Second World War. I outline Hahn's efforts to mobilize his institute for military-related research; his inclusion in high-level scientific structures of the military and the state; and his institute's research programs in neutron physics, isotope separation, transuranium elements, and fission products, all of potential military importance for a bomb or a reactor and almost all of it secret. These activities are contrasted with Hahn's deliberate misrepresentations after the war, when he claimed that his wartime work had been nothing but "purely scientific" fundamental research that was openly published and of no military relevance.

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

NASA Astrophysics Data System (ADS)

Bundgaard, Jeremy J.

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

METHOD AND APPARATUS FOR IMPROVING PERFORMANCE OF A FAST REACTOR

DOEpatents

Koch, L.J.

1959-01-20