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Sample records for low-enriched uranium fuel

  1. Feasibility of Low Enriched Uranium Fuel for Space Nuclear Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, Paolo; Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-05-15

    The purpose of this initial study is to create a baseline with which to perform further analysis and to build a solid understanding of the neutronic characteristics of a solid core for the nuclear thermal rocket. Once consistency with work done at Idaho National Laboratory (INL) is established, this paper will provide a study of other fuel types, such as low and medium-enriched uranium fuels. This paper will examine how the implementation of each fuel type affects the multiplication factor of the reactor, and will then explore different possibilities for alterations needed to accommodate their successful usage. The reactor core analysis was done using the MCNP5 code. While this study has not shown that the SNRE can be easily retrofitted for low-enriched U fuel, it has made a detailed study of the SNRE, and identified the difficulties of the implementation of low-enriched fuels in small nuclear rockets. These difficulties are the need for additional moderation and fuel mass in order to achieve a critical mass. Neither of these is insurmountable. Future work includes finding the best method by which to increase the internal moderation of the reactor balanced with appropriate sizing to prevent neutron leakage. Both of these are currently being studied. This paper will present a study of the Small Nuclear Rocket Engine (SNRE) and the feasibility of using low enriched Uranium (LEU) instead of the traditional high enriched Uranium (HEU) fuels.

  2. Conversion of research reactors to low-enrichment uranium fuels

    International Nuclear Information System (INIS)

    Muranaka, R.G.

    1983-01-01

    There are at present approximately 350 research reactors in 52 countries ranging in power from less than 1 watt to 100 Megawatt and over. In the 1970's, many people became concerned about the possibility that some fuels and fuel cycles could provide an easy route to the acquisition of nuclear weapons. Since enrichment to less than 20% is internationally recognized as a fully adequate barrier to weapons usability, certain Member States have moved to minimize the international trade in highly enriched uranium and have established programmes to develop the technical means to help convert research reactors to the use of low-enrichment fuels with minimum penalties. This could involve modifications in the design of the reactor and development of new fuels. As a result of these programmes, it is expected that most research reactors can be converted to the use of low-enriched fuel

  3. The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

    International Nuclear Information System (INIS)

    Townes, B.M.; Hilborn, J.W.

    1985-06-01

    A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

  4. The low enriched uranium fuel cycle in Ontario

    International Nuclear Information System (INIS)

    Archinoff, G.H.

    1979-02-01

    Six fuel-cycle strategies for use in CANDU reactors are examined in terms of their uranium-conserving properties and their ease of commercialization for three assumed growth rates of installed nuclear capacity in Ontario. The fuel cycle strategies considered assume the continued use of the natural uranium cycle up to the mid-1990's. At that time, the low-enriched uranium (LEU) cycle is gradually introduced into the existing power generation grid. In the mid-2020's one of four advanced cycles is introduced. The advanced cycles considered are: mixed oxide, intermediate burn-up thorium (Pu topping), intermediate burn-up thorium (U topping), and LMFBR. For comparison purposes an all natural uranium strategy and a natural uranium-LEU strategy (with no advanced cycle) are also included. None of the strategies emerges as a clear, overall best choice. (LL)

  5. Low-enriched uranium-molybdenum fuel plate development

    International Nuclear Information System (INIS)

    Wiencek, T.C.; Prokofiev, I.G.

    2000-01-01

    To examine the fabricability of low-enriched uranium-molybdenum powders, full-size 450 x 60 x 0.5-mm (17.7 x 2.4 x 0.020-in.) fuel zone test plates loaded to 6 g U/cm 3 were produced. U-10 wt.% Mo powders produced by two methods, centrifugal atomization and grinding, were tested. These powders were supplied at no cost to Argonne National Laboratory by the Korean Atomic Energy Research Institute and Atomic Energy of Canada Limited, respectively. Fuel homogeneity indicated that both of the powders produced acceptable fuel plates. Operator skill during loading of the powder into the compacting die and fuel powder morphology were found to be important when striving to achieve homogeneous fuel distribution. Smaller, 94 x 22 x 0.6-mm (3.7 x 0.87 x 0.025-in.) fuel zone, test plates were fabricated using U-10 wt.% Mo foil disks instead of a conventional powder metallurgy compact. Two fuel plates of this type are currently undergoing irradiation in the RERTR-4 high-density fuel experiment in the Advanced Test Reactor. (author)

  6. Qualification status of LEU [low enriched uranium] fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.

    1987-01-01

    Sufficient data has been obtained from tests of high-density, low-enriched fuels for research and test reactors to declare them qualified for use. These fuels include UZrH x (TRIGA fuel) and UO 2 (SPERT fuel) for rod-type reactors and UAl x , U 3 O 8 , U 3 Si 2 , and U 3 Si dispersed in aluminium for plate-type reactors. Except for U 3 Si, the allowable fission density for LEU applications is limited only by the available 235 U. Several reactors are now using these fuels, and additional conversions are in progress. The basic performance characteristics and limits, if any, of the qualified low-enriched (and medium-enriched) fuels are discussed. Continuing and planned work to qualify additional fuels is also discussed. (Author)

  7. Yalina booster subcritical assembly performance with low enriched uranium fuel

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gohar, Yousry

    2011-01-01

    The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

  8. Yalina booster subcritical assembly performance with low enriched uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto; Gohar, Yousry, E-mail: alby@anl.gov [Argonne National Laboratory, Lemont, IL (United States)

    2011-07-01

    The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

  9. Low enriched uranium fuel conversion and fuel shipping guide

    International Nuclear Information System (INIS)

    1997-01-01

    The analysis of reactor core physics and thermal hydraulics was completed in 1993. A supplement to the Final Safety Analysis Report describing the results of these analyses was submitted to the Nuclear Regulatory Commission along with proposed Technical Specifications in May, 1993. Discussions with the NRC staff led to a submittal of revised proposed Technical Specifications in February, 1994. The analytical work is complete. A second portion of the grant was to develop a fuel shipping guide for university research reactors. Such a guide was developed and is available for use by the research reactor community

  10. Advanced fuel cycles: a rationale and strategy for adopting the low-enriched-uranium fuel cycle

    International Nuclear Information System (INIS)

    James, R.A.

    1980-01-01

    A two-year study of alternatives to the natural uranium fuel cycle in CANDU reactors is summarized. The possible advanced cycles are briefly described. Selection criteria for choosing a cycle for development include resource utilization, economics, ease of implementaton, and social acceptability. It is recommended that a detailed study should be made with a view to the early implementation of the low-enriched uranium cycle. (LL)

  11. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

    Gietzen, A.

    1993-01-01

    Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with General Atomic's standard commercial warranty

  12. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

    Gietzen, A.

    1993-01-01

    Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium-zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched-uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with GA's standard commercial warranty

  13. Experiments of JRR-4 low-enriched-uranium-silicied fuel core

    International Nuclear Information System (INIS)

    Hirane, Nobuhiko; Ishikuro, Yasuhiro; Nagadomi, Hideki; Yokoo, Kenji; Horiguchi, Hironori; Nemoto, Takumi; Yamamoto, Kazuyoshi; Yagi, Masahiro; Arai, Nobuyoshi; Watanabe, Shukichi; Kashima, Yoichi

    2006-03-01

    JRR-4, a light-water-moderated and cooled, swimming pool type research reactor using high-enriched uranium plate-type fuels had been operated from 1965 to 1996. In order to convert to low-enriched-uranium-silicied fuels, modification work had been carried out for 2 years, from 1996 to 1998. After the modification, start-up experiments were carried out to obtain characteristics of the low-enriched-uranium-silicied fuel core. The measured excess reactivity, reactor shutdown margin and the maximum reactivity addition rate satisfied the nuclear limitation of the safety report for licensing. It was confirmed that conversion to low-enriched-uranium-silicied fuels was carried out properly. Besides, the necessary data for reactor operation were obtained, such as nuclear, thermal hydraulic and reactor control characteristics. This report describes the results of start-up experiments and burnup experiments. The first criticality of low-enriched-uranium-silicied core was achieved on 14th July 1998, and the operation for joint-use has been carried out since 6th October 1998. (author)

  14. Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2009-01-01

    The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 deg. C to 50 deg. C and 100 deg. C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U 3 O 8 -Al was about 2% more than the original UAl x -Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

  15. Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2009-01-01

    The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U 3 Si 2 -Al followed by 0.03% for U 3 Si-Al, and 0.01% for U 3 O 8 -Al fuel. The U 3 O 8 -Al fueled reactor gave the maximum ρ excess at BOL which was 21.67% more than the original fuel followed by U 3 Si-Al which was 2.55% more, while that of U 3 Si 2 -Al was 2.50% more than the original UAl x -Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U 3 O 8 -Al followed by U 3 Si-Al and then U 3 Si 2 -Al fuel.

  16. Development of IAEA safeguards at low enrichment uranium fuel fabrication plants

    International Nuclear Information System (INIS)

    Badawy, I.

    1988-01-01

    In this report the nuclear material at low enrichment uranium fuel fabrication plants under IAEA safeguards is studied. The current verification practices of the nuclear material and future improvements are also considered. The problems met during the implementation of the the verification measures of the nuclear material - particularly for the fuel assemblies are discussed. The additional verification activities as proposed for future improvements are also discussed including the physical inventory verification and the verification of receipts and shipments. It is concluded that the future development of the present IAEA verification practices at low enrichment uranium fuel fabrication plants would necessitate the application of quantitative measures of the nuclear material and the implementation of advanced measurement techniques and instruments. 2 fig., 4 tab

  17. Research reactor core conversion from the use of highly enriched uranium to the use of low enriched uranium fuels guidebook

    International Nuclear Information System (INIS)

    1980-08-01

    In view of the proliferation concerns caused by the use of highly enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors will be more restricted in the future, this document has been prepared to assist reactor operators in determining whether conversion to the use of low enriched uranium (LEU) fuel designs is technically feasible for their specific reactor, and to assist in making a smooth transition to the use of LEU fuel designs where appropriate

  18. Report of the Working Party on the conversion of HIFAR to low enrichment uranium fuel

    International Nuclear Information System (INIS)

    1986-06-01

    This report states the effect on research reactor operations and applications of international and national political decisions relating to fuel enrichment. Technical work done in Australia and overseas to establish parameters for conversion of research reactors from High Enrichment Uranium (HEU) to Low Enrichment Uranium (LEU) have been considered in developing a strategy for HIFAR. The requirements of the research groups, isotope production group and reactor operating staff have been considered. For HIFAR to continue to provide the required facilities in support of the national need, it is concluded these should be no reduction of neutron flux

  19. Conversion of research and test reactors to low enriched uranium fuel: technical overview and program status

    International Nuclear Information System (INIS)

    Roglans-Ribas, J.

    2008-01-01

    Many of the nuclear research and test reactors worldwide operate with high enriched uranium fuel. In response to worries over the potential use of HEU from research reactors in nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel by converting research reactors to low enriched uranium (LEU) fuel. The Reactor Conversion program is currently under the DOE's National Nuclear Security Administration's Global Threat Reduction Initiative (GTRI). 55 of the 129 reactors included in the scope have been already converted to LEU fuel or have shutdown prior to conversion. The major technical activities of the Conversion Program include: (1) the development of advanced LEU fuels; (2) conversion analysis and conversion support; and (3) technology development for the production of Molybdenum-99 (Mo 99 ) with LEU targets. The paper provides an overview of the status of the program, the technical challenges and accomplishments, and the role of international collaborations in the accomplishment of the Conversion Program objectives. Nuclear research and test reactors worldwide have been in operation for over 60 years. Many of these facilities operate with high enriched uranium fuel. In response to increased worries over the potential use of HEU from research reactors in the manufacturing of nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel in research reactors by converting them to low enriched uranium (LEU) fuel. The reactor conversion program was initially focused on U.S.-supplied reactors, but in the early 1990s it expanded and began to collaborate with Russian institutes with the objective of converting Russian supplied reactors to the use of LEU fuel.

  20. Moderator configuration options for a low-enriched uranium fueled Kilowatt-class Space Nuclear Reactor

    International Nuclear Information System (INIS)

    King, Jeffrey C.; Mencarini, Leonardo de Holanda; Guimaraes, Lamartine N. F.

    2015-01-01

    The Brazilian Air Force, through its Institute for Advanced Studies (Instituto de Estudos Avancados, IEAv/DCTA), and the Colorado School of Mines (CSM) are studying the feasibility of a space nuclear reactor with a power of 1-5 kW e and fueled with Low-Enriched Uranium (LEU). This type of nuclear reactor would be attractive to signatory countries of the Non-Proliferation Treaty (NPT) or commercial interests. A LEU-fueled space reactor would avoid the security concerns inherent with Highly Enriched Uranium (HEU) fuel. As an initial step, the HEU-fueled Kilowatt Reactor Using Stirling Technology (KRUSTY) designed by the Los Alamos National Laboratory serves as a basis for a similar reactor fueled with LEU fuel. Using the computational code MCNP6 to predict the reactor neutronics performance, the size of the resulting reactor fueled with 19.75 wt% enriched uranium-10 wt% molybdenum alloy fuel is adjusted to match the excess reactivity of KRUSTY. Then, zirconium hydride moderator is added to the core to reduce the size of the reactor. This work presents the preliminary results of the computational modeling, with special emphasis on the comparison between homogeneous and heterogeneous moderator systems, in terms of the core diameter required to meet a specific multiplication factor (k eff = 1.035). This comparison illustrates the impact of moderator configuration on the size and performance of a LEU-fueled kilowatt-class space nuclear reactor. (author)

  1. Development of very-high-density low-enriched-uranium fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Trybus, C.L.; Wiencek, T.C.

    1997-01-01

    Following a hiatus of several years and following its successful development and qualification of 4.8 g U cm -3 U 3 Si 2 -Al dispersion fuel for application with low-enriched uranium in research and test reactors, the US Reduced Enrichment for Research and Test Reactors program has embarked on the development of even-higher-density fuels. Our goal is to achieve uranium densities of 8-9 g cm -3 in aluminum-based dispersion fuels. Achieving this goal will require the use of high-density, γ-stabilized uranium alloy powders in conjunction with the most-advanced fuel fabrication techniques. Key issues being addressed are the reaction of the fuel alloys with aluminum and the irradiation behavior of the fuel alloys and any reaction products. Test irradiations of candidate fuels in very-small (micro) plates are scheduled to begin in the Advanced Test Reactor during June, 1997. Initial results are expected to be available in early 1998. We are performing out-of-reactor studies on the phase structure of the candidate alloys on diffusion of the matrix material into the aluminum. In addition, we are modifying our current dispersion fuel irradiation behavior model to accommodate the new fuels. Several international partners are participating in various phases of this work. (orig.)

  2. Moderator configuration options for a low-enriched uranium fueled Kilowatt-class Space Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    King, Jeffrey C., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Colorado School of Mines (CSM), Golden, CO (United States); Mencarini, Leonardo de Holanda; Guimaraes, Lamartine N. F., E-mail: guimaraes@ieav.cta.br, E-mail: mencarini@ieav.cta.br [Instituto de Estudos Avancados (IEAV), Sao Jose dos Campos, SP (Brazil). Divisao de Energia Nuclear

    2015-07-01

    The Brazilian Air Force, through its Institute for Advanced Studies (Instituto de Estudos Avancados, IEAv/DCTA), and the Colorado School of Mines (CSM) are studying the feasibility of a space nuclear reactor with a power of 1-5 kW{sub e} and fueled with Low-Enriched Uranium (LEU). This type of nuclear reactor would be attractive to signatory countries of the Non-Proliferation Treaty (NPT) or commercial interests. A LEU-fueled space reactor would avoid the security concerns inherent with Highly Enriched Uranium (HEU) fuel. As an initial step, the HEU-fueled Kilowatt Reactor Using Stirling Technology (KRUSTY) designed by the Los Alamos National Laboratory serves as a basis for a similar reactor fueled with LEU fuel. Using the computational code MCNP6 to predict the reactor neutronics performance, the size of the resulting reactor fueled with 19.75 wt% enriched uranium-10 wt% molybdenum alloy fuel is adjusted to match the excess reactivity of KRUSTY. Then, zirconium hydride moderator is added to the core to reduce the size of the reactor. This work presents the preliminary results of the computational modeling, with special emphasis on the comparison between homogeneous and heterogeneous moderator systems, in terms of the core diameter required to meet a specific multiplication factor (k{sub eff} = 1.035). This comparison illustrates the impact of moderator configuration on the size and performance of a LEU-fueled kilowatt-class space nuclear reactor. (author)

  3. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    International Nuclear Information System (INIS)

    Montierth, Leland M.

    2016-01-01

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  4. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  5. Operational impacts of low-enrichment uranium fuel conversion on the Ford Nuclear Reactor

    International Nuclear Information System (INIS)

    Bernal, F.E.; Brannon, C.C.; Burgard, N.E.; Burn, R.R.; Cook, G.M.; Simpson, P.A.

    1985-01-01

    The University of Michigan Department of Nuclear Engineering and the Michigan Memorial-Phoenix Project have been engaged in a cooperative effort with Argonne National Laboratory to test and analyze low-enrichment fuel in the Ford Nuclear Reactor (FNR). The effort was begun in 1979, as part of the Reduced Enrichment Research and Test Reactor Program, to demonstrate on a whole-core basis the feasibility of enrichment reduction from 93% to <20% in Materials Test Reactor-type fuel designs. The first low-enrichment uranium (LEU) core was loaded into the FNR and criticality was achieved on December 8, 1981. The final LEU core was established October 11, 1984. No significant operational impacts have resulted from conversion of the FNR to LEU fuel. Thermal flux in the core has decreased slightly; thermal leakage flux has increased. Rod worths, temperature coefficient, and void coefficient have changed imperceptibly. Impressions from the operators are that power defect has increased slightly and that fuel lifetime has increased

  6. Conversion and standardization of university reactor fuels using low-enrichment uranium - Options and costs

    International Nuclear Information System (INIS)

    Harris, D.R.; Matos, J.E.; Young, H.H.

    1985-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. The U.S. Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the U.S. Department of Energy. (author)

  7. Conversion and standardization of university reactor fuels using low-enrichment uranium - options and costs

    International Nuclear Information System (INIS)

    Harris, D.R.; Matos, J.E.; Young, H.H.

    1985-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. The US Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the US Department of Energy. 20 refs., 1 tab

  8. CONCEPTUAL PROCESS DESCRIPTION FOR THE MANUFACTURE OF LOW-ENRICHED URANIUM-MOLYBDENUM FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Daniel M. Wachs; Curtis R. Clark; Randall J. Dunavant

    2008-02-01

    The National Nuclear Security Agency Global Threat Reduction Initiative (GTRI) is tasked with minimizing the use of high-enriched uranium (HEU) worldwide. A key component of that effort is the conversion of research reactors from HEU to low-enriched uranium (LEU) fuels. The GTRI Convert Fuel Development program, previously known as the Reduced Enrichment for Research and Test Reactors program was initiated in 1978 by the United States Department of Energy to develop the nuclear fuels necessary to enable these conversions. The program cooperates with the research reactors’ operators to achieve this goal of HEU to LEU conversion without reduction in reactor performance. The programmatic mandate is to complete the conversion of all civilian domestic research reactors by 2014. These reactors include the five domestic high-performance research reactors (HPRR), namely: the High Flux Isotope Reactor at the Oak Ridge National Laboratory, the Advanced Test Reactor at the Idaho National Laboratory, the National Bureau of Standards Reactor at the National Institute of Standards and Technology, the Missouri University Research Reactor at the University of Missouri–Columbia, and the MIT Reactor-II at the Massachusetts Institute of Technology. Characteristics for each of the HPRRs are given in Appendix A. The GTRI Convert Fuel Development program is currently engaged in the development of a novel nuclear fuel that will enable these conversions. The fuel design is based on a monolithic fuel meat (made from a uranium-molybdenum alloy) clad in Al-6061 that has shown excellent performance in irradiation testing. The unique aspects of the fuel design, however, necessitate the development and implementation of new fabrication techniques and, thus, establishment of the infrastructure to ensure adequate fuel fabrication capability. A conceptual fabrication process description and rough estimates of the total facility throughput are described in this document as a basis for

  9. Calculation of parameters for inspection planning and evaluation: low enriched uranium conversion and fuel fabrication facilities

    International Nuclear Information System (INIS)

    Reardon, P.T.; Mullen, M.F.; Harms, N.L.

    1981-02-01

    As part of Task C.35 (Calculation of Parameters for Inspection Planning and Evaluation) of the US Program of Technical Assistance to IAEA Safeguards, Pacific Northwest Laboratory has performed some quantitative analyses of IAEA inspection activities at low-enriched uranium (LEU) conversion and fuel fabrication facilities. This report presents the results and conclusions of those analyses. Implementation of IAEA safeguards at LEU conversion and fuel fabrication facilities must take into account a variety of practical problems and constraints. One of the key concerns is the problem of flow verification, especially product verification. The objective of this report is to help put the problem of flow verification in perspective by presenting the results of some specific calculations of inspection effort and probability of detection for various product measurement strategies. In order to provide quantitative information about the advantages and disadvantages of the various strategies, eight specific cases were examined

  10. Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium

    International Nuclear Information System (INIS)

    Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed

    2013-01-01

    The operations with the fissile materials such as U 235 introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k eff ) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed

  11. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David [ORNL; Chandler, David [ORNL; Cook, David [ORNL; Ilas, Germina [ORNL; Jain, Prashant [ORNL; Valentine, Jennifer [ORNL

    2014-10-30

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The

  12. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David G [ORNL; Chandler, David [ORNL; Cook, David Howard [ORNL; Ilas, Germina [ORNL; Jain, Prashant K [ORNL; Valentine, Jennifer R [ORNL

    2014-11-01

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present

  13. Development of Nitride Coating Using Atomic Layer Deposition for Low-Enriched Uranium Fuel Powder

    Science.gov (United States)

    Bhattacharya, Sumit

    High-performance research reactors require fuel that operates at high specific power and can withstand high fission density, but at relatively low temperatures. The design of the research reactor fuels is done for efficient heat emission, and consists of assemblies of thin-plates cladding made from aluminum alloy. The low-enriched fuels (LEU) were developed for replacing high-enriched fuels (HEU) for these reactors necessitates a significantly increased uranium density in the fuel to counterbalance the decrease in enrichment. One of the most promising new fuel candidate is U-Mo alloy, in a U-Mo/Al dispersion fuel form, due to its high uranium loading as well as excellent irradiation resistance performance, is being developed extensively to convert from HEU fuel to LEU fuel for high-performance research reactors. However, the formation of an interaction layer (IL) between U-Mo particles and the Al matrix, and the associated pore formation, under high heat flux and high burnup conditions, degrade the irradiation performance of the U-Mo/Al dispersion fuel. From the recent tests results accumulated from the surface engineering of low enriched uranium fuel (SELENIUM) and MIR reactor displayed that a surface barrier coating like physical vapor deposited (PVD) zirconium nitride (ZrN) can significantly reduce the interaction layer. The barrier coating performed well at low burn up but above a fluence rate of 5x 1021 ions/cm2 the swelling reappeared due to formation interaction layer. With this result in mind the objective of this research was to develop an ultrathin ZrN coating over particulate uranium-molybdenum nuclear fuel using a modified savannah 200 atomic layer deposition (ALD) system. This is done in support of the US Department of Energy's (DOE) effort to slow down the interaction at fluence rate and reach higher burn up for high power research reactor. The low-pressure Savannah 200 ALD system is modified to be designed as a batch powder coating system using the

  14. Conversion of the University of Missouri-Rolla Reactor from high-enriched uranium to low-enriched uranium fuel

    International Nuclear Information System (INIS)

    Bolon, A.E.; Straka, M.; Freeman, D.W.

    1997-01-01

    The objectives of this project were to convert the UMR Reactor fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel and to ship the HEU fuel back to the Department of Energy Savannah River Site. The actual core conversion was completed in the summer of 1992. The HEU fuel was offloaded to an onsite storage pit where it remained until July, 1996. In July, 1996, the HEU fuel was shipped to the DOE Savannah River Site. The objectives of the project have been achieved. DOE provided the following funding for the project. Several papers were published regarding the conversion project and are listed in the Attachment. In retrospect, the conversion project required much more time and effort than originally thought. Several difficulties were encountered including the unavailability of a shipping cask for several years. The authors are grateful for the generous funding provided by DOE for this project but wish to point out that much of their efforts on the conversion project went unfunded

  15. Low-enriched fuel particle performance review

    International Nuclear Information System (INIS)

    Homan, F.; Nabielek, H.; Yang, L.

    1978-08-01

    The available data on low-enriched uranium (LEU) fuel particles were reviewed under the United States-Federal Republic of Germany Agreement. The most influential factors controlling the irradiation performance of LEU fuel particles were found to be plutonium transport, fission product transport, fuel particle mechanical performance, and fuel particle chemical performance

  16. Materials safeguards and accountability in the low enriched uranium conversion-fabrication sector of the fuel cycle

    International Nuclear Information System (INIS)

    Schneider, R.A.; Nilson, R.; Jaech, J.L.

    1978-01-01

    Today materials accounting in the low enriched conversion-fabrication sector of the LWR fuel cycle is of increased importance. Low enriched uranium is rapidly becoming a precious metal with current dollar values in the range of one dollar per gram comparing with gold and platinum at 7-8 dollars per gram. In fact, people argue that its dollar value exceeds its safeguards value. Along with this increased financial incentive for better material control, the nuclear industry is faced with the impending implementation of international safeguards and increased public attention over its ability to control nuclear materials. Although no quantity of low enriched uranium (LEU) constitutes a practical nuclear explosive, its control is important to international safeguards because of plutonium production or further enrichment to an explosive grade material. The purpose of the paper is to examine and discuss some factors in the area of materials safeguards and accountability as they apply to the low enriched uranium conversion-fabrication sector. The paper treats four main topics: basis for materials accounting; our assessment of the proposed new IAEA requirements; adequacy of current practices; and timing and direction of future modifications

  17. Reactivity worth of the thermal column of a MTR type swimming pool research reactor using low enriched uranium fuel

    International Nuclear Information System (INIS)

    Ali Khan, L.; Ahmad, N.

    2002-01-01

    The reactivity worth of the thermal column of a typical MTR type swimming pool research reactor using low enriched uranium fuel has been determined by modeling the core using standard computer codes. It was also measured experimentally by operating the reactor in the stall and open ends. The calculated value of the reactivity worth of the thermal column is about 14% greater than the experimentally determined value

  18. Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

    An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

  19. From high enriched to low enriched uranium fuel in research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Van Den Berghe, S.; Leenaers, A.; Koonen, E.; Moons, F.; Sannen, L. [Nuclear Materials Science Institute, SCK.CEN, Boeretang 200, B-2400 Mol (Belgium)

    2010-07-01

    Since the 1970's, global efforts have been going on to replace the high-enriched (>90% {sup 235}U), low-density UAlx research reactor fuel with high-density, low enriched (<20% {sup 235}U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has presently been obtained with U{sub 3}Si{sub 2} dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher density, which will also allow the conversion of some high flux research reactors that currently cannot change to U{sub 3}Si{sub 2} (eg. BR2 in Belgium), have continued and are for the moment mainly directed towards the U(Mo) alloy fuel (7-10 w% Mo). This paper provides an overview of the past efforts and presents the current status of the U(Mo) development. (authors)

  20. From high enriched to low enriched uranium fuel in research reactors

    International Nuclear Information System (INIS)

    Van Den Berghe, S.; Leenaers, A.; Koonen, E.; Moons, F.; Sannen, L.

    2010-01-01

    Since the 1970's, global efforts have been going on to replace the high-enriched (>90% 235 U), low-density UAlx research reactor fuel with high-density, low enriched ( 235 U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has presently been obtained with U 3 Si 2 dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher density, which will also allow the conversion of some high flux research reactors that currently cannot change to U 3 Si 2 (eg. BR2 in Belgium), have continued and are for the moment mainly directed towards the U(Mo) alloy fuel (7-10 w% Mo). This paper provides an overview of the past efforts and presents the current status of the U(Mo) development. (authors)

  1. Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David G [ORNL; Cook, David Howard [ORNL; Freels, James D [ORNL; Griffin, Frederick P [ORNL; Ilas, Germina [ORNL; Sease, John D [ORNL; Chandler, David [ORNL

    2012-03-01

    This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

  2. Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

    International Nuclear Information System (INIS)

    Renfro, David G.; Cook, David Howard; Freels, James D.; Griffin, Frederick P.; Ilas, Germina; Sease, John D.; Chandler, David

    2012-01-01

    This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

  3. Thermal-hydraulic analysis for core conversion to the use of low-enriched uranium fuels in the KUR

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Kanda, Keiji; Shibata, Toshikazu

    1985-01-01

    A feasibility study has been performed on the core conversion to the use of low-enriched uranium (LEU) fuels in the KUR. Five fuel element geometries are studied. For each fuel element, the relation between the pressure drop and the flow rate, critical heat flux, and heat fluxes for the onset of flow instability and the onset of nucleate boiling are calculated using the computer code PLTEMP3.MOD1 which has been developed for this analysis. The effect of fuel material (UAl x -Al, U 3 O 8 -Al and U 3 Si 2 -Al) on the peak fuel temperatures is also studied. A particular interest in the mixed core which may be constructed on the way to the use of LEU fuels, the change in the bypass flow rate due to the change in the gap between different fuel elements is investigated. (author)

  4. Preliminary experience and near future utilization programmes of the MPR-30 fueled by LEU [low enriched uranium

    International Nuclear Information System (INIS)

    Arbie, B.; Soentono, S.

    1987-01-01

    The MTR type reactor MPR-30 G.A. Siwabessy, located at PUSPIPTEK Serpong has recently reached its first criticality. This multipurpose reactor is supposed to be the first MTR type reactor in the world that is designed and constructed to be fueled by low enriched uranium. Preliminary experience covering the approach to the first criticality and the excess reactivity loading as well as some thermal hydraulics and power ascension tests are briefly presented and discussed. The near future utilization programmes during and after commissioning are also presented. (Author)

  5. Establishing a Cost Basis for Converting the High Flux Isotope Reactor from High Enriched to Low Enriched Uranium Fuel

    International Nuclear Information System (INIS)

    Primm, Trent; Guida, Tracey

    2010-01-01

    Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration/Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.

  6. Progress in developing very-high-density low-enriched-uranium fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Hayes, S.L.; Wiencek, T.C.; Strain, R.V.

    1999-01-01

    Preliminary results from the postirradiation examinations of microplates irradiated in the RERTR-1 and -2 experiments in the ATR have shown several binary and ternary U-Mo alloys to be promising candidates for use in aluminum-based dispersion fuels with uranium densities up to 8 to 9 g/cm 3 . Ternary alloys of uranium, niobium, and zirconium performed poorly, however, both in terms of fuel/matrix reaction and fission-gas-bubble behavior, and have been dropped from further study. Since irradiation temperatures achieved in the present experiments (approximately 70 deg. C) are considerably lower than might be experienced in a high-performance reactor, a new experiment is being planned with beginning-of-cycle temperatures greater than 200 deg. C in 8-g U/cm 3 fuel. (author)

  7. Development of very-high-density low-enriched uranium fuels

    International Nuclear Information System (INIS)

    Snegrove, J.L.; Hofmann, G.L.; Trybus, C.L.; Wiencek, T.C.

    1997-01-01

    The RERTR (=Reduced Enrichment for Research and Test Reactors) program has begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun. (author)

  8. Development of very high-density low-enriched uranium fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L.; Wiencek, T.C.

    1997-02-01

    The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and first results should be available by end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

  9. Evaluating the effectiveness of dilution of the recovered uranium with depleted uranium and low-enriched uranium to obtain fuel for VVER reactors

    International Nuclear Information System (INIS)

    Smirnov, A Yu; Sulaberidze, G A; Dudnikov, A A; Nevinitsa, V A

    2016-01-01

    The possibility of the recovered uranium enrichment in a cascade of gas centrifuges with three feed flows (depleted uranium, low-enriched uranium, recovered uranium) with simultaneous dilution of U-232,234,236 isotopes was shown. A series of numerical experiments were performed for different content of U-235 in low-enriched uranium. It has been demonstrated that the selected combination of diluents can simultaneously reduce the cost of separative work and the consumption of natural uranium, not only with respect to the previously used multi-flow cascade schemes, but also in comparison to the standard cascade for uranium enrichment. (paper)

  10. The proposed use of low enriched uranium fuel in the High Flux Australian Reactor (HIFAR)

    International Nuclear Information System (INIS)

    Vittorio, D.; Durance, G.

    2002-01-01

    The Australian Nuclear Science and Technology Organisation (ANSTO) operates the High Flux Australian Reactor (HIFAR). HIFAR commenced operation in the late 1950's with fuel elements containing uranium enriched to 93%. From that time the level of enrichment has gradually decreased to the current level of 60%. It is now proposed to further reduce the enrichment of HIFAR fuel to <20% by utilising LEU fuel assemblies manufactured by RISO National Laboratory, that were originally intended for use in the DR-3 reactor. Minor modifications have been made to the assemblies to adapt them for use in HIFAR. A detailed design review has been performed and initial safety analysis and reactor physics calculations are to be submitted to ARPANSA as part of a four-stage approval process. (author)

  11. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE and AFTER IRRADIATION

    International Nuclear Information System (INIS)

    SCHWINKENDORF, K.N.

    2006-01-01

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements. The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprising two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with ''green'' (fresh) fuel and one with spent fuel. Both the green and spent fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k eff = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, 3 green fuel and 4 spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements can supply useful

  12. Detailed description of an SSAC at the facility level for a low-enriched uranium conversion and fuel fabrication facility

    International Nuclear Information System (INIS)

    Jones, R.J.

    1984-09-01

    Some States have expressed a need for more detailed guidance with regard to the technical elements in the design and operation of SSACs for both the national and the international objectives. To meet this need the present document has been prepared, describing the technical elements of an SSAC in considerable detail. The purpose of this document is therefore, to provide a detailed description of a system for the accounting for and control of nuclear material in a model low enriched uranium conversion and fuel fabrication facility which can be used by a facility operator to establish his own system in a way which will provide the necessary information for compliance with a national system for nuclear material accounting and control and for the IAEA to carry out its safeguards responsibilities

  13. DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Almond, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-28

    The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Site (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.

  14. Latest developments in rolled fuels for materials-testing reactors: a trend towards the use of low-enriched uranium

    International Nuclear Information System (INIS)

    Fanjas, Y.

    1981-01-01

    The properties of rolled fuels and the work done in this field by CERCA is described. The technology developed conforms to low enrichment requirements, whilst guaranteeing a satisfactory level of reactor performance [fr

  15. Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Sease, J.D.; Primm, R.T. III; Miller, J.H.

    2007-01-01

    The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are within the

  16. Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sease, J.D.; Primm, R.T. III; Miller, J.H.

    2007-09-30

    The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are

  17. Replacement of highly enriched uranium by medium or low-enriched uranium in fuels for research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    To exclude the possibility of an explosive use of the uranium obtained from an elementary chemical process, one needs to use a fuel less enriched than 20 weight percent in U 235 . This goal can be reached by two ways: 1. The low density fuels, i.e. U or U 3 O 8 /Al fuels. One has to increase their U content from 1.3 g U/cm 3 presently qualified under normal operation conditions. Several manufacturers such as CERCA in France developed these fuels with a near-term objective of about 2 g U/cm 3 and a long-term objective of 3 g U/cm 3 . 2. The high density fuels. They are the UO 2 Caramel plate type fuels now under consideration, and U 3 Si and UMo as a long-term potential

  18. Fabrication and irradiation testing of LEU [low enriched uranium] fuels at CRNL status as of 1987 September

    International Nuclear Information System (INIS)

    Sears, D.F.; Berthiaume, L.C.; Herbert, L.N.

    1987-01-01

    The current status of Chalk River Nuclear Laboratories' (CRNL) program to develop and test low-enriched uranium (LEU), proliferation-resistant fuels for use in research reactors is reviewed. CRNL's fuel manufacturing process has been qualified by the successful demonstration irradiation of 7 full-size rods in the NRU reactor. Now industrial-scale production equipment has been commissioned, and a fuel-fabrication campaign for 30 NRU rods and a MAPLE-X core is underway. Excess capacity could be used for commercial fuel fabrication. In the irradiation testing program, mini-elements with deliberately included core surface defects performed well in-reactor, swelling by only 7 to 8 vol% at 93 atomic percent burnup of the original U-235. The additional restraint provided by the aluminium cladding which flowed into the defects during extrusion contributed to this good performance. Mini-elements containing a variety of particle size distributions were also successfully irradiated to 93 at% burnup in NRU, as part of a study to establish the optimum particle size distribution. Swelling was found to be proportional to the percentage of fines (<44μm particles) contained in the cores. The mini-elements containing the composition normally used at CRNL had swollen by 5.8 vol%, and mini-elements with a much higher percentage of fines had swollen by 6.8 vol%, at 93 at% burnup. Also, a program to develop LEU targets for Mo-99 production, via the technology developed to fabricate dispersed silicide fuel, has started, and preliminary scoping studies are underway. (Author)

  19. Development program for fuel elements with low enriched uranium for high temperature reactors

    International Nuclear Information System (INIS)

    1987-12-01

    The results of HTR fuel development taking place at the THTR's can be summarized as follows for the main points of core manufacture coating matrix and fuel emenent manufacture: 1. The well known gel precipitation process was modified for the manufacture of UO 2 cores. 2. The TRISO coating (additional SiC layer between two very dense PyC layers) can be applied with the required quality on an economical 10 kg scale. 3. The particle fracture in the complete fuel element due to manufacture was lowered during the course of the project to below the target values of -6 U/U total. For testing fuel elements, the required irradiation samples were designed in agreement with the reactor constructors, were prepared and the first phase of the irradiation program was successfully completed in the context of the HBK project. (orig./HP) [de

  20. The use of low enriched uranium fuel cycle in high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    The present paper begins with a brief review of the status of research and development of experimental VHTR in Japan. On the basis of the experience gained from these work, assessment is made of commercial HTRs. Material balance with fuel burnup is calculated for the two core models; one is HTGR for steam cycle and the other VHTR for process heat application. The results of assessment of commercial HTRs are compared with those for LWR

  1. Study on usage of low enriched uranium Russian type fuel elements for design of an experimental ADS research reactor

    International Nuclear Information System (INIS)

    Pesic, M.P.

    2005-01-01

    Conceptual design of an accelerator driven sub-critical experimental research reactor (ADSRR) was initiated in 1999 at the Vinca Institute of Nuclear Sciences, Serbia and Montenegro. Initial results of neutronic analyses of the proposed ADSRR-H were carried out by Monte Carlo based codes and available high-enriched uranium dioxide (HEU) dispersed Russian type TVR-S fuel elements (FE) placed in a lead matrix. Beam of charged particles (proton or deuteron) would be extracted from the high-energy channel H5B of the VINCY cyclotron of the TESLA Accelerator Installation. In 2002, the Vinca Institute has, in compliance with the Reduced Enrichment for Research and Test Reactors (RERTR) Program, returned fresh HEU TVR-S type FEs back to the Russian Federation. Since usage of HEU FEs in research reactors is not further recommended, a new study of an ADSRR-L conceptual design has initiated in Vinca Institute in last two years, based on assumed availability of low-enriched uranium (LEU) dispersed type TVR-S FEs. Initial results of numerical simulations of this new ADSRR-L, published for the first time in this paper, shows that such a small low neutron flux system can be used as an experimental - 'demonstration' - ADS with neutron characteristics similar to proposed well-known lead moderated and cooled power sub-critical ADS with intermediate neutron spectrum. Neutron spectrum characteristics of the ADSRR-L are compared to ones of the ADSRR-H with the same mass (7.7 g) of 235 U nuclide per TVR-S FE. (author)

  2. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    Energy Technology Data Exchange (ETDEWEB)

    Evans, Louise G., E-mail: evanslg@ornl.gov [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Swinhoe, Martyn T.; Menlove, Howard O. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Schwalbach, Peter; Baere, Paul De [European Commission, Euratom Safeguards Office (Luxembourg); Browne, Michael C. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2013-11-21

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd{sub 2}O{sub 3}) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available {sup 241}AmLi (α,n) interrogation source strength of 5.7×10{sup 4} s{sup −1}. Furthermore, the calibration range of the new collar has been extended to verify {sup 235}U content in variable PWR fuel

  3. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    International Nuclear Information System (INIS)

    Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

    2013-01-01

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd 2 O 3 ) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241 AmLi (α,n) interrogation source strength of 5.7×10 4 s −1 . Furthermore, the calibration range of the new collar has been extended to verify 235 U content in variable PWR fuel designs in the presence of up to

  4. Production of Mo-99 using low-enriched uranium silicide

    International Nuclear Information System (INIS)

    Hutter, J.C.; Srinivasan, B.; Vicek, M.; Vandegrift, G.F.

    1994-01-01

    Over the last several years, uranium silicide fuels have been under development as low-enriched uranium (LEU) targets for Mo-99. The use of LEU silicide is aimed at replacing the UAl x alloy in the highly-enriched uranium dissolution process. A process to recover Mo-99 from low-enriched uranium silicide is being developed at Argonne National Laboratory. The uranium silicide is dissolved in alkaline hydrogen peroxide. Experiments performed to determine the optimum dissolution procedure are discussed, and the results of dissolving a portion of a high-burnup (>40%) U 3 Si 2 miniplate are presented. Future work related to Mo-99 separation and waste disposal are also discussed

  5. Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan)], E-mail: mfarhan_73@yahoo.co.uk; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan)

    2008-09-15

    The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease.

  6. Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2008-01-01

    The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease

  7. 31 CFR 540.308 - Low Enriched Uranium (LEU).

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Low Enriched Uranium (LEU). 540.308... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.308 Low Enriched Uranium (LEU). The term low enriched...

  8. Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N. (U. of Cincinnati)

    2006-02-01

    A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

  9. Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

    International Nuclear Information System (INIS)

    Primm, R.T. III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N.

    2006-01-01

    A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U 3 O 8 mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties

  10. Post-pulse detail metallographic examinations of low-enriched uranium silicide plate-type miniature fuel

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1991-10-01

    Pulse irradiation at Nuclear Safety Research Reactor (NSRR) was performed using low-enriched (19.89 w% 235 U) unirradiated silicide plate-type miniature fuel which had a density of 4.8 gU/cm 3 . Experimental aims are to understand the dimensional stability and to clarify the failure threshold of the silicide plate-type miniature fuel under power transient conditions through post-pulse detail metallographic examinations. A silicide plate-type miniature fuel was loaded into an irradiation capsule and irradiated by a single pulse. Deposited energies given in the experiments were 62, 77, 116 and 154 cal/g·fuel, which lead to corresponding peak fuel plate temperatures, 201 ± 28degC, 187 ± 10degC, 418 ± 74degC and 871 ± 74degC, respectively. Below 400degC, reliability and dimensional stability of the silicide plate fuel was sustained, and the silicide plate fuel was intact. Up to 540degC, wall-through intergranular crackings occurred in the Al-3%Mg alloy cladding. With the increase of the temperature, the melting of the aluminum cladding followed by recrystallization, the denudation of fuel core and the plate-through intergranular cracking were observed. With the increase of the temperature beyond 400degC, the bowing of fuel plate became significant. Above the temperature of 640degC molten aluminum partially reacted with the fuel core, partially flowed downward under the influence of surface tension and gravity, and partially formed agglomerations. Judging from these experimental observations, the fuel-plate above 400degC tends to reduce its dimensional stability. Despite of the apparent silicide fuel-plate failure, neither generation of pressure pulse nor that of mechanical energy occurred at all. (J.P.N.)

  11. Low-enriched fuel particle performance review

    International Nuclear Information System (INIS)

    Homan, F.; Nabielek, H.; Yang, L.

    1978-08-01

    The available data on low-enriched (LEU) fuel particles were reviewed under the United States-Federal Republic of Germany Agreement. The most influential factors controlling the irradiation performance of LEU fuel particles were found to be plutonium transport, fission product transport, fuel particle mechanical performance and fuel particle chemical performande. (orig.) [de

  12. Effects of high density dispersion fuel loading on the uncontrolled reactivity insertion transients of a low enriched uranium fueled material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)], E-mail: farhan73@hotmail.com; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)

    2009-08-15

    The effects of using high density low enriched uranium on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U-Mo (9w/o) LEU fuels currently being developed under the RERTR program having uranium densities of 6.57 gU/cm{sup 3}, 7.74 gU/cm{sup 3} and 8.57 gU/cm{sup 3}. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained increased by 5.8% as uranium density increases from 6.57 gU/cm{sup 3} to 8.90 gU/cm{sup 3}. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved at the new power level, by 4.7 K, 4.4 K and 2.4 K, respectively. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point.

  13. Development of quality assurance methods for low enriched fuel assemblies

    International Nuclear Information System (INIS)

    Woolstenhulme, N.E.; Moore, G.A.; Perez, D.M.; Wachs, D.M.

    2010-01-01

    As the Reduced Enrichment for Research and Test Reactors (RERTR) fuel development program has furthered the technology of low enriched uranium fuels, much effort has been expended to specify requirements, perform appropriate inspections, and to qualify experimental fuel plates and assemblies for irradiation. A great deal of consideration has been given to generate examinations and criteria that are both applicable to the unique fuel types being developed and consistent with industry practices for inspecting plate-type reactor fuel. Recent developments in quality assurance (QA) methodologies have given a heightened confidence in satisfactory fuel plate performance. At the same time, recommendations are given to further develop a system suitable for the testing and acceptance of production fuel elements containing low enriched uranium fuels. (author)

  14. Development of long-life low enrichment fuel

    International Nuclear Information System (INIS)

    Gietzen, A.J.; West, G.B.

    1978-01-01

    With only a few exceptions, TRIGA reactors have always used low-enriched-uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on non-proliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U. S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of this year, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with GA's standard commercial warranty

  15. Proposal of new 235U nuclear data to improve keff biases on 235U enrichment and temperature for low enriched uranium fueled lattices moderated by light water

    International Nuclear Information System (INIS)

    Wu, Haicheng; Okumura, Keisuke; Shibata, Keiichi

    2005-06-01

    The under prediction of k eff depending on 235 U enrichment in low enriched uranium fueled systems, which had been a long-standing puzzle especially for slightly enriched ones, was studied in this report. Benchmark testing was carried out with several evaluated nuclear data files, including the new uranium evaluations from preliminary ENDF/B-VII and CENDL-3.1. Another problem reviewed here was k eff underestimation vs. temperature increase, which was observed in the sightly enriched system with recent JENDL and ENDF/B uranium evaluations. Through the substitute analysis of nuclear data of 235 U and 238 U, we propose a new evaluation of 235 U data to solve both of the problems. The new evaluation was tested for various uranium fueled systems including low or highly enriched metal and solution benchmarks in the ICSBEP handbook. As a result, it was found that the combination of the new evaluation of 235 U and the 238 U data from the preliminary ENDF/B-VII gives quite good results for most of benchmark problems. (author)

  16. Investigations of uraniumsilicide-based dispersion fuels for the use of low enrichment uranium (LEU) in research and test reactors

    International Nuclear Information System (INIS)

    Nazare, S.

    1982-07-01

    The work presents at the outset, a review of the preparation and properties of uranium silicides (U 3 Si and U 3 Si 2 ) in so far as these are relevant for their use as dispersants in research reactor fuels. The experimental work deals with the preparation and powder metallurgical processing of Al-clad miniature fuel element plates with U 3 Si- und U 3 Si-Al up to U-densities of 6.0 g U/cm 3 . The compatibility of these silicides with the Al-matrix under equilibrium conditions (873 K) and the influence of the reaction on the dimensional stability of the miniplates is described and discussed. (orig.) [de

  17. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  18. Field test of short-notice random inspections for inventory-change verification at a low-enriched-uranium fuel-fabrication plant: Preliminary summary

    International Nuclear Information System (INIS)

    Fishbone, L.G.; Moussalli, G.; Naegele, G.; Ikonomou, P.; Hosoya, M.; Scott, P.; Fager, J.; Sanders, C.; Colwell, D.; Joyner, C.J.

    1994-01-01

    An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. This report details a six-month field test of the feasibility of such SNRIs which took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division. Westinghouse personnel made daily declarations about both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ''mailbox''. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. Items from both strata were verified during the SNRIs by means of nondestructive assay equipment. The field test demonstrated the feasibility and practicality of key elements of the SNRI approach for a large LEU fuel fabrication plant

  19. A study of a zone approach to IAEA [International Atomic Energy Agency] safeguards: The low-enriched-uranium zone of a light-water-reactor fuel cycle

    International Nuclear Information System (INIS)

    Fishbone, L.G.; Higinbotham, W.A.

    1986-06-01

    At present the IAEA designs its safeguards approach with regard to each type of nuclear facility so that the safeguards activities and effort are essentially the same for a given type and size of nuclear facility wherever it may be located. Conclusions regarding a state are derived by combining the conclusions regarding the effectiveness of safeguards for the individual facilities within a state. In this study it was convenient to define three zones in a state with a closed light-water-reactor nuclear fuel cycle. Each zone contains those facilities or parts thereof which use or process nuclear materials of the same safeguards significance: low-enriched uranium, radioactive spent fuel, or recovered plutonium. The possibility that each zone might be treated as an extended material balance area for safeguards purposes is under investigation. The approach includes defining the relevant features of the facilities in the three zones and listing the safeguards activities which are now practiced. This study has focussed on the fresh-fuel zone, the several facilities of which use or process low-enriched uranium. At one extreme, flows and inventories would be verified at each material balance area. At the other extreme, the flows into and out of the zone and the inventory of the whole zone would be verified. There are a number of possible safeguards approaches which fall between the two extremes. The intention is to develop a rational approach which will make it possible to compare the technical effectiveness and the inspection effort for the facility-oriented approach, for the approach involving the zone as a material balance area, and for some reasonable intermediate safeguards approaches

  20. A study of a zone approach to IAEA (International Atomic Energy Agency) safeguards: The low-enriched-uranium zone of a light-water-reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Fishbone, L.G.; Higinbotham, W.A.

    1986-06-01

    At present the IAEA designs its safeguards approach with regard to each type of nuclear facility so that the safeguards activities and effort are essentially the same for a given type and size of nuclear facility wherever it may be located. Conclusions regarding a state are derived by combining the conclusions regarding the effectiveness of safeguards for the individual facilities within a state. In this study it was convenient to define three zones in a state with a closed light-water-reactor nuclear fuel cycle. Each zone contains those facilities or parts thereof which use or process nuclear materials of the same safeguards significance: low-enriched uranium, radioactive spent fuel, or recovered plutonium. The possibility that each zone might be treated as an extended material balance area for safeguards purposes is under investigation. The approach includes defining the relevant features of the facilities in the three zones and listing the safeguards activities which are now practiced. This study has focussed on the fresh-fuel zone, the several facilities of which use or process low-enriched uranium. At one extreme, flows and inventories would be verified at each material balance area. At the other extreme, the flows into and out of the zone and the inventory of the whole zone would be verified. There are a number of possible safeguards approaches which fall between the two extremes. The intention is to develop a rational approach which will make it possible to compare the technical effectiveness and the inspection effort for the facility-oriented approach, for the approach involving the zone as a material balance area, and for some reasonable intermediate safeguards approaches.

  1. Field test of short-notice random inspections for inventory-change verification at a low-enriched-uranium fuel-fabrication plant

    International Nuclear Information System (INIS)

    Fishbone, L.G.; Moussalli, G.; Naegele, G.

    1995-01-01

    An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. Then a statistical inference can be made from verification results for items verified during SNRIs to the entire populations, i.e. the entire strata, even if inspectors were not present when many items were received or produced. A six-month field test of the feasibility of such SNRIs took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division during 1993. Westinghouse personnel made daily declarations about both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ''mailbox''. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. They arrived unannounced at the plant, in most cases immediately after travel from Canada, where the IAEA maintains a regional office. Items from both strata were verified during the SNRIs by meant of nondestructive assay equipment

  2. A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Drera, Saleem S., E-mail: saleem.drera@gmail.com [Mechanical Engineering, Colorado School of Mines, Golden, CO 80401 (United States); Hofman, Gerard L. [Argonne National Laboratory, Chicago, IL 60439 (United States); Kee, Robert J. [Mechanical Engineering, Colorado School of Mines, Golden, CO 80401 (United States); King, Jeffrey C. [Metallurgical and Materials Engineering, Colorado School of Mines, Golden, CO 80401 (United States)

    2014-10-15

    Highlights: • This article presents a cellular automata (CA) algorithm to synthesize the growth of intermetallic interaction layers in U–Mo/Al dispersion fuel. • The method utilizes a 3D representation of the fuel, which is discretized into separate voxels that can change identy based on derived CA rules. • The CA model is compared to ILT measurements for RERTR experimental data. • The primary objective of the model is to synthesize three-dimensional microstructures that can be used in subsequent thermal and mechanical modeling. • The CA model can be used for predictive analysis. For example, it can be used to study the dependence of temperature on interaction layer growth. - Abstract: Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium–molybdenum (U–Mo) particles within an aluminum matrix. Fresh U–Mo particles typically range between 10 and 100 μm in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction–diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the present paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates.

  3. The low-enrichment fuel development program

    International Nuclear Information System (INIS)

    Stahl, D.

    1993-01-01

    In the 1950s and 1960s, low-power research reactors were built around the world utilized MTR-type fuel elements containing 20% enriched uranium. However, the demand for higher specific power created a need for greater uranium-235 concentrations. Early difficulties in increasing uranium content led to the substitution of highly enriched uranium in place of the 20% enriched fuel previously utilized. The highly enriched material also yielded other benefits including longer core residence time, higher specific reactivity, and somewhat lower cost. Highly enriched material then became readily available and was used for high-power reactors as well as in low-power reactors where 20% enriched material would have sufficed. The trend toward higher and higher specific power also led to the development of the dispersion-type fuels which utilized highly enriched uranium at a concentration of about 40 wt%. In the 1970's, however, concerns were raised about the proliferation resistance of fuels and fuel cycles. As a consequence, the U.S. Department of State has recently prohibited the foreign shipment of highly enriched material, except where prior contractual obligation or special merit exists. This will impact on the availability and utilization of highly enriched uranium for research and test reactor fuel. It has also stimulated development programs on fuels with higher uranium content which would allow the use of uranium of lower enrichment. The purpose of this report is to briefly describe the overall fuel-development program which is coordinated by Argonne National Laboratory for the Department of Energy, and to indicate the current and potential uranium loadings. Other reports will address the individual fuel-development activities in greater detail

  4. Licensing considerations in converting NRC-licensed non-power reactors from high-enriched to low-enriched uranium fuels

    International Nuclear Information System (INIS)

    Carter, R.E.

    1985-01-01

    During the mid-1970s, there was increasing concern with the possibility that highly enriched uranium (HEU), widely used in non-power reactors around the world, might be diverted from its intended peaceful uses. In 1982 the U.S. Nuclear Regulatory Commission (NRC) issued a policy statement that was intended to conform with the perceived international thinking, and that addressed the two relevant areas in which NRC has statutory responsibility, namely, export of special nuclear materials for non-USA non-power reactors, and the licensing of USA-based non-power reactors not owned by the Federal government. To further address the second area, NRC issued a proposed rule for public comment that would require all NRC-licensed non-power reactors using HEU to convert to low enriched uranium (LEU) fuel, unless they could demonstrate a unique purpose. Currently the NRC staff is revising the proposed rule. An underlying principle guiding the staff is that as long as a change in enrichment does not lead to safety-related reactor modifications, and does not involve an unreviewed safety question, the licensee could convert the core without prior NRC approval. At the time of writing this paper, a regulatory method of achieving this principle has not been finalized. (author)

  5. Licensing considerations in converting NRC-licensed non-power reactors from high-enriched to low-enriched uranium fuels

    Energy Technology Data Exchange (ETDEWEB)

    Carter, R E

    1985-07-01

    During the mid-1970s, there was increasing concern with the possibility that highly enriched uranium (HEU), widely used in non-power reactors around the world, might be diverted from its intended peaceful uses. In 1982 the U.S. Nuclear Regulatory Commission (NRC) issued a policy statement that was intended to conform with the perceived international thinking, and that addressed the two relevant areas in which NRC has statutory responsibility, namely, export of special nuclear materials for non-USA non-power reactors, and the licensing of USA-based non-power reactors not owned by the Federal government. To further address the second area, NRC issued a proposed rule for public comment that would require all NRC-licensed non-power reactors using HEU to convert to low enriched uranium (LEU) fuel, unless they could demonstrate a unique purpose. Currently the NRC staff is revising the proposed rule. An underlying principle guiding the staff is that as long as a change in enrichment does not lead to safety-related reactor modifications, and does not involve an unreviewed safety question, the licensee could convert the core without prior NRC approval. At the time of writing this paper, a regulatory method of achieving this principle has not been finalized. (author)

  6. Comparison of the thorium- and low-enriched uranium fuel cycle in the OTTO pebble bed HTR

    Energy Technology Data Exchange (ETDEWEB)

    Teuchert, E; Maly, V

    1973-03-15

    From the study of the physical, technical and economical properties of the LOTTO and TOTTO fuel cycle the authors draw the conclusion that no fundamental reason can be found for a definite preference of one of the two cycles. Either of them can be developed to become a very attractive concept. The flexibility of this reactor allows the design performance in that way that the reactor becomes a safe system with reasonable economy. The decision for the preference of the LOTTO or TOTTO can be allowed to be governed by the requirements of the outer fuel cycle or by politics.

  7. Some economic aspects of the low enriched uranium production

    International Nuclear Information System (INIS)

    1990-05-01

    At the Technical Committee Meeting on Economics of Low Enriched Uranium 14 papers were presented. A separate abstract was prepared for each of these papers. The five technical sessions covered several economic aspects of uranium concentrates production, conversion into uranium hexafluoride and uranium enrichment and the recycling of U and Pu in LWR. Four Panel discussions were held to discuss the uranium market trends, the situation of conversion industry, the reprocessing and the uranium market, the future trends of enrichment and the economics of LWRs compared with other reactors. Refs, figs and tabs

  8. Research and Development of Multiphysics Models in Support of the Conversion of the High Flux Isotope Reactor to Low Enriched Uranium Fuel

    International Nuclear Information System (INIS)

    Bodey, Isaac T.; Curtis, Franklin G.; Arimilli, Rao V.; Ekici, Kivanc; Freels, James D.

    2015-01-01

    The findings presented in this report are results of a five year effort led by the RRD Division of the ORNL, which is focused on research and development toward the conversion of the High Flux Isotope Reactor (HFIR) fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU). This report focuses on the tasks accomplished by the University of Tennessee Knoxville (UTK) team from the Department of Mechanical, Aerospace, and Biomedical Engineering (MABE) that provided expert support in multiphysics modeling of complex problems associated with the LEU conversion of the HFIR reactor. The COMSOL software was used as the main computational modeling tool, whereas Solidworks was also used in support of computer-aided-design (CAD) modeling of the proposed LEU fuel design. The UTK research has been governed by a statement of work (SOW), which was updated annually to clearly define the specific tasks reported herein. Ph.D. student Isaac T. Bodey has focused on heat transfer and fluid flow modeling issues and has been aided by his major professor Dr. Rao V. Arimilli. Ph.D. student Franklin G. Curtis has been focusing on modeling the fluid-structure interaction (FSI) phenomena caused by the mechanical forces acting on the fuel plates, which in turn affect the fluid flow in between the fuel plates, and ultimately the heat transfer, is also affected by the FSI changes. Franklin Curtis has been aided by his major professor Dr. Kivanc Ekici. M.Sc. student Adam R. Travis has focused two major areas of research: (1) on accurate CAD modeling of the proposed LEU plate design, and (2) reduction of the model complexity and dimensionality through interdimensional coupling of the fluid flow and heat transfer for the HFIR plate geometry. Adam Travis is also aided by his major professor, Dr. Kivanc Ekici. We must note that the UTK team, and particularly the graduate students, have been in very close collaboration with Dr. James D. Freels (ORNL technical monitor and mentor) and have

  9. Research and Development of Multiphysics Models in Support of the Conversion of the High Flux Isotope Reactor to Low Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Curtis, Franklin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Arimilli, Rao V. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ekici, Kivanc [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-11-01

    The findings presented in this report are results of a five year effort led by the RRD Division of the ORNL, which is focused on research and development toward the conversion of the High Flux Isotope Reactor (HFIR) fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU). This report focuses on the tasks accomplished by the University of Tennessee Knoxville (UTK) team from the Department of Mechanical, Aerospace, and Biomedical Engineering (MABE) that provided expert support in multiphysics modeling of complex problems associated with the LEU conversion of the HFIR reactor. The COMSOL software was used as the main computational modeling tool, whereas Solidworks was also used in support of computer-aided-design (CAD) modeling of the proposed LEU fuel design. The UTK research has been governed by a statement of work (SOW), which was updated annually to clearly define the specific tasks reported herein. Ph.D. student Isaac T. Bodey has focused on heat transfer and fluid flow modeling issues and has been aided by his major professor Dr. Rao V. Arimilli. Ph.D. student Franklin G. Curtis has been focusing on modeling the fluid-structure interaction (FSI) phenomena caused by the mechanical forces acting on the fuel plates, which in turn affect the fluid flow in between the fuel plates, and ultimately the heat transfer, is also affected by the FSI changes. Franklin Curtis has been aided by his major professor Dr. Kivanc Ekici. M.Sc. student Adam R. Travis has focused two major areas of research: (1) on accurate CAD modeling of the proposed LEU plate design, and (2) reduction of the model complexity and dimensionality through interdimensional coupling of the fluid flow and heat transfer for the HFIR plate geometry. Adam Travis is also aided by his major professor, Dr. Kivanc Ekici. We must note that the UTK team, and particularly the graduate students, have been in very close collaboration with Dr. James D. Freels (ORNL technical monitor and mentor) and have

  10. Low enrichment of uranium in the light of the nuclear weapon problem

    International Nuclear Information System (INIS)

    Barstad, G.

    1979-09-01

    A difficult problem in the immediate future will be to direct civil nuclear technology in such a way that the ability to produce nuclear weapons by additional countries is prevented. There are two main problems. First, enrichment plants can be used to produce high enriched uranium, which can be used in nuclear weapons, as well as low enriched reactor fuel. Second, plutonium produced during reactor operation can be used as nuclear weapon material, as well as for nuclear fuel. The problem discussed here is particularly the development of an enrichment process which is economic for low enriched reactor fuel, but which may not easily be adapted to produce high enriched uranium. (JIW)

  11. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

  12. Low-enriched uranium holdup measurements in Kazakhstan

    International Nuclear Information System (INIS)

    Barham, M.A.; Ceo, R.; Smith, S.E.

    1998-01-01

    Quantification of the residual nuclear material remaining in process equipment has long been a challenge to those who work with nuclear material accounting systems. Fortunately, nuclear material has spontaneous radiation emissions that can be measured. If gamma-ray measurements can be made, it is easy to determine what isotope a deposit contains. Unfortunately, it can be quite difficult to relate this measured signal to an estimate of the mass of the nuclear deposit. Typically, the measurement expert must work with incomplete or inadequate information to determine a quantitative result. Simplified analysis models, the distribution of the nuclear material, any intervening attenuation, background(s), and the source-to-detector distance(s) can have significant impacts on the quantitative result. This presentation discusses the application of a generalized-geometry holdup model to the low-enriched uranium fuel pellet fabrication plant in Ust-Kamenogorsk, Kazakhstan. Preliminary results will be presented. Software tools have been developed to assist the facility operators in performing and documenting the measurements. Operator feedback has been used to improve the user interfaces

  13. Simulation of transportation of low enriched uranium solutions

    International Nuclear Information System (INIS)

    Hope, E.P.; Ades, M.J.

    1996-01-01

    A simulation of the transportation by truck of low enriched uranium solutions has been completed for NEPA purposes at the Savannah River Site. The analysis involves three distinct source terms, and establishes the radiological risks of shipment to three possible destinations. Additionally, loading accidents were analyzed to determine the radiological consequences of mishaps during handling and delivery. Source terms were developed from laboratory measurements of chemical samples from low enriched uranium feed materials being stored at SRS facilities, and from manufacturer data on transport containers. The transportation simulations were accomplished over the INTERNET using the DOE TRANSNET system at Sandia National Laboratory. The HIGHWAY 3.3 code was used to analyze routing scenarios, and the RADTRAN 4 code was used to analyze incident free and accident risks of transporting radiological materials. Loading accidents were assessed using the Savannah River Site AXAIR89Q and RELEASE 2 codes

  14. Assay of low-enriched uranium using spontaneous fission neutrons

    International Nuclear Information System (INIS)

    Zucker, M.S.; Fainberg, A.

    1980-01-01

    Low-enriched uranium oxide in bulk containers can be assayed for safeguards purposes, using the neutrons from spontaneous fission of 238 U as a signature, to complement enrichment and mass measurement. The penetrability of the fast fission neutrons allows the inner portion of bulk samples to register. The measurement may also be useful for measuring moisture content, of significance in process control. The apparatus used can be the same as for neutron correlation counting for Pu assay. The neutron multiplication observed in 238 U is of intrinsic interest

  15. Low Enrichment Uranium (LEU)-fueled SLOWPOKE-2 nuclear reactor simulation with the Monte-Carlo based MCNP 4A code

    International Nuclear Information System (INIS)

    Pierre, J.R.M.

    1996-01-01

    Following the commissioning of the Low Enrichment Uranium (LEU) Fuelled SLOWPOKE-2 research reactor at the Royal Military College-College Militaire Royal (RMC-CMR), excess reactivity measurements were conducted over a range of temperature and power. The results showed a maximum excess reactivity of 3.37 mk at 33 o C. Several deterministic models using computer codes like WIMS-CRNL, CITATION, TRIVAC and DRAGON have been used to try to reproduce the excess reactivity and temperature trend of both the LEU and HEU SLOWPOKE-2 reactors. The best simulations had been obtained at Ecole Polytechnique de Montreal. They were able to reproduce the temperature trend of their HEU-fuelled reactor using TRIVAC calculations, but this model over-estimated the absolute value of the excess reactivity by 119 mk. Although calculations using DRAGON did not reproduce the temperature trend as well as TRIVAC, these calculations represented a significant improvement on the absolute value at 20 o C reducing the discrepancy to 13 mk. Given the advance in computer technology, a probabilistic approach was tried in this work, using the Monte-Carlo N-Particle Transport Code System MCNP 4A, to model the RMC-CMR SLOWPOKE-2 reactor.

  16. Low enrichment fuel development at INEL

    International Nuclear Information System (INIS)

    Newton, D.G.

    1993-01-01

    EG and G Idaho, Inc. is under contract to the Department of Energy to operate the Idaho National Engineering Laboratory (INEL). The INEL is located in southeastern Idaho. This facility has been operating since 1949 and was originally called the National Reactor Testing Station. Several contractors manage projects on this facility. Most projects at INEL are concerned with either reactor safety or irradiation testing. At Test Area North, for example, experiments are being conducted on the effects of loss of coolant. At the Test Reactor Area the ATR (Advanced Test Reactor) and ETR (Engineering Test Reactor) are used for irradiation testing and, of course, those of you working at Argonne will recognize the Experimental Breeder Reactors I and II. SPERT is an acronym for Special Power Excursion Reactor Test. A part of this former reactor facility has been converted into a fuel fabrication laboratory facility. At SPERT IV a miniature fabrication facility has been set up to duplicate the aluminide plate fuel processing line at Atomics International. In other words, a model of the supplier's processing has been created, so that what process changes are developed here can then be scaled up to production. The process is described showing: making UAI x powder, making compact for fuel core, making experimental fuel plate and compact assembly, inspection and testing the fuel plate. Main concern was related to possible swelling

  17. Selection and use of a low enriched fuel in high performance research reactors

    International Nuclear Information System (INIS)

    Cerles, J.M.; Schwartz, J.P.

    1978-08-01

    A new nuclear fuel composition for research reactors (Osiris, Siloe) is studied using low enriched (E<20%) uranium oxide. Its utilization leads to modifications in the facilities of these experimental reactors: increase of primary coolant flow, modifications in failed element detection system, handling of materials and storage

  18. Structure and thermal properties of as-fabricated U-7Mo/Mg and U-10Mo/Mg low-enriched uranium research reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kulakov, Mykola, E-mail: mykola.kulakov@cnl.ca [Fuel Development Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Saoudi, Mouna [Fuel Development Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Piro, Markus H.A. [Fuel and Fuel Channel Safety Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Donaberger, Ronald L. [Canadian Neutron Beam Centre, Chalk River, ON K0J 1J0 Canada (Canada)

    2017-02-15

    Aluminum-clad U-7Mo/Mg and U-10Mo/Mg pin-type mini-elements (with a core uranium loading of 4.5 gU/cm{sup 3}) have been fabricated at the Canadian Nuclear Laboratories for experimental tests and ultimately for use in research and test reactors. In this study, the microstructure and phase composition of unirradiated U-7Mo/Mg and U-10Mo/Mg fuel cores were analyzed using optical and scanning electron microscopy, and neutron powder diffraction. Thermal properties were characterized using a combination of experimental measurements and thermodynamic calculations. The thermal diffusivity was measured using the laser flash method. The temperature-dependent specific heat capacities were calculated based on the linear rule of mixture using the weight fraction of different crystalline phases and their specific heat capacity values taken from the literature. The thermal conductivity was then calculated using the measured thermal diffusivity, the measured density and the calculated specific heat capacity. The resulting thermal conductivity is practically identical for both types of fuel. The in-reactor temperatures were predicted using conjugate heat transfer simulations. - Highlights: • Neutron diffraction analysis shows that most of the γ-U(Mo) phase was retained in as-fabricated U-7Mo/Mg and U-10Mo/Mg fuel cores. • The experimental thermal conductivity of both types of fuel is practically identical. • Based on conjugate heat transfer simulations, under normal operating conditions, the in-reactor fuel centreline temperature is about 510 K.

  19. Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Ross, Kyle W. (Los Alamos National Laboratory, Los Alamos, NM); Smith, James Dean; Longmire, Pamela

    2010-04-01

    The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

  20. A comparison between thorium-uranium and low enrichment uranium cycles in the high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cerles, J M

    1973-03-15

    In a previous report, it was shown that the Uranium cycle could be used as well with multi-hole block (GGA type) as with tubular elements. Now, in a F.S.V. geometry, a comparison is made between Thorium cycle and Uranium cycle. This comparison will be concerned with the physical properties of the materials, the needs of natural Uranium, the fissile material inventory and, at last, an attempt of economical considerations. In this report the cycle will be characterizd by the fertile material. So, we write ''Thorium cycle'' for Highly Enriched Uranium - Thorium cycle and ''Uranium cycle'' for low Enrichment Uranium cycle.

  1. Some Main Results of Commissioning of the Dalat Research Reactor with Low Enriched Fuel

    International Nuclear Information System (INIS)

    Nguyen Nhi Dien; Luong Ba Vien; Pham Van Lam; Le Vinh Vinh; Huynh Ton Nghiem

    2014-01-01

    After completion of design calculation of the Dalat Nuclear Research Reactor (DNRR) for conversion from high-enriched uranium fuel (HEU) to low-enriched uranium (LEU) fuel, the commissioning programme for DNRR with entire core loaded with LEU fuel was successfully carried out from 24 November 2011 to 13 January 2012. The experimental results obtained during the implementation of commissioning programme showed a good agreement with design calculations and affirmed that the DNRR with LEU core have met all safety and exploiting requirements. (author)

  2. Status report on conversion of the Georgia Tech Research Reactor to low enrichment fuel

    International Nuclear Information System (INIS)

    Karam, R.A.; Matos, J.E.; Mo, S.C.; Woodruff, W.L.

    1995-01-01

    The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. The US Department of Energy is funding a program to compare reactor performance with high and low enrichment fuels. The goals of the program are: (1) to amend the SAR and the technical specifications of the GTRR so that LEU U 3 Si 2 -Al dispersion fuel plates can replace the current HEU U-Al alloy fuel, and (2) to optimize the LEU core such that maximum value neutron beams can be extracted for possible neutron capture therapy application. This paper presents a status report on the LEU conversion effort. (author)

  3. Status report on conversion of the Georgia Tech Research Reactor to low enrichment fuel

    International Nuclear Information System (INIS)

    Karam, R.A.; Matos, J.E.; Mo, S.C.; Woodruff, W.L.

    1991-01-01

    The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. The US Department of Energy is funding a program to compare reactor performance with high and low enrichment fuels. The goals of the program are: (1) to amend the SAR and the Technical Specifications of the GTRR so that LEU U 3 Si 2 -Al dispersion fuel plates can replace the current HEU U-Al alloy fuel, and (2) to optimize the LEU core such that maximum value neutron beams can be extracted for possible neutron capture therapy application. This paper presents a status report on the LEU conversion effort

  4. DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

    Energy Technology Data Exchange (ETDEWEB)

    Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

    2011-02-01

    This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

  5. Continuing investigations for technology assessment of 99Mo production from LEU [low enriched uranium] targets

    International Nuclear Information System (INIS)

    Vandegrift, G.F.; Kwok, J.D.; Marshall, S.L.; Vissers, D.R.; Matos, J.E.

    1987-01-01

    Currently much of the world's supply of 99m Tc for medical purposes is produced from 99 Mo derived from the fissioning of high enriched uranium (HEU). This paper presents the results of our continuing studies on the effects of substituting low enriched uranium (LEU) for HEU in targets for the production of fission product 99 Mo. Improvements in the electrodeposition of thin films of uranium metal continue to increase the appeal for the substitution of LEU metal for HEU oxide films in cylindrical targets. The process is effective for targets fabricated from stainless steel or zircaloy. Included is a cost estimate for setting up the necessary equipment to electrodeposit uranium metal on cylindrical targets. Further investigations on the effect of LEU substitution on processing of these targets are also reported. Substitution of uranium silicides for the uranium-aluminium alloy or uranium aluminide dispersed fuel used in current target designs will allow the substitution of LEU for HEU in these targets with equivalent 99 Mo-yield per target and no change in target geometries. However, this substitution will require modifications in current processing steps due to 1) the insolubility of uranium silicides in alkaline solutions and 2) the presence of significant quantities of silicate in solution. Results to date suggest that substitution of LEU for HEU can be achieved. (Author)

  6. Nuclear characteristics evaluation for Kyoto University Research Reactor with low-enriched uranium core

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Ken; Unesaki, Hironobu [Kyoto University Research Reactor Institute, Kumatori-cho Sennan-gun Osaka (Japan)

    2008-07-01

    A project to convert the fuel of Kyoto University Research Reactor (KUR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) is in progress as a part of RERTR program. Prior to the operation of LEU core, the nuclear characteristics of the core have been evaluated to confirm the safety operation. In the evaluation, nuclear parameters, such as the excess reactivity, shut down margin control rod worth, reactivity coefficients, were calculated, and they were compared with the safety limits. The results of evaluation show that the LEU core is able to satisfy the safety requirements for operation, i.e. all the parameters satisfy the safety limits. Consequently, it was confirmed that the LEU fuel core has the proper nuclear characteristics for the safety operation. (authors)

  7. A data base for PHW reactor operating on a once-through, low enriched uranium-thorium cycle

    International Nuclear Information System (INIS)

    Lungu, S.

    1984-04-01

    The study of a detailed data base for a new once-through uranium-thorium cycle using low enriched uranium (4 and 5,5% wt. U-235) and distinct UO 2 and ThO 2 fuel channels has been performed. With reference to a standard 638 MWe CANDU-type PHWR with 380 channels, evaluation of economics, fuel behaviour and safety has been performed. The Feinberg-Galanin method (code FEINGAL) has been used for calculation of axial flux distribution. All parameters have been provided by LATREP code following up the irradiation history. Economical assessment has shown that this fuel cycle is competitive with the natural uranium fuel cycle for 1979-based values of the parameters. Fuel behaviour and safety features modelling has shown that core behaviour of the uranium-thorium reactor under abnormal and accident conditions would be at least as good as that of the standard natural uranium reactor

  8. Performance and economic penalties of some LEU [low enriched uranium] conversion options for the Australian Reactor HIFAR

    International Nuclear Information System (INIS)

    McCulloch, D.B.; Robinson, G.S.

    1987-01-01

    Performance calculations for the conversion of HIFAR to low enriched uranium (LEU) fuel have been extended to a wide range of 235 U loadings per fuel element. Using a simple approximate algorithm for the likely costs of LEU compared with highly enriched uranium (HEU) fuel elements, the increases in annual fuelling costs for LEU compared with HEU fuel are examined for a range of conversion options involving different performance penalties. No significant operational/safety problems were found for any of the options canvassed. (Author)

  9. Low enrichment fuel conversion for Iowa State University. Final report

    International Nuclear Information System (INIS)

    Bullen, D.B.; Wendt, S.E.

    1996-01-01

    The UTR-10 research and teaching reactor at Iowa State University (ISU) has been converted from high-enriched fuel (HEU) to low- enriched fuel (LEU) under Grant No. DE-FG702-87ER75360 from the Department of Energy (DOE). The original contract period was August 1, 1987 to July 31, 1989. The contract was extended to February 28, 1991 without additional funding. Because of delays in receiving the LEU fuel and the requirement for disassembly of the HEU assemblies, the contract was renewed first through May 31, 1992, then through May 31, 1993 with additional funding, and then again through July 31, 1994 with no additional funding. In mid-August the BMI cask was delivered to Iowa State. Preparations are underway to ship the HEU fuel when NRC license amendments for the cask are approved

  10. PULSTAR fuel, low enrichment, long lifetime, economical, proven

    International Nuclear Information System (INIS)

    Carter, Robert E.; Leonard, Bobby E.

    1993-01-01

    sufficiently low enrichment to preclude use as nuclear weapons material. Although Pu is produced, its continuous burnup precludes the use of the fuel to produce Pu for nuclear weapons material. The fuel is readily available on the world-wide market, and therefore purchasers do not have to depend on one sole-source fuel supplier. The fuel is inherently safe against any conceivable accident condition. Unlike other research reactor fuel, the fuel can be reprocessed with present technology and present reprocessing facilities. Due to the similarity to power reactor fuel, the fuel facilitates research related to fuel and core development for power reactors. The under moderated neutron fluence spectrum enables studies of fast neutron effects and provides high thermal neutron fluence peaking at the core boundary and in 'flux traps'. The long fuel burnup cycle provides a fuel lifetime at least a factor of 4 greater than competitive low enrichment 20%) fuels. The cost per megawatt day of operation is a factor of 2 less than competitive low enrichment 20% fuel

  11. PULSTAR fuel, low enrichment, long lifetime, economical, proven

    Energy Technology Data Exchange (ETDEWEB)

    Carter, Robert E; Leonard, Bobby E [Institute for Resource Management, Inc., Bethesda, MD (United States)

    1993-08-01

    is of sufficiently low enrichment to preclude use as nuclear weapons material. Although Pu is produced, its continuous burnup precludes the use of the fuel to produce Pu for nuclear weapons material. The fuel is readily available on the world-wide market, and therefore purchasers do not have to depend on one sole-source fuel supplier. The fuel is inherently safe against any conceivable accident condition. Unlike other research reactor fuel, the fuel can be reprocessed with present technology and present reprocessing facilities. Due to the similarity to power reactor fuel, the fuel facilitates research related to fuel and core development for power reactors. The under moderated neutron fluence spectrum enables studies of fast neutron effects and provides high thermal neutron fluence peaking at the core boundary and in 'flux traps'. The long fuel burnup cycle provides a fuel lifetime at least a factor of 4 greater than competitive low enrichment 20%) fuels. The cost per megawatt day of operation is a factor of 2 less than competitive low enrichment 20% fuel.

  12. Refueling the RPI reactor facility with low-enrichment fuel

    International Nuclear Information System (INIS)

    Harris, D.R.; Rodriguez-Vera, F.; Wicks, F.E.

    1985-01-01

    The RPI Critical Facility has operated since 1963 with a core of thin, highly enriched fuel plates in twenty-five fuel assembly boxes. A program is underway to refuel the reactor with 4.81 w/o enriched SPERT (F-1) fuel rods. Use of these fuel rods will upgrade the capabilities of the reactor and will eliminate a security risk. Adequate quantities of SPERT (F-1) fuel rods are available, and their use will result in a great cost saving relative to manufacturing new low-enrichment fuel plates. The SPERT fuel rods are 19 inches longer than are the present fuel plates, so a modified core support structure is required. It is planned to support and position the SPERT fuel pins by upper and lower lattice plates, thus avoiding the considerable cost of new fuel assembly boxes. The lattice plates will be secured to the existing top and bottom plates. The design permits the fabrication and use of other lattice plates for critical experiment research programs in support of long-lived full development for power reactors. (author)

  13. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

    2009-03-01

    This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

  14. Continuing investigations for technology assessment of 99Mo production from LEU [low enriched Uranium] targets

    International Nuclear Information System (INIS)

    Vandergrift, G.F.; Kwok, J.D.; Marshall, S.L.; Vissers, D.R.; Matos, J.E.

    1987-01-01

    Currently much of the world's supply of /sup 99m/Tc for medical purposes is produced from 99 Mo derived from the fissioning of high enriched uranium (HEU). The need for /sup 99m/Tc is continuing to grow, especially in developing countries, where needs and national priorities call for internal production of 99 Mo. This paper presents the results of our continuing studies on the effects of substituting low enriched Uranium (LEU) for HEU in targets for the production of fission product 99 Mo. Improvements in the electrodeposition of thin films of uranium metal are reported. These improvements continue to increase the appeal for the substitution of LEU metal for HEU oxide films in cylindrical targets. The process is effective for targets fabricated from stainless steel or hastaloy. A cost estimate for setting up the necessary equipment to electrodeposit uranium metal on cylindrical targets is reported. Further investigations on the effect of LEU substitution on processing of these targets are also reported. Substitution of uranium silicides for the uranium-aluminum alloy or uranium aluminide dispersed fuel used in other current target designs will allow the substitution of LEU for HEU in these targets with equivalent 99 Mo-yield per target and no change in target geometries. However, this substitution will require modifications in current processing steps due to (1) the insolubility of uranium silicides in alkaline solutions and (2) the presence of significant quantities of silicate in solution. Results to date suggest that both concerns can be handled and that substitution of LEU for HEU can be achieved

  15. Atomics International fuel fabrication facility and low enrichment program [contributed by T.A. Moss, AI

    International Nuclear Information System (INIS)

    Moss, T.A.

    1993-01-01

    The AI facility is approximately 30,000 square feet in area and consists of four general areas. One area is devoted to the production of UAl x powder. It consists of a series of arc melting furnaces, crushing lines, glove boxes, and compacting presses. The second area is used for the rolling of fuel plates. The third area is used for the machining of the plates to final size and also the machining of the fuel elements. In the fourth area the fuel plates are swaged into assemblies, and all welding and inspection operations are performed. As part of the lower enrichment program we are scheduled to put a second UAl x powder line into operation and we have had to expand some of our storage area. Under the low enrichment program the AI fuel facility will be modified to accommodate a separate low enrichment Al x production line and compacting line. This facility modification should be done by the end of the fiscal year. We anticipate producing fuel with an enrichment slightly less than 20% We anticipate powder being available for plate production shortly after the facility is completed. Atomics International is scheduled to conduct plate LEU verification work using fully enriched material in the June-July time period, at which time we will investigate what level of uranium loadings we can go to using the current process. It is anticipated that 55 volume percent uranium compound in our fuel form can be achieved

  16. Using low-enriched uranium in research reactors: The RERTR program

    International Nuclear Information System (INIS)

    Travelli, A.

    1994-01-01

    The goal of the RERTR program is to minimize and eventually eliminate use of highway enriched uranium (HEU) in research and test reactors. The program has been very successful, and has developed low-enriched uranium (LEU) fuel materials and designs which can be used effectively in approximately 90 percent of the research and test reactors which used HEU when the program began. This progress would not have been possible without active international cooperation among fuel developers, commercial vendors, and reactor operators. The new tasks which the RERTR program is undertaking at this time include development of new and better fuels that will allow use of LEU fuels in all research and test reactors; cooperation with Russian laboratories, which will make it possible to minimize and eventually eliminate use of HEU in research reactors throughout the world, irrespective of its origin; and development of an LEU-based process for the production of 99 Mo. Continuation and intensification of international cooperation are essential to the achievement of the ultimate goals of the RERTR program

  17. Supply of low enriched (LEU) and highly enriched uranium (HEU) for research reactors

    International Nuclear Information System (INIS)

    Mueller, H.

    1997-01-01

    Enriched uranium for research reactors in the form of LEU /= low enriched uranium at 19.75% U-235) and HEU (= highly enriched uranium at 90 to 93% U-235) was and is - due to its high U-235 enrichment - a political fuel other than enriched uranium for power reactors. The sufficient availability of LEU and HEU is a vital question for research reactors, especially in Europe, in order to perform their peaceful research reactor programs. In the past the USA were in the Western hemisphere sole supplier of LEU and HEU. Today the USA have de facto stopped the supply of LEU and HEU, for HEU mainly due to political reasons. This paper deals, among others, with the present availability of LEU and HEU for European research reactors and touches the following topics: - historical US supplies, - influence of the RERTR-program, - characteristics of LEU and HEU, - military HEU enters the civil market, -what is the supply situation for LEU and HEU today? - outlook for safe supplies of LEU and HEU. (author)

  18. Structure, conduct, and sustainability of the international low-enriched fuel fabrication industry

    International Nuclear Information System (INIS)

    Rothwell, Geoffrey

    2008-01-01

    This paper examines the cost structures of fabricating Low-Enriched Uranium fuel (LEU, enriched to 5% enrichment) light water reactor fuels. The LEU industry is decades old, and (except for high entry cost, i.e., the cost of designing and licensing a fuel fabrication facility and its fuel), labor and additional fabrication lines can be added by industry incumbents at Nth-of-a-Kind cost to the maximum capacity allowed by the license. On the other hand, new entrants face higher First-of-a-Kind costs and high new-facility licensing costs, increasing the scale required for entry thus discouraging small scale entry by countries with only a few nuclear power plants. Therefore, the industry appears to be competitive with sustainable investment in fuel-cycle states, and structural barriers-to-entry increase its proliferation resistance. (author)

  19. Conversion of the Worcester Polytechnic Institute nuclear reactor to low enriched uranium

    International Nuclear Information System (INIS)

    Newton, T.H. Jr.

    1991-01-01

    The Training Reactor was converted to Low-Enriched Uranium (LEU) aluminide fuel in 1988 and 1989. Tests on the Highly-Enriched Uranium (HEU) core and LEU cores were performed and comparisons made. The testing consisted of critical loading, thermal neutron flux distribution, excess reactivity, regulating blade reactivity worth, and temperature coefficient of reactivity measurement. Comparisons between the LEU and HEU showed that the critical loading configurations were somewhat different with the HEU core consisting of 24 elements and the LEU core consisting of 21 1/3 elements with excess reactivities of 0.24% ΔK/K for the HEU and 0.16% for the LEU. Thermal neutron flux distributions showed similar trends in both the LEU and HEU cores. The regulating blade worth showed a larger LEU value due to thermal peaking in the blade region and temperature coefficients showed a more negative LEU value due to Doppler broadening. Low induced activity of the HEU fuel permitted shipment to the Westinghouse Savannah River Facility using DOT-6M type B containers on 8 August, 1989. (orig.)

  20. Post-irradiation studies of test plates for low enriched fuel elements for research reactors

    International Nuclear Information System (INIS)

    Groos, E.; Buecker, H.J.; Derz, H.; Schroeder, R.

    1988-07-01

    In developing new fuels for research reactor elements that allow the use of low enriched uranium (LEU) 3 Si 2 , U 3 Si 1.5 , U 3 Si 1.3 and U 3 Si. Even up to high burnup rates (80% fifa) U 3 Si 2 was proved to be a reliable fuel that according to the test results achieved to date complies with all necessary requirements above all with respect to dimensional stability. U 3 Si showed significant changes of the fuel microstructure associated with considerably higher fuel swelling, that will probably exclude its use in research reactor operation. The irradiation of U 3 Si 1.3 and U 3 Si 1.5 plates had to be terminated untimely. Up to a burnup of 40% fifa these plates behaved quite well. An extrapolation to higher burnup rates, however only seems to be possible with reservations. (orig./HP) [de

  1. 78 FR 66898 - Low Enriched Uranium From France: Final Results of Changed Circumstances Review

    Science.gov (United States)

    2013-11-07

    ... DEPARTMENT OF COMMERCE International Trade Administration [A-427-818] Low Enriched Uranium From... Administration, International Trade Administration, Department of Commerce. SUMMARY: The Department of Commerce...: Andrew Huston or Mark Hoadley, AD/CVD Operations, Office VII, Enforcement and Compliance, International...

  2. Atomics International fuel fabrication facility and low enrichment program. Part 2

    International Nuclear Information System (INIS)

    Hassel, H.W.

    1993-01-01

    Most of you know our company from the last meeting in May in Vienna, so I won't steal your time with explaining and demonstrating the same techniques that we have heard this morning f rom the other speakers. I would just take some words to explain the order of business with highly enriched uranium. NUKEM handles around almost two tons of highly enriched uranium a year and it was necessary to satisfy all the new physical protection philosophies. That means that we have to install storage and safe fabrication sites for a lot of money, 2.5 meter thick concrete walls, and different alarm systems. So just to demonstrate how silly this business is, we have just overcome this for highly enriched uranium, and now we speak about low enriched uranium for which we don't need all of these investments to make this business safe. I would just like to concentrate my words on the status of fabrication and considerations in my company concerning the medium enriched uranium and low enriched uranium. In TABLE I are the different fuel types (see column 1) and then we have the fabrication in column 2; (The reason that I use the blackboard this morning is that I try to demonstrate all the techniques. However, all the speakers before me did this and in theory we are not so far away from each other.) the experience of my company in kg. In column 3 is the irradiation experience of these fuels types. Column 4 shows the studies and calculations made in our company for lower and medium enriched fuels. The preliminary fabrication tests and calculations are in column 5, and in column 6 we have the delivery time for a prototype core in months after UF 6 supply. Column 7 shows the time for the development of specifications including irradiation time in years for 6 and 7, and column 8 is the estimated cost of 6 and 7. There is just one fuel that is not in this summary and that is U-Zr

  3. Conversion and Blending Facility Highly enriched uranium to low enriched uranium as uranium hexafluoride. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-05

    This report describes the Conversion and Blending Facility (CBF) which will have two missions: (1) convert surplus HEU materials to pure HEU UF{sub 6} and a (2) blend the pure HEU UF{sub 6} with diluent UF{sub 6} to produce LWR grade LEU-UF{sub 6}. The primary emphasis of this blending be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The chemical and isotopic concentrations of the blended LEU product will be held within the specifications required for LWR fuel. The blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry.

  4. Conversion and Blending Facility Highly enriched uranium to low enriched uranium as uranium hexafluoride. Revision 1

    International Nuclear Information System (INIS)

    1995-01-01

    This report describes the Conversion and Blending Facility (CBF) which will have two missions: (1) convert surplus HEU materials to pure HEU UF 6 and a (2) blend the pure HEU UF 6 with diluent UF 6 to produce LWR grade LEU-UF 6 . The primary emphasis of this blending be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The chemical and isotopic concentrations of the blended LEU product will be held within the specifications required for LWR fuel. The blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry

  5. Atomics international fuel fabrication facility and low enrichment program [contributed by H.W. Hassel, NUKEM

    International Nuclear Information System (INIS)

    Hassel, H.W.

    1993-01-01

    NUKEM handles around almost two tons of highly enriched uranium a year and it was necessary to satisfy all the new physical protection philosophies. That means that we have to install storage and safe fabrication sites for a lot of money, 25 meter thick concrete walls, and different alarm systems. So just to demonstrate how silly this business is, we have just overcome this for highly enriched uranium, and now we speak about low enriched uranium for which we don't need all of these investments to make this business safe. I would-just like to concentrate my words on the status of fabrication and considerations in my company concerning the medium enriched uranium and low enriched uranium. In the table are the different fuel types (see column) and then we have the fabrication in column 2 the experience of my comp any in kg. In column 3 is the irradiation experience of these fuels types. Column 4 shows the studies and calculations made in our company for lower and medium enriched fuels. The preliminary fabrication tests and calculations are in column 5, and in column 6 we have the delivery time for a prototype core in months after UF 6 supply. Column 7 shows the time for the development of specifications including irradiation time in years for 6 and 7 and column is the estimated cost of 6 and 7 There is just one fuel that is not in this summary and that is U-Zr. We now see how complex and sophisticated this business is. I have told you already that we have installed for a lot of millions of Deutsche Mark the physical protection, storage vaults and things like that. Now we have to investigate all these different types of fuels for, as you see, a lot of money. Maybe these are a lot of optimistic figures; anyway the question is, does this make all the overall nuclear situation worldwide easier or not. One cannot answer for the moment, but anyway we have a lot of problems

  6. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Foyto, L [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Kutikkad, K [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; McKibben, J C [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Peters, N. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

  7. Reactivity feedback coefficients of a low enriched uranium fuelled material test research reactor at end-of-life

    International Nuclear Information System (INIS)

    Muhammad, Farhan

    2011-01-01

    Highlights: → The isotopic concentration in the fuel changes as soon as it starts its operation. → The neutronic properties of a reactor also change with fuel burnup. → The reactivity feedbacks at end-of-life of a material test reactor fuelled with low enriched uranium fuel are calculated. → Codes used include WIMS-D4 and CITATION. - Abstract: The reactivity feedback coefficients at end-of-life of a material test reactor fuelled with low enriched uranium fuel were calculated. The reactor used for the study was the IAEA's 10 MW benchmark reactor. Simulations were carried out to calculate the different reactivity feedback coefficients including Doppler feedback coefficient, reactivity coefficient for change of water temperature and reactivity coefficient for change of water density. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitude of all the reactivity feedback coefficients increased at end of life of the reactor by almost 2-5%.

  8. Low enrichment fuel conversion for Iowa State University

    International Nuclear Information System (INIS)

    Rohach, A.F.; Hendrickson, R.A.

    1990-08-01

    Work during the reported period was centered primarily in preparation for receiving the LEU fuel and the shipping of the HEU fuel. The LEU fuel has not been received. The HEU fuel assemblies for the UTR-10 reactor will not fit into any current research reactor shipping containers; therefore, the fuel assemblies must be disassembled and the fuel shipped as fuel plates. Procedures and practices have been developed so that the fuel assemblies will be disassembled in a shielded environment

  9. Active-interrogation measurements of fast neutrons from induced fission in low-enriched uranium

    International Nuclear Information System (INIS)

    Dolan, J.L.; Marcath, M.J.; Flaska, M.; Pozzi, S.A.; Chichester, D.L.; Tomanin, A.; Peerani, P.

    2014-01-01

    A detection system was designed with MCNPX-PoliMi to measure induced-fission neutrons from U-235 and U-238 using active interrogation. Measurements were then performed with this system at the Joint Research Centre in Ispra, Italy on low-enriched uranium samples. Liquid scintillators measured induced fission neutrons to characterize the samples in terms of their uranium mass and enrichment. Results are presented to investigate and support the use of organic liquid scintillators with active interrogation techniques to characterize uranium containing materials. -- Highlights: • We studied low-enriched uranium using active-interrogation experiments including a deuterium–tritium neutron generator and an americium–lithium isotopic neutron source. • Liquid scintillators measured induced-fission neutrons from the active-interrogation methods. • Fast-neutron (DT) and thermal-neutron (Am–Li) interrogation resulted in the measurement of trends in uranium mass and 235 U enrichment respectively. • MCNPX-PoliMi, the Monte Carlo transport code, simulated the measured induced-fission neutron trends in the liquid scintillators

  10. Active-interrogation measurements of fast neutrons from induced fission in low-enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    Dolan, J.L., E-mail: jldolan@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Marcath, M.J.; Flaska, M.; Pozzi, S.A. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Chichester, D.L. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Tomanin, A.; Peerani, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Ispra (Italy)

    2014-02-21

    A detection system was designed with MCNPX-PoliMi to measure induced-fission neutrons from U-235 and U-238 using active interrogation. Measurements were then performed with this system at the Joint Research Centre in Ispra, Italy on low-enriched uranium samples. Liquid scintillators measured induced fission neutrons to characterize the samples in terms of their uranium mass and enrichment. Results are presented to investigate and support the use of organic liquid scintillators with active interrogation techniques to characterize uranium containing materials. -- Highlights: • We studied low-enriched uranium using active-interrogation experiments including a deuterium–tritium neutron generator and an americium–lithium isotopic neutron source. • Liquid scintillators measured induced-fission neutrons from the active-interrogation methods. • Fast-neutron (DT) and thermal-neutron (Am–Li) interrogation resulted in the measurement of trends in uranium mass and {sup 235}U enrichment respectively. • MCNPX-PoliMi, the Monte Carlo transport code, simulated the measured induced-fission neutron trends in the liquid scintillators.

  11. Design Study for a Low-enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Ellis, Ronald James [ORNL; Gehin, Jess C [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL

    2007-11-01

    This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with a modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from the current, HEU fuel cycle. Though manufacturing can be accomplished with existing technology

  12. Conversion and Blending Facility highly enriched uranium to low enriched uranium as oxide. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-05

    This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials into pure HEU oxide and (2) blend the pure HEU oxide with depleted and natural uranium oxide to produce an LWR grade LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU will be produced as a waste suitable for storage or disposal.

  13. Low enrichment fuel conversion for Iowa State University

    International Nuclear Information System (INIS)

    Rohach, A.F.; Hendrickson, R.A.

    1991-08-01

    Work during the reported period was centered primarily in preparation for receiving the LEU fuel and the shipping of the HEU fuel. This included development of procedures and tools for the disassembly process. During the period we held many practice sessions applying these tools and practices to a dummy fuel assembly. The LEU fuel was received on April 10, 1991 and the reactor was shut down on May 3, 1991 for refueling. The twelve HEU fuel assemblies in the UTR-10 reactor core were removed and disassembled during the week of May 6--9, 1991. The disassembly process went smoothly with only a few minor problems. Also during this reporting period several experimental measurements and preventative maintenance tasks were accomplished. Finally procedures and practices have been developed for the new LEU fuel loading and critical experiments which are to be completed during the late summer of 1991

  14. Irradiation behavior of low-enriched U/sub 6/Fe-Al dispersion fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Hofman, G.L.; Domagala, R.F.; Copeland, G.L.

    1987-10-01

    An irradiation test of miniature fuel plates containing low-enriched (20% /sup 235/U)U/sub 6/Fe dispersed and clad in Al was performed. The postirradiation examination shows U/sub 6/Fe to form extensive fission gas bubbles at burnups of only approx. = 20% of the original 20% fuel enrichment. Plate failure by fission gas-driven pillowing occurred at approx. = 40% burnup. This places U/sub 6/FE at the lowest burnup capability among low enriched dispersion fuels that have been tested for use in research and test reactors

  15. Validity of Hansen-Roach cross sections in low-enriched uranium systems

    International Nuclear Information System (INIS)

    Busch, R.D.; O'Dell, R.D.

    1991-01-01

    Within the nuclear criticality safety community, the Hansen-Roach 16 group cross section set has been the ''standard'' for use in k eff calculations over the past 30 years. Yet even with its widespread acceptance, there are still questions about its validity and adequacy, about the proper procedure for calculating the potential scattering cross section, σ p , for uranium and plutonium, and about the concept of resonance self shielding and its impact on cross sections. This paper attempts to address these questions. It provides a brief background on the Hansen-Roach cross sections. Next is presented a review of resonances in cross sections, self shielding of these resonances, and the use of σ p to characterize resonance self shielding. Three prescriptions for calculating σ p are given. Finally, results of several calculations of k eff on low-enriched uranium systems are provided to confirm the validity of the Hansen-Roach cross sections when applied to such systems

  16. Validity of Hansen-Roach cross sections in low-enriched uranium systems

    International Nuclear Information System (INIS)

    Busch, R.D.; O'Dell, R.D.

    1991-01-01

    Within the nuclear criticality safety community, the Hansen-Roach 16 group cross section set has been the standard for use in k eff calculations over the past 30 years. Yet even with its widespread acceptance, there are still questions about its validity and adequacy, about the proper procedure for calculating the potential scattering cross section, σ p , for uranium and plutonium, and about the concept of resonance self shielding and its impact on cross sections. This paper attempts to address these questions. It provides a brief background on the Hansen-Roach cross sections. Next is presented a review of resonances in cross sections, self shielding of these resonances, and the use of σ p to characterize resonance self shielding. Three prescriptions for calculating σ p are given. Finally, results of several calculations of k eff on low-enriched uranium systems are provided to confirm the validity of the Hansen-Roach cross sections when applied to such systems. (Author)

  17. Preliminary investigations for technology assessment of 99Mo production from LEU [low enriched uranium] targets

    International Nuclear Information System (INIS)

    Vandegrift, G.F.; Chaiko, D.J.; Heinrich, R.R.; Kucera, E.T.; Jensen, K.J.; Poa, D.S.; Varma, R.; Vissers, D.R.

    1986-11-01

    This paper presents the results of preliminary studies on the effects of substituting low enriched uranium (LEU) for highly enriched uranium (HEU) in targets for the production of fission product 99 Mo. Issues that were addressed are: (1) purity and yield of the 99 Mo//sup 99m/Tc product, (2) fabrication of LEU targets and related concerns, and (3) radioactive waste. Laboratory experimentation was part of the efforts for issues (1) and (2); thus far, radioactive waste disposal has only been addressed in a paper study. Although the reported results are still preliminary, there is reason to be optimistic about the feasibility of utilizing LEU targets for 99 Mo production. 37 refs., 1 fig., 5 tabs

  18. Argentine activities related to the development of low enriched fuel elements

    International Nuclear Information System (INIS)

    Giorsetti, Domingo R.; Perez, Edmundo E.

    1983-01-01

    Within the framework of the RERTR Program and supported by the technical cooperation work agreed upon between the U.S.A. and Argentina in May 1979, the CNEA Nuclear Fuel Department - Low Enriched Fuel Elements Project (ECBE Project), has carried on its own program for developing fuels with low enrichment for research and test reactors. Up to the present, its main objective has been to replace the highly enriched fuel used in its only reactor (RA-3) for research, development and radioisotopes production. The basic stages of the Argentine Program are shown in Table 1. At a meeting held in Vienna in March, 1980, the CNEA stated that its development of fuels with low enrichment would be in two fuel lines: UAl x -Al and U 3 O 8 -Al, and that its aim would be to reach uranium densities of 18-2.2 g/cm 3 for the UAI x -Al line and 2.4-3.0 g/cm 3 for the U 3 O 8 line. At the international meeting held at ANL in November, 1980, and after having received depleted uranium and uranium with 20% and 45% enrichment (purchased from the U.S.A. for manufacturing miniplates and possible standard fuels) to carry on the proposed development, CNEA anticipated -- after its first tests -- that the conditions were satisfactory for reaching uranium densities of 2.4-3.0 g/cm 3 in U 3 O 8 -Al fuel and of 2.4 g/cm 3 in UAI x -Al fuel. In February 1981, after Argentina accepted the obligation of paying for the irradiation service, authorization was obtained for irradiating miniplates in the Oak Ridge Reactor within the RERTR Program. In June 1981, the first set of miniplates was sent to Oak Ridge National Laboratory (ORNL). The maximum actual densities reached at that time were 3.12 g/cm 3 with U 3 O 8 -Al and 2.52 g/cm 3 with UAl x -Al. During a visit of the CNEA Project Technical Manager to the Argonne National Laboratory (ANL) in July 1981, and after exchanging ideas with ANL professional staff, the CNEA decided to incorporate a new line of development, that of U 3 Si-Al. Three months later

  19. Shielding Studies for Reducing the associated Radiological Risks Due To Irradiated Low Enriched Uranium Foil

    International Nuclear Information System (INIS)

    Margeanu, C.A.

    2011-01-01

    Present work estimates the radiation dose rates corresponding to irradiated Low Enriched Uranium (20 wt % 235 U) foil as part of shielding studies for radiological risks reduction after irradiation inside TRIGA 14 MW Research Reactor in an investigation on 99 Mo production possibility. Post-Irradiation Examination Laboratory's cell shielding calculations have been performed; radiation source was obtained by using ORIGEN-S code with specific cross-sections libraries. Different post-irradiation cooling times have been considered, gamma dose rates being estimated by using MAVRIC module from Scale 6 programs package, for following exposure situations (relative to Pie cell): i) front side, ii) lateral side and iii) back side. Three different calculations were performed: a) without any protection shield between operator and cell, except for the cell stainless steel wall; b) with a Lead protection shield between operator and cell and c) with a depleted Uranium shield, located inside the cell in between the radiation source and cell window. Radiation dose rates to cell external wall surface and for other eight fixed distances from cell wall were estimated. To obtain a consistent set of solutions, the study was done for various Uranium foil weights and different Lead and depleted Uranium shields thicknesses. Calculations were focused to assure that the dose rate to an operator positioned at 60 cm working distance from the cell will not exceed 0.02 mSv/h, maximum allowed dose rate for professionally exposed personnel according to Romanian regulations.

  20. Neutronics substantiation of possibility for conversion of the WWR-K reactor core to operation with low-enriched fuel

    International Nuclear Information System (INIS)

    Arinkin, F.M.; Gizatulin, Sh.H.; Zhantikin, T.M.; Koltochnik, S.N.; Takibaev, A.Zh.; Talanov, S.V.; Chakrov, P.V.; Chekushina, L.V.

    2002-01-01

    The studies are aimed to calculation and experimental justification of possibility for conversion of the WWR-R reactor core to low-enriched nuclear fuel (the 19.75-% enrichment in isotope U-235), resulting in reducing the risk of non-sanctioned proliferation of nuclear materials which can be used as weapons materials. The analysis of available published data, related to problem of reduction of enrichment in the fuel used in research thermal reactors, has been carried out. Basing on the analysis results, reference fuel compositions have been chosen, in particular, uranium dioxide (UO 2 ) in aluminum master form and the UA1 4 alloy. Preliminary calculations have shown that, with the WWR-K reactor core preserved existing critical characteristics (the fuel composition: UA1 4 ), the uranium concentration in the fuel element is to be increased by a factor of 2.0-2.2, being impossible technologically. The calculations have been performed by means of the Monte Carlo computational codes. The program of optimal conversion of the WWR-K reactor core to low-enriched fuel has been developed, including: development of calculation models of the reactor core, composed of various designs of fuel elements and fuel assemblies (FA), on a base of corresponding computational codes (diffusion, statistical, etc.); implementation of experiments in the zero-power reactor (critical assembly) with the WWR-C-type FA, in view of correction of the computational constants used in calculations; implementation of reactor core neutronics calculations, in view of selection of the U-235 optimal content in the low-enriched fuel elements and choice of FA reload strategy at the regime of reactor core after burning; determination of the fuel element specification; determination of the critical and operational loads for the reactor core composed of rod/tubular fuel elements; calculation of the efficiency of the protection control system effectors, optimization of its composition, number and locations in the

  1. Low-enriched uranium high-density target project. Compendium report

    Energy Technology Data Exchange (ETDEWEB)

    Vandegrift, George; Brown, M. Alex; Jerden, James L.; Gelis, Artem V.; Stepinski, Dominique C.; Wiedmeyer, Stanley; Youker, Amanda; Hebden, Andrew; Solbrekken, G; Allen, C; Robertson., D; El-Gizawy, Sherif; Govindarajan, Srisharan; Hoyer, Annemarie; Makarewicz, Philip; Harris, Jacob; Graybill, Brian; Gunn, Andy; Berlin, James; Bryan, Chris; Sherman, Steven; Hobbs, Randy; Griffin, F. P.; Chandler, David; Hurt, C. J.; Williams, Paul; Creasy, John; Tjader, Barak; McFall, Danielle; Longmire, Hollie

    2016-09-01

    At present, most 99Mo is produced in research, test, or isotope production reactors by irradiation of highly enriched uranium targets. To achieve the denser form of uranium needed for switching from high to low enriched uranium (LEU), targets in the form of a metal foil (~125-150 µm thick) are being developed. The LEU High Density Target Project successfully demonstrated several iterations of an LEU-fission-based Mo-99 technology that has the potential to provide the world’s supply of Mo-99, should major producers choose to utilize the technology. Over 50 annular high density targets have been successfully tested, and the assembly and disassembly of targets have been improved and optimized. Two target front-end processes (acidic and electrochemical) have been scaled up and demonstrated to allow for the high-density target technology to mate up to the existing producer technology for target processing. In the event that a new target processing line is started, the chemical processing of the targets is greatly simplified. Extensive modeling and safety analysis has been conducted, and the target has been qualified to be inserted into the High Flux Isotope Reactor, which is considered above and beyond the requirements for the typical use of this target due to high fluence and irradiation duration.

  2. Minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, W.C.; Turner, J.C.

    1992-12-01

    A parametric calculational analysis has been performed in order to estimate the minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems. The analysis was performed using a version of the SCALE-4.0 code system and the 27-group ENDF/B-IV cross-section library. Water-moderated uranyl fluoride (UO{sub 2}F{sub 2} and H{sub 2}O) and hydrofluoric-acid-moderated uranium hexaflouride (UF{sub 6} and HF) systems were considered in the analysis over enrichments of 1.4 to 5 wt % {sup 235}U. Estimates of the minimum critical volume, minimum critical mass of uranium, and the minimum mass of moderator required for criticality are presented. There was significant disagreement between the values generated in this study when compared with a similar undocumented study performed in 1983 using ANISN and the Knight-modified Hansen-Roach cross sections. An investigation into the cause of the disagreement was made, and the results are presented.

  3. Minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, W.C.; Turner, J.C.

    1992-12-01

    A parametric calculational analysis has been performed in order to estimate the minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems. The analysis was performed using a version of the SCALE-4.0 code system and the 27-group ENDF/B-IV cross-section library. Water-moderated uranyl fluoride (UO[sub 2]F[sub 2] and H[sub 2]O) and hydrofluoric-acid-moderated uranium hexaflouride (UF[sub 6] and HF) systems were considered in the analysis over enrichments of 1.4 to 5 wt % [sup 235]U. Estimates of the minimum critical volume, minimum critical mass of uranium, and the minimum mass of moderator required for criticality are presented. There was significant disagreement between the values generated in this study when compared with a similar undocumented study performed in 1983 using ANISN and the Knight-modified Hansen-Roach cross sections. An investigation into the cause of the disagreement was made, and the results are presented.

  4. Development of dissolution process for metal foil target containing low enriched uranium

    International Nuclear Information System (INIS)

    Srinivasan, B.; Hutter, J.C.; Johnson, G.K.; Vandegrift, G.F.

    1994-01-01

    About six times more low enriched uranium (LEU) metal is needed to produce the same quantity of 99 Mo as from a high enriched uranium (HEU) oxide target, under similar conditions of neutron irradiation. In view of this, the post-irradiation processing procedures of the LEU target are likely to be different from the Cintichem process procedures now in use for the HEU target. The authors have begun a systematic study to develop modified procedures for LEU target dissolution and 99 Mo separation. The dissolution studies include determination of the dissolution rate, chemical state of uranium in the solution, and the heat evolved in the dissolution reaction. From these results the authors conclude that a mixture of nitric and sulfuric acid is a suitable dissolver solution, albeit at higher concentration of nitric acid than in use for the HEU targets. Also, the dissolver vessel now in use for HEU targets is inadequate for the LEU target, since higher temperature and higher pressure will be encountered in the dissolution of LEU targets. The desire is to keep the modifications to the Cintichem process to a minimum, so that the switch from HEU to LEU can be achieved easily

  5. Analysis of the production of U3O8 powder for low enrichment fuel plates

    International Nuclear Information System (INIS)

    Boero, N.L.; Celora, J.; Parodi, C.A.; Ponieman, G.; Kellner, M.; Marajofsky, A.

    1987-01-01

    Description is made of the processes used in the production of U 3 O 8 powder for low enrichment plates for fuel elements for Research Reactors. The analysis of the efficiency of each batch is foccused on the relationship between milling and sieving times and the morphology of the product in each production step. (Author)

  6. SIMMER model of a low-enriched uranium non-power reactor

    International Nuclear Information System (INIS)

    Wilhelm, Dirk; Biaut, Guillaume; Tobita, Yoshiharu

    2008-01-01

    IRSN has started using the coupled neutronics-fluid dynamics code SIMMER [] to study core-disruptive accidents induced by insertions of large reactivities to produce very short period power excursions in fuel plate-type and water-moderated experimental research reactors. Until now, French safety analyses retain a bounding thermal energy released and mechanical yields, deduced from analysis of destructive in-pile test programs, to study the behavior of such reactors and design their structures and containment. Contrary to this approach, the present research program aims at modeling the design basis accident of research reactors with a low-enriched fuel using a CFD code. The objective is to analyze the effects of reactivity feedbacks and how they would limit the generated thermal energy released in the fuel. These aspects require a close coupling of the neutronics to the fluid dynamics analysis. The consequences of the nuclear power excursion, the changes of state of the fuel and the coolant, and ultimately the mechanical energy released are calculated by SIMMER. For large step-wise reactivity introductions, the Doppler effect and, at a lower extent, the fuel element thermal dilatation, which generates locally a decrease of the moderator to fuel ratio, limit the power excursion before the energy released is high enough to melt a large part of the fuel. Moreover, it has been shown that imposing an external reactivity as a step-wise or time-dependent reactivity introduction yields results quite different from those of the physical movement of control rods

  7. Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Peters, N. J. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Cowherd, W. M. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program; Rickman, B. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program

    2014-12-01

    This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo

  8. Development of industrial-scale fission {sup 99}Mo production process using low enriched uranium target

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Kon; Lee, Jun Sig [Radioisotope Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Beyer, Gerd J. [Grunicke Strasse 15, Leipzig (Germany)

    2016-06-15

    Molybdenum-99 ({sup 99}Mo) is the most important isotope because its daughter isotope, technetium-99m ({sup 99}mTc), has been the most widely used medical radioisotope for more than 50 years, accounting for > 80% of total nuclear diagnostics worldwide. In this review, radiochemical routes for the production of {sup 99}Mo, and the aspects for selecting a suitable process strategy are discussed from the historical viewpoint of {sup 99}Mo technology developments. Most of the industrial-scale {sup 99}Mo processes have been based on the fission of {sup 235}U. Recently, important issues have been raised for the conversion of fission {sup 99}Mo targets from highly enriched uranium to low enriched uranium (LEU). The development of new LEU targets with higher density was requested to compensate for the loss of {sup 99}Mo yield, caused by a significant reduction of {sup 235}U enrichment, from the conversion. As the dramatic increment of intermediate level liquid waste is also expected from the conversion, an effective strategy to reduce the waste generation from the fission {sup 99}Mo production is required. The mitigation of radioxenon emission from medical radioisotope production facilities is discussed in relation with the monitoring of nuclear explosions and comprehensive nuclear test ban. Lastly, the {sup 99}Mo production process paired with the Korea Atomic Energy Research Institute's own LEU target is proposed as one of the most suitable processes for the LEU target.

  9. Development of Industrial-Scale Fission 99Mo Production Process Using Low Enriched Uranium Target

    Directory of Open Access Journals (Sweden)

    Seung-Kon Lee

    2016-06-01

    Full Text Available Molybdenum-99 (99Mo is the most important isotope because its daughter isotope, technetium-99m (99mTc, has been the most widely used medical radioisotope for more than 50 years, accounting for > 80% of total nuclear diagnostics worldwide. In this review, radiochemical routes for the production of 99Mo, and the aspects for selecting a suitable process strategy are discussed from the historical viewpoint of 99Mo technology developments. Most of the industrial-scale 99Mo processes have been based on the fission of 235U. Recently, important issues have been raised for the conversion of fission 99Mo targets from highly enriched uranium to low enriched uranium (LEU. The development of new LEU targets with higher density was requested to compensate for the loss of 99Mo yield, caused by a significant reduction of 235U enrichment, from the conversion. As the dramatic increment of intermediate level liquid waste is also expected from the conversion, an effective strategy to reduce the waste generation from the fission 99Mo production is required. The mitigation of radioxenon emission from medical radioisotope production facilities is discussed in relation with the monitoring of nuclear explosions and comprehensive nuclear test ban. Lastly, the 99Mo production process paired with the Korea Atomic Energy Research Institute's own LEU target is proposed as one of the most suitable processes for the LEU target.

  10. Conversion and Blending Facility highly enriched uranium to low enriched uranium as metal. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-05

    The mission of this Conversion and Blending Facility (CBF) will be to blend surplus HEU metal and alloy with depleted uranium metal to produce an LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal.

  11. Conversion and Blending Facility highly enriched uranium to low enriched uranium as metal. Revision 1

    International Nuclear Information System (INIS)

    1995-01-01

    The mission of this Conversion and Blending Facility (CBF) will be to blend surplus HEU metal and alloy with depleted uranium metal to produce an LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal

  12. Simmer model of a low-enriched uranium non-power reactor

    International Nuclear Information System (INIS)

    Wilhelm, Dirk; Biaut, Guillaume; Tobita, Yoshiharu

    2006-01-01

    IRSN has started to use the coupled neutronics - fluid dynamics code SIMMER to study core-disruptive accident induced by insertions of large reactivities sufficient to very short period power excursions in fuel plate-type and water-moderated experimental research reactors. Until now, French safety analysis retain thermal energy released and mechanical yields, deduced from analysis of destructive test programs SPERT-I and BORAX-I to demonstrate the behavior of such reactors and design their structures and containment. The present research program models the design basis accident of a low enriched fuel currently used in experimental research reactors contrary to SPERT-I or BORAX-I. The objective is to analyze the effects of counter reactivities and how these would limit the generated thermal energy in the fuel. This part demands a close coupling to the fluid dynamics analysis. The consequences of the nuclear power excursion, the changes of state of the fuel and the coolant, and ultimately the mechanical energy released are calculated by SIMMER. For large step-wise reactivity introductions, the Doppler effect limits the power excursion before energy is released high enough to melt a large part of the fuel. Moreover, it has been shown that imposing an external reactivity as a step-wise or time dependant reactivity introduction yields results quite different from those of the physical movement of control rods. (author)

  13. Loading and initial start-up testing of the low-enrichment uranium core for the Ohio State University research reactor

    International Nuclear Information System (INIS)

    Talnagi, J.W.

    1989-01-01

    Conversion of the Ohio State University Research Reactor (OSURR) from high-enrichment uranium (HEU) fuel to low-enrichment uranium (LEU) fuel elements was begun in August 1985, with funding provided by the U.S. Department of Energy (DOE) and the university. Conversion of the OSURR from HEU to LEU fuel was successfully completed. The reactor is operational at 10-kW steady-state thermal power. Measurements of selected core parameters have been made and compared with predicted values and previous values for the HEU core. In general, measured results agree well with predicted performance, and minor changes have been detected in certain core parameters as a result of the change to LEU fuel. Future plans include additional core testing and a possible increase in operating power

  14. Moderation control in low enriched 235U uranium hexafluoride packaging operations and transportation

    International Nuclear Information System (INIS)

    Dyer, R.H.; Kovac, F.M.; Pryor, W.A.

    1993-01-01

    Moderation control is the basic parameter for ensuring nuclear criticality safety during the packaging and transport of low 235 U enriched uranium hexafluoride before its conversion to nuclear power reactor fuel. Moderation control has permitted the shipment of bulk quantities in large cylinders instead of in many smaller cylinders and, therefore, has resulted in economies without compromising safety. Overall safety and uranium accountability have been enhanced through the use of the moderation control. This paper discusses moderation control and the operating procedures to ensure that moderation control is maintained during packaging operations and transportation

  15. TYPE AF CERTIFICATE FOR TRANSPORTATION OF LOW ENRICHED URANIUM OXIDE (LEUO) FOR DISPOSAL

    International Nuclear Information System (INIS)

    Opperman, E; Kenneth Yates, K

    2007-01-01

    Washington Savannah River Company (WSRC) operates the Savannah River Site (SRS) in Aiken, SC under contract with the U.S. Department of Energy (DOE). SRS had the need to ship 227 drums of low enriched uranium oxide (LEUO) to a disposal site. The LEUO had been packaged nearly 25 years ago in U.S. Department of Transportation (DOT) 17C 55-gallon drums and stored in a warehouse. Since the 235U enrichment was just above 1 percent by weight (wt%) the material did not qualify for the fissile material exceptions in 49 CFR 173.453, and therefore was categorized as 'fissile material' for shipping purposes. WSRC evaluated all existing Type AF packages and did not identify any feasible packaging. Applying for a new Type AF certificate of compliance was considered too costly for a one-time/one-way shipment for disposal. Down-blending the material with depleted uranium (to reduce enrichment below 1 wt% and enable shipment as low specific activity (LSA) radioactive material) was considered, but appropriate blending facilities do not exist at SRS. After reviewing all options, WSRC concluded that seeking a DOT Special Permit was the best option to enable shipment of the material for permanent disposal. WSRC submitted the Special Permit application to the DOT, and after one request-for-additional-information (RAI) the permit was considered acceptable. However, in an interesting development that resulted from the DOT Special Permit application process, it was determined that it was more appropriate for the DOE to issue a Type AF certificate [Ref. 1] for this shipping campaign. This paper will outline the DOT Special Permit application and Type AF considerations, and will discuss the issuance of the new DOE Type AF certificate of compliance

  16. The low enriched fuel cycle in the GA 1160 MW design and the switch-over to thorium

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, H.

    1974-03-15

    Calculations for the GA 1160 MW HTR are presented. The aim of these investigations was to compare the Low Enriched Uranium (LEU) cycle and the Thorium cycle for the GA 1160 MW HTR both using the same GA designed integral block fuel element. The total fuel cycle cost for the equilibrium cycle comes out to be about 16% cheaper for the Thorium cycle than for the Low-Enriched cycle. However, these favorable results for the thorium cycle are completely dependent on the availability of reprocessing and refabrication facilities, for costs comparable with the costs used for these investigations. The possibility of starting the reactor on a LEU 3 year cycle and later switching over to a thorium 4 year cycle was investigated. No cost penalties were found to be paid during the switch-over. The problems of local power peaks and age factors were not investigated in greater detail as only integral physical quantities were obtained from the neutron physics calculations. However, no indications of any problem in the switch-over phase were given. Elaborate 3-dimensional methods are necessary for further investigation of these types of problems.

  17. Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jaluvka, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States); Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States); McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States); Peters, N. J. [Univ. of Missouri, Columbia, MO (United States)

    2017-02-01

    This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization (M3).

  18. Low-enriched research reactor fuel: Post-Irradiation Examinations at SCK-CEN

    International Nuclear Information System (INIS)

    Van den Berghe, S.; Leenaers, A.

    2007-01-01

    Generally, research and test reactors are fuelled with fuel plates instead of pins. In most cases in the past, these plates consisted of high enriched (higher than 95 percent 235 U) UAl 3 powder mixed with a pure Al matrix (called the meat) in between two aluminium alloy plates (the cladding). These plates are then assembled in fuel elements of different designs to fit the needs of the various reactors. Since the 1970's, efforts have been going on to replace the high-enriched, low-density UAl 3 fuel with high-density, low enriched ( 235 U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched materials because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative and the Reduced Enrichment for Research and Test Reactors program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has been obtained with U 3 Si 2 fuel, which is currently used in many research reactors in the world. However, efforts to search for a better replacement have continued and are currently directed towards the U-Mo alloy fuel (7-10 weight percent Mo)

  19. Examinations of the irradiation behaviour of U3Si2 test fuel plates with low enrichment

    International Nuclear Information System (INIS)

    Muellauer, J.

    1989-01-01

    Five low-enriched (19.7% 235 U), high-density (4.7 gU/cm/ 3 ) U 3 Si 2 -test fuel plates (miniplates) with different fine grain contents have been qualified under irradiation. During the course of irradiation up to burnup of 63% 235 U depletion, no released fractions of gaseous or solid fission products from the fuel plate to the rig coolant were detected. The measured swelling rate of the fuel zone (meat) is less than 0.45% ΔV/10 20 fissions/cm 3 the blister-threshold temperature of the fuel plates is above 520 0 C. The favourable irradiation behavior of the U 3 Si 2 fuel plates was not influenced by using higher amounts of fine grained particles (40% [de

  20. Feasibility studies to establish at the Kazakhstan Ulba metallurgical plant the manufacturing capability to produce low-enriched uranium certified reference materials

    Energy Technology Data Exchange (ETDEWEB)

    Kuzminski, Jozef [Los Alamos National Laboratory; Nesuhoff, J [NBL; Cratto, P [NBL; Pfennigwerth, G [Y12 NATIONAL SEC. COMPLEX; Mikhailenko, A [ULBA METALLURGICAL PLANT; Maliutina, I [ULBA METALLURGICAL PLANT; Nations, J [GREGG PROTECTION SERVICES

    2009-01-01

    One of the salient features of the transition plan that the United States Department of Energy/National Nuclear Security Administration (DOE/NNSA) is presently implementing in the Former Soviet Union countries is the availability of uranium certified reference materials for calibration of nondestructive assay (NDA) measurement equipment. To address this challenge, DOE/NNSA and U.S. national laboratories have focused their cooperative efforts on establishing a reliable source for manufacturing, certifying, and supplying of such standards. The Ulba Metallurgical Plant (UMP), Kazakhstan, which processes large quantities of low-enriched uranium to produce ceramic fuel pellets for nuclear-powered reactors, is well situated to become a key supplier of low-enriched uranium certified reference materials for the country and Central Asia region. We have recently completed Phase I of a feasibility study to establish at UMP capabilities of manufacturing these standards. In this paper we will discuss details of a proposed methodology for uranium down-blending, material selection and characterization, and a proposed methodology of measurement by destructive (DA) and non-destructive (NDA) analysis to form a database for material certification by the competent State authorities in the Republic of Kazakhstan. In addition, we will discuss the prospect for manufacturing of such standards at UMP.

  1. 2nd RCM of the CRP on Analytical and Experimental Benchmark Analyses of Accelerator Driven Systems (ADS) and Technical Meeting on Low Enriched Uranium (LEU) Fuel Utilization in Accelerator Driven Sub-critical Systems. Working Material

    International Nuclear Information System (INIS)

    2010-01-01

    The overall objective of the CRP is contributing to the generic R&D efforts in various fields common to innovative fast neutron system development, i.e., heavy liquid metal thermal hydraulics, dedicated transmutation fuels and associated core designs, theoretical nuclear reaction models, measurement and evaluation of nuclear data for transmutation, and development and validation of calculational methods and codes. Ultimately, the CRP’s overall objective is to make contributions towards the realization of a transmutation demonstration facility

  2. A programme for Euratom safeguards inspectors, used in the assay of high enriched (H.E.U.) and low enriched (L.E.U.) uranium fuel materials by active neutron interrogation

    International Nuclear Information System (INIS)

    Vocino, V.; Farese, N.; Maucq, T.; Nebuloni, M.

    1991-01-01

    The programme AECC (Active Euratom Coincidence Counters) has been developed at the Joint Research Center, Ispra by the Euratom Safeguards Directorate, Luxembourg and the Safety Technology Institute, Ispra for the acquisition, evaluation, management and storage of measurement data originating from active neutron interrogation of HEU and LEU fuel materials. The software accommodates the implementation of the NDA (Non Destructive Assay) procedures for the Active Well Coincidence Counters and Active Neutron Coincidence Counters deployed by the Euratom Safeguards Directorate, Luxembourg

  3. An assessment of the radiological consequences of the Greek Research Reactor's design basis accident with the use of low enriched uranium

    International Nuclear Information System (INIS)

    Kollas, J.G.

    1985-09-01

    An analysis of the radiological consequences of the design basis accident in the low enriched uranium fueled 5 MW Greek Research Reactor is presented. For the source term thirty-five isotopes are taken into consideration and conservative figures of fission product release are adopted. To estimate the reactor's consequences for Athens population a CRAC2 consequence model version is used. The results indicate that limiting dose and effects are respectively the thyroid dose and the thyroid effects induced in the 3,081,000 inhabitants of Athens region. (author)

  4. U.S. progress in the development of very high density low enrichment research reactor fuels

    International Nuclear Information System (INIS)

    Meyer, M. K.; Wachs, D. M.; Jue, J.-F.; Keiser, D. D.; Gan, J.; Rice, F.; Robinson, A.; Woolstenhulme, N. E.; Medvedev, P.; Hofman, G. L.; Kim, Y.-S.

    2012-01-01

    The effort to develop low-enriched fuels for high power research reactors began world-wide in 1996. Since that time, hundreds of fuel specimens have been tested to investigate the operational limits of many variations of U-Mo alloy dispersion and monolithic fuels. In the U.S., the fuel development program has focused on the development of monolithic fuel, and is currently transitioning from conducting research experiments to the demonstration of large scale, prototypic element assemblies. These larger scale, integral fuel performance demonstrations include the AFIP-7 test of full-sized, curved plates configured as an element, the RERTR-FE irradiation of hybrid fuel elements in the Advanced Test Reactor, reactor specific Design Demonstration Experiments, and a multi-element Base Fuel Demonstration. These tests are conducted alongside mini-plate tests designed to prove fuel stability over a wide range of operating conditions. Along with irradiation testing, work on collecting data on fuel plate mechanical integrity, thermal conductivity, fission product release, and microstructural stability is underway. (authors)

  5. Environmental assessment: Transfer of normal and low-enriched uranium billets to the United Kingdom, Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1995-11-01

    Under the auspices of an agreement between the U.S. and the United Kingdom, the U.S. Department of Energy (DOE) has an opportunity to transfer approximately 710,000 kilograms (1,562,000 pounds) of unneeded normal and low-enriched uranium (LEU) to the United Kingdom; thus, reducing long-term surveillance and maintenance burdens at the Hanford Site. The material, in the form of billets, is controlled by DOE's Defense Programs, and is presently stored as surplus material in the 300 Area of the Hanford Site. The United Kingdom has expressed a need for the billets. The surplus uranium billets are currently stored in wooden shipping containers in secured facilities in the 300 Area at the Hanford Site (the 303-B and 303-G storage facilities). There are 482 billets at an enrichment level (based on uranium-235 content) of 0.71 weight-percent. This enrichment level is normal uranium; that is, uranium having 0.711 as the percentage by weight of uranium-235 as occurring in nature. There are 3,242 billets at an enrichment level of 0.95 weight-percent (i.e., low-enriched uranium). This inventory represents a total of approximately 532 curies. The facilities are routinely monitored. The dose rate on contact of a uranium billet is approximately 8 millirem per hour. The dose rate on contact of a wooden shipping container containing 4 billets is approximately 4 millirem per hour. The dose rate at the exterior of the storage facilities is indistinguishable from background levels

  6. Non-destructive evaluation methods to improve quality control in low enrichment MTR fuel plate production

    International Nuclear Information System (INIS)

    Milne, J.M.; Lidington, B.; Hawker, B.M.

    1991-01-01

    This paper summarises some preliminary non-destructive measurements made recently at the Harwell Laboratory on a prototype low enrichment MTR fuel plate. The measurements were intended to indicate the potential of two different techniques for improving quality control in plate production. Pulse Video Thermography (PVT) is being considered as an alternative to ultrasound transmission measurements for the detection and sizing of lack of thermal bonding between the fuel and the clad layers, either to verify the indications from the established ultrasonic methods before destroying the plate or as a replacement method of inspection. High frequency pulse-echo ultrasonics is being considered for providing maps of clad layer thickness on each side of the plate. The measurements have indicated the potential for both methods, but more work is required, using a test plate containing controlled defects, to establish their capability. (orig.)

  7. The Impact of Climatological Conditions on Low Enriched Uranium Loading Station Operations for the HEU Blend Down Project

    International Nuclear Information System (INIS)

    Chang, R.C.

    2002-01-01

    A computer model was developed using COREsim to perform a time motion study for the Low Enriched Uranium (LEU) Loading Station operations. The project is to blend Highly Enriched Uranium (HEU) with Natural Uranium (NU) to produce LEU to be shipped to Tennessee Valley Authority (TVA) for further processing. To cope with a project cost reduction, the LEU Loading Station concept has changed from an enclosed building with air-conditioning to a partially enclosed building without air conditioning. The LEU Loading Station is within a radiological contaminated area; two pairs of coveralls and negative pressure respirator are required. As a result, inclement weather conditions, especially heat stress, will affect and impact the LEU loading operations. The purposes of the study are to determine the climatological impacts on LEU Loading operations, resources required for committed throughputs, and to find out the optimum process pathways for multi crews working simultaneously in the space-lim ited LEU Loading Station

  8. Progress in development of low-enriched U-Mo dispersion fuels

    International Nuclear Information System (INIS)

    Hofman, G.L.; Snelgrove, J.L.; Hayes, S.L.; Meyer, M.K.

    2002-01-01

    Results from post irradiation examinations and analyses of U-Mo/Al dispersion mini plates are presented. Irradiation test RERTR-5 contained mini- fuel plates with fuel loadings of 6 and 8 g U cm -3 . The fuel material consisted of 6, 7 and 10 wt. % Mo-uranium-alloy powders in atomized and machined form. The swelling behavior of the various fuel types is analyzed, indicating athermal swelling of the U-Mo alloy and temperature-dependent swelling owing to U-Mo/Al interdiffusion. (author)

  9. Criticality experiments with low enriched UO2 fuel rods in water containing dissolved gadolinium

    International Nuclear Information System (INIS)

    Bierman, S.R.; Murphy, E.S.; Clayton, E.D.; Keay, R.T.

    1984-02-01

    The results obtained in a criticality experiments program performed for British Nuclear Fuels, Ltd. (BNFL) under contract with the United States Department of Energy (USDOE) are presented in this report along with a complete description of the experiments. The experiments involved low enriched UO 2 and PuO 2 -UO 2 fuel rods in water containing dissolved gadolinium, and are in direct support of BNFL plans to use soluble compounds of the neutron poison gadolinium as a primary criticality safeguard in the reprocessing of low enriched nuclear fuels. The experiments were designed primarily to provide data for validating a calculation method being developed for BNFL design and safety assessments, and to obtain data for the use of gadolinium as a neutron poison in nuclear chemical plant operations - particularly fuel dissolution. The experiments program covers a wide range of neutron moderation (near optimum to very under-moderated) and a wide range of gadolinium concentration (zero to about 2.5 g Gd/l). The measurements provide critical and subcritical k/sub eff/ data (1 greater than or equal to k/sub eff/ greater than or equal to 0.87) on fuel-water assemblies of UO 2 rods at two enrichments (2.35 wt % and 4.31 wt % 235 U) and on mixed fuel-water assemblies of UO 2 and PuO 2 -UO 2 rods containing 4.31 wt % 235 U and 2 wt % PuO 2 in natural UO 2 respectively. Critical size of the lattices was determined with water containing no gadolinium and with water containing dissolved gadolinium nitrate. Pulsed neutron source measurements were performed to determine subcritical k/sub eff/ values as additional amounts of gadolinium were successively dissolved in the water of each critical assembly. Fission rate measurements in 235 U using solid state track recorders were made in each of the three unpoisoned critical assemblies, and in the near-optimum moderated and the close-packed poisoned assemblies of this fuel

  10. Environmental assessment for the purchase of Russian low enriched uranium derived from the dismantlement of nuclear weapons in the countries of the former Soviet Union

    International Nuclear Information System (INIS)

    1994-01-01

    The United States is proposing to purchase from the Russian Federation low enriched uranium (LEU) derived from highly enriched uranium (HEU) resulting from the dismantlement of nuclear weapons in the countries of the former Soviet Union. The purchase would be accomplished through a proposed contract requiring the United States to purchase 15,250 metric tons (tonnes) of LEU (or 22,550 tonnes of UF 6 ) derived from blending 500 metric tones uranium (MTU) of HEU from nuclear warheads. The LEU would be in the form of uranium hexafluoride (UF 6 ) and would be converted from HEU in Russia. The United States Enrichment Corporation (USEC) is the entity proposing to undertake the contract for purchase, sale, and delivery of the LEU from the Russian Federation. The US Department of Energy (DOE) is negotiating the procedure for gaining confidence that the LEU is derived from HEU that is derived from dismantled nuclear weapons (referred to as ''transparency),'' and would administer the transparency measures for the contract. There are six environments that could potentially be affected by the proposed action; marine (ocean); US ports of entry; truck or rail transportation corridors; the Portsmouth GDP; the electric power industry; and the nuclear fuel cycle industry. These environmental impacts are discussed

  11. Environmental assessment for the purchase of Russian low enriched uranium derived from the dismantlement of nuclear weapons in the countries of the former Soviet Union

    Energy Technology Data Exchange (ETDEWEB)

    1994-01-01

    The United States is proposing to purchase from the Russian Federation low enriched uranium (LEU) derived from highly enriched uranium (HEU) resulting from the dismantlement of nuclear weapons in the countries of the former Soviet Union. The purchase would be accomplished through a proposed contract requiring the United States to purchase 15,250 metric tons (tonnes) of LEU (or 22,550 tonnes of UF{sub 6}) derived from blending 500 metric tones uranium (MTU) of HEU from nuclear warheads. The LEU would be in the form of uranium hexafluoride (UF{sub 6}) and would be converted from HEU in Russia. The United States Enrichment Corporation (USEC) is the entity proposing to undertake the contract for purchase, sale, and delivery of the LEU from the Russian Federation. The US Department of Energy (DOE) is negotiating the procedure for gaining confidence that the LEU is derived from HEU that is derived from dismantled nuclear weapons (referred to as ``transparency),`` and would administer the transparency measures for the contract. There are six environments that could potentially be affected by the proposed action; marine (ocean); US ports of entry; truck or rail transportation corridors; the Portsmouth GDP; the electric power industry; and the nuclear fuel cycle industry. These environmental impacts are discussed.

  12. Criticality of moderated and undermoderated low-enriched uranium oxide systems

    International Nuclear Information System (INIS)

    Goebel, G.R.

    1980-06-01

    Uranium oxide was enriched to 4.46 wt % 235 U compacted to a density of 4.68 g/cm 3 . The uranium oxide was packed into cubical aluminum cans and water added to the oxide until an H/U atomic ratio of 0.77 was achieved. A 5 x 5 x 5 array of uranium oxide cans for the experiments were used when no plastic moderator material was placed between cans. High enriched uranium drivers were used to achieve criticality. Criticality was achieved for smaller arrays without a driver when 24.5 mm plastic moderator material was placed between the cans. Twelve critical experiments are reported, six in each reflector

  13. Benchmark critical experiments on low-enriched uranium oxide systems with H/U = 0.77

    International Nuclear Information System (INIS)

    Tuck, G.; Oh, I.

    1979-08-01

    Ten benchmark experiments were performed at the Critical Mass Laboratory at Rockwell International's Rocky Flats Plant, Golden, Colorado, for the US Nuclear Regulatory Commission. They provide accurate criticality data for low-enriched damp uranium oxide (U 3 O 8 ) systems. The core studied consisted of 152 mm cubical aluminum cans containing an average of 15,129 g of low-enriched (4.46% 235 U) uranium oxide compacted to a density of 4.68 g/cm 3 and with an H/U atomic ratio of 0.77. One hundred twenty five (125) of these cans were arranged in an approx. 770 mm cubical array. Since the oxide alone cannot be made critical in an array of this size, an enriched (approx. 93% 235 U) metal or solution driver was used to achieve criticality. Measurements are reported for systems having the least practical reflection and for systems reflected by approx. 254-mm-thick concrete or plastic. Under the three reflection conditions, the mass of the uranium metal driver ranged from 29.87 kg to 33.54 kg for an oxide core of 1864.6 kg. For an oxide core of 1824.9 kg, the weight of the high concentration (351.2 kg U/m 3 ) solution driver varied from 14.07 kg to 16.14 kg, and the weight of the low concentration (86.4 kg U/m 3 ) solution driver from 12.4 kg to 14.0 kg

  14. Metallurgical and reactor physics aspects of using low enrichment fuel in Safari-I

    International Nuclear Information System (INIS)

    1978-09-01

    The feasibility of using lower than 93% enriched fuel in the SAFARI-I research and materials testing reactor is reviewed. Metallurgical experiments show that, using standard U-Al alloy technology and keeping the 235 U loading per element constant without altering the fuel plate thickness, a maximum of 35 weight percent of uranium in the meat can be achieved. This corresponds to using a minimum enrichment of 40% 235 U in order to retain the same mass of 235 U in the core. Even then a loss of approximately 3,3% in reactivity is calculated, which is more than the 2,8% sup(deltak)/k which is normally allowed for burnup. Using current U-Al alloy fuel technology, and an enrichment of approximately 45% 235 U, no changes in core configuration or coolant requirements will be necessary. The use of 20% enriched uranium will require the development of a new fuel design and technology if drastic redesign and modification of the reactor and coolant circuits is to be avoided. Without such new technology, the redesign and modification of the reactor will cost upwards of 3 million dollars and take up to 5 years to complete, requiring a complete shutdown of the reactor for approximately 2 years

  15. Conversion and Blending Facility highly enriched uranium to low enriched uranium as uranyl nitrate hexahydrate. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-05

    This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials to pure HEU uranyl nitrate (UNH) and (2) blend pure HEU UNH with depleted and natural UNH to produce HEU UNH crystals. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU Will be produced as a waste suitable for storage or disposal.

  16. Advances of the low enriched uranium utilization project in CNA-1 during 1998 and 1999

    International Nuclear Information System (INIS)

    Fink, Jose M.; Higa, Manabu; Sidelnik, Jorge I.; Perez, Ramon A.; Casario, Jose A.; Alvarez, Luis A.

    1999-01-01

    In this work, a general description of advances of the Enriched Fuel Introduction Project in CNA-1 and the main tasks performed during 1998 and 1999 are presented. The program is being satisfactorily developed and during that period the number of slightly enriched fuels (LEU) introduced had significantly increased in relation to previous years. At present, there are 181 LEU fuel elements in the core and 125 LEU fuel elements have been extracted. The number of full power burnt fuel elements per day decreased from 1.31 FE/dpp in 1994 (when all fuel was natural) to 0.92 in 1998 and 0.83 in 1999, reaching the predicted value for homogeneous LEU core of 0.7. The cost of burnt fuel in 1998 was 25% lower that if only natural fuel would have been used. (author)

  17. Low enriched uranium foil targets with different geometries for the production of Molybdenum-99 in the BMR (Brazilian Multipurpose Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Muniz, Rafael O.R.; Coelho, Talita S., E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    A new research reactor is being planned in Brazil to take care of the demand of radiopharmaceuticals in the country and conduct research in various areas. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Two low enriched (<20% {sup 235}U) metallic uranium foil targets (cylinder and plate geometries) are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of {sup 99}Mo for these targets in the RMB and to determine the temperatures achieved in the targets. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations were utilized the computer codes MTRCR-IEA-R1 and ANSYS CFX. (author)

  18. Fifth Supply Agreement. Agreement for the Transfer of Low Enriched Uranium for a Research Reactor in Romania

    International Nuclear Information System (INIS)

    2008-01-01

    The text of the Fifth Supply Agreement among the Government of Romania, the Government of the United States of America and the International Atomic Energy Agency for the Transfer of Low Enriched Uranium for a Research Reactor in Romania is reproduced in this document for the information of all Members of the Agency. The Agency's Board of Governors approved the text of the Agreement on 20 November 2003, which was signed by the authorized representatives of Romania and the United States, and by the Director General of the IAEA, on 24 November 2003. Pursuant to Article V of the Agreement, the Agreement entered into force on 24 November 2003, upon signature by the representatives of Romania, the United States and the Director General of the IAEA

  19. Low enriched uranium foil targets with different geometries for the production of Molybdenum-99 in the BMR (Brazilian Multipurpose Reactor)

    International Nuclear Information System (INIS)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Muniz, Rafael O.R.; Coelho, Talita S.

    2011-01-01

    A new research reactor is being planned in Brazil to take care of the demand of radiopharmaceuticals in the country and conduct research in various areas. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Two low enriched ( 235 U) metallic uranium foil targets (cylinder and plate geometries) are being considered for production of Molybdenum-99 ( 99 Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of 99 Mo for these targets in the RMB and to determine the temperatures achieved in the targets. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations were utilized the computer codes MTRCR-IEA-R1 and ANSYS CFX. (author)

  20. Critical experiments on low-enriched uranium oxide system with H/U=1.25

    International Nuclear Information System (INIS)

    Oh, I.; Rothe, R.E.; Tuck, G.

    1982-01-01

    Fifteen (15) critical experiments were performed on a horizontal split table machine using 4.48%-enriched sup(235)U uranium oxide(U 3 O 8 ). The oxide was compacted to a density of 4.68g/cm 3 and placed in 152 mm cubical aluminum cans. Water was added to achive an H/U of 1.25. Various arrays of oxide cans were distributed on each half of the split table, and the separation between halves reduced until criticality occurred. The critical table separation varied from 3.59 mm to 18.40 mm. Twelve (12) experiments required the addition of a high-enriched(-93 %sup(235)U) metal or solution driver to achieve criticality. These experiments were performed in a plastic, concrete, or thin steel reflector. Three additional experiments in the plastic reflector contained either 9.3-mm- or 24.3-mm-thick plastic moderator material between the oxide cans and did not require a driver to achieve criticality. Critical uranium driver masses ranged from 9.999 kg to 14.000 kg (solution driver), and from 25.378 kg to 29.278 kg (metal driver) for 5X5X5 arrays of uranium oxide cans. Always, one or four of these 125 cans had to be removed to make room for the drivers. Therefore, the uranium oxide masses used were 1823.8 kg and 1863.5 kg. For the moderated experiments, the uranium oxide mass ranged between 574.4 kg and 1210.0 kg. (Author)

  1. Conversion to low-enriched fuel in research reactor aspects of licensing the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Jacquemin, J.

    1985-01-01

    Conversion to low-enriched fuel and usage of new developed highly densified fuel in research-reactors will be an essential alteration in operating the reactor. According to the German Energy Act this has to be licensed. here might be some risk to the licensee of an older research-reactor by suspending his operating license because he cannot meet current requirements to be fulfilled or because of a court decision.Disposal of irradiated fuel elements of the new fuel type is a further significant problem which has to be solved before issuing a new license. (author)

  2. Converting targets and processes for fission-product molybdenum-99 from high- to low-enriched uranium

    International Nuclear Information System (INIS)

    Vandegrift, G.F.; Snelgrove, J.L.; Aase, S.

    1999-01-01

    Most of the world's supply of 99 Mo is produced by the fissioning of 235 U in high-enriched uranium targets (HEU, generally 93% 235 U). To reduce nuclear-proliferation concerns, the U.S. Reduced Enrichment for Research and Test Reactor Program is working to convert the current HEU targets to low-enriched uranium (LEU, 235 U). Switching to LEU targets also requires modifying the separation processes. Current HEU processes can be classified into two main groups based on whether the irradiated target is dissolved in acid or base. Our program has been working on both fronts, with development of targets for acid-side processes being the furthest along. However, using an LEU metal foil target may allow the facile replacement of HEU for both acid and basic dissolution processes. Demonstration of the irradiation and 99 Mo separation processes for the LEU metal-foil targets is being done in cooperation with researchers at the Indonesian PUSPIPTEK facility. We are also developing LEU UO 2 /Al dispersion plates as substitutes for HEU UA1 x /A1 dispersion plates for base-side processes. Results show that conversion to LEU is technically feasible; working with producers is essential to lowering any economic penalty associated with conversion. (author)

  3. The Supply of Medical Radioisotopes. Market impacts of converting to low-enriched uranium targets for medical isotope production

    International Nuclear Information System (INIS)

    Westmacott, Chad; Cameron, Ron

    2012-01-01

    The reliable supply of molybdenum-99 ( 99 Mo) and its decay product, technetium-99m ( 99m Tc), is a vital component of modern medical diagnostic practices. At present, most of the global production of 99 Mo is from highly enriched uranium (HEU) targets. However, all major 99 Mo-producing countries have recently agreed to convert to using low-enriched uranium (LEU) targets to advance important non-proliferation goals, a decision that will have implications for the global supply chain of 99 Mo/ 99m Tc and the long-term supply reliability of these medical isotopes. This study provides the findings and analysis from an extensive examination of the 99 Mo/ 99m Tc supply chain by the OECD/NEA High-level Group on the Security of Supply of Medical Radioisotopes (HLG-MR). It presents a comprehensive evaluation of the potential impacts of converting to the use of LEU targets for 99 Mo production on the global 99 Mo/ 99m Tc market in terms of costs and available production capacity, and the corresponding implications for long-term supply reliability. In this context, the study also briefly discusses the need for policy action by governments in their efforts to ensure a stable and secure long-term supply of 99 Mo/ 99m Tc

  4. Postirradiation examination of a low enriched U3Si2-Al fuel element manufactured and irradiated at Batan, Indonesia

    International Nuclear Information System (INIS)

    Suripto, A.; Sugondo, S.; Nasution, H.

    1994-01-01

    The first low-enriched U 3 Si 2 -Al dispersion plate-type fuel element produced at the Nuclear Fuel Element Center, BATAN, Indonesia, was irradiated to a peak 235 U burnup of 62%. Postirradiation examinations performed to data shows the irradiation behavior of this element to be similar to that of U 3 Si 2 -Al plate-type fuel produced and tested at other institutions. The main effect of irradiation on the fuel plates is a thickness increase of 30--40 μm (2.5-3.0%). This thickness increase is almost entirely due to the formation of a corrosion layer (Boehmite). The contribution of fuel swelling to the thickness increase is rather small (less than 10 μm) commensurate with the burnup of the fuel and the relatively moderate as-fabricated fuel volume fraction of 27% in the fuel meat

  5. Comparison of control rod effectiveness for thorium and low-enriched fuel cycles in the GA-1, 160 MW(e) design

    Energy Technology Data Exchange (ETDEWEB)

    Neef, Hans Joachim

    1974-03-15

    In an investigation of the properties of the Thorium-Uranium (Th) and the Low-Enriched Uranium (LEU) fuel cycles it is also necessary to compare the effectiveness of the control rods in a reactor system operating with these sorts of fuel. Furthermore, it is under consideration to start a reactor with LEU fuel and switch-over to a Th cycle. It is also of interest to look at the switch-over phase in respect to the control rod effectiveness. The various fuel cycles have been studied for the same fuel element and control rod design, namely the one of GA's commercially available 1,160 MW(e) reference power station. This paper gives the first results on the control rod calculations and is presented mainly in two parts. Part 1 describes spectral effects which have been investigated by cell calculations with a discrete ordinates transport code. The main result is the higher effectiveness of a rod in a Th-cycle compared with a LEU-cycle. Part 2 reports on reactor calculations with a diffusion code and shows that this advantage can partially disappear in the reactor because of the spatial flux distribution. This effect has to be studied in further investigations for a full understanding.

  6. Validation of SCALE 4.0 -- CSAS25 module and the 27-group ENDF/B-IV cross-section library for low-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, W.C.

    1993-02-01

    A version of KENO V.a and the 27-group library in SCALE-4.0 were validated for use in evaluating the nuclear criticality safety of low-enriched uranium systems. A total of 59 critical systems were analyzed. A statistical analysis of the results was performed, and subcritical acceptanced criteria are established.

  7. Validation of SCALE 4. 0 -- CSAS25 module and the 27-group ENDF/B-IV cross-section library for low-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, W.C.

    1993-02-01

    A version of KENO V.a and the 27-group library in SCALE-4.0 were validated for use in evaluating the nuclear criticality safety of low-enriched uranium systems. A total of 59 critical systems were analyzed. A statistical analysis of the results was performed, and subcritical acceptanced criteria are established.

  8. Neutronic design of a LEU [low enriched uranium] core for the Ohio State University research reactor

    International Nuclear Information System (INIS)

    Seshadri, M.D.; Aybar, H.S.; Aldemir, T.

    1987-01-01

    The 10 kw HEU fuelled Ohio State University Reactor (OSURR) will be upgraded to operate at 500 kW with standardized 125 g 235 U LEU U 3 Si 2 fuel plates. An earlier scoping study based on two-dimensional diffusion calculations has identified the potential LEU core configurations for the conversion/upgrade of OSURR using the standardized plates in a 16-plate (+ 2 dummy plates) standard and 10-scoping study is improved for a more precise determination of the excess reactivities and safety rod worths for these potential configurations. Comparison of the results obtained by the improved model to experimental results and to the results of full-core Monte Carlo simulations shows excellent agreement. The results also indicate that the conversion/upgrade of OSURR can be realized with three possible LEU core configurations while maintaining a cold, clean shutdown margin of 1.57-1.91 % Δ k/k, depending on the configuration used. (Author)

  9. Remote Handling Devices for Disposition of Enriched Uranium Reactor Fuel Using Melt-Dilute Process

    International Nuclear Information System (INIS)

    Heckendorn, F.M.

    2001-01-01

    Remote handling equipment is required to achieve the processing of highly radioactive, post reactor, fuel for the melt-dilute process, which will convert high enrichment uranium fuel elements into lower enrichment forms for subsequent disposal. The melt-dilute process combines highly radioactive enriched uranium fuel elements with deleted uranium and aluminum for inductive melting and inductive stirring steps that produce a stable aluminum/uranium ingot of low enrichment

  10. Radioactive Waste Issues related to Production of Fission-based Mo-99 by using Low Enriched Uranium (LEU)

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Muhmood ul; Ryu, Ho Jin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    In order to produce fission-based Mo-99 from research reactors, two types of targets are being used and they are highly enriched uranium (HEU) targets with {sup 235}U enrichment more than 90wt% of {sup 235}U and low enriched uranium (LEU) targets with {sup 235}U enrichment less than 20wt% of {sup 235}U. It is worth noting that medium enriched uranium i.e. 36wt% of {sup 235}U as being used in South Africa is also regarded as non-LEU from a nuclear security point of view. In order to cope with the proliferation issues, international nuclear security policy is promoting the use of LEU targets in order to minimize the civilian use of HEU. It is noteworthy that Mo-99 yield of the LEU target is less than 20% of the HEU target, which requires approximately five times more LEU targets to be irradiated and consequently results in increased volume of waste. The waste generated from fission Mo-99 production can be mainly due to: target fabrication, assembling of target, irradiation in reactor and processing of irradiated targets. During the fission of U-235 in a reactor, a large number of radionuclides with different chemical and physical properties are formed. The waste produced from these practices may be a combination of low level waste (LLW) and intermediate level waste (ILW) comprised of all three types, i.e., solid, liquid and gas. Handling and treatment of the generated waste are dependent on its form and activity. In case of the large production facility, waste storage facility should be constructed in order to limit the radiation exposures of the workers and the environment. In this study, we discuss and compare mainly the radioactive waste generated by alkaline digestion of both HEU and LEU targets to assist in planning and deciding the choice of the technology with better arrangements for proper handling and disposal of generated waste. With the use of the LEU targets in Mo-99 production facility, significant increase in liquid and solid waste has been expected.

  11. Preliminary Accident Analyses for Conversion of the Massachusetts Institute of Technology Reactor (MITR) from Highly Enriched to Low Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, Erik H. [Argonne National Lab. (ANL), Argonne, IL (United States); Sun, Kaichao S. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Newton, Jr., Thomas H. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2013-09-30

    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. This report presents the preliminary accident analyses for MITR cores fueled with LEU monolithic U-Mo alloy fuel with 10 wt% Mo. Preliminary results demonstrate adequate performance, including thermal margin to expected safety limits, for the LEU accident scenarios analyzed.

  12. Environmental Assessment for the shipment of low enriched uranium billets to the United Kingdom from the Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1992-08-01

    This Environmental Assessment provides the necessary information so that a decision can be made on whether a Finding of No Significant Impact Environmental Impact Statement should be prepared for the proposed action. The proposed action is to transfer 2,592 low enriched uranium billets to the United Kingdom. The billets are currently stored in the 300 Area of the Hanford Site, Richland, Washington. The proposed action would consist of two types of activities: loading and transportation. The loading activities would include placing the billets into the appropriate containers for transportation. The transportation activities would include the tasks required to transport the containers 215 miles (344 km) via highway to the Port of Seattle, Washington, and transfer the containers aboard an ocean cargo vessel for transportation to the United Kingdom. The Department of Energy would only be responsible for conducting the loading activities. The United Kingdom would be responsible for conducting the transportation activities in compliance with all applicable United States and international transportation laws. The tasks associated with the proposed action activities have been performed before and are well defined in terms of requirements and consequences. A risk assessment and a nuclear safety evaluation were performed to address safety issues associated with the proposed action. The risk assessment determined the exposure risk from normal operation and from the maximum credible accident that involves a truck or ship collision followed by a fire that engulfs all the billets in the shipment and the release of the radiological contents of the shipment to the environment. The criticality assessment determined the nuclear safety limits for handling, transporting and storing the shipment under incident-free and accident transport conditions

  13. A comparison of integral block and tubular interacting fuel element concepts for low enrichment HTR

    Energy Technology Data Exchange (ETDEWEB)

    Desoisa, J A

    1972-04-15

    The tubular interacting fuel element has to date been the favoured U.K. high temperature reactor design. Recent attempts to lower fuel costs and the progress of the Fort St. Vrain reactor has focussed attention on alternative designs, and in particular on the attractive design simplicity of the integral block concept. The aim of this investigation is to compare the merits of both concepts from fuel cycle cost and thermal performance viewpoints and to determine whether optimization of the integral block concept leads to changes in the current design values of (a) fuel density, (b) Nc/Nu, and/or (c) mean discharge irradiation within the framework of present design limits.

  14. Description of the CNEA U308 powder production plant for low enrichment fuel plates

    International Nuclear Information System (INIS)

    Boero, N.L.; Celora, J.; Parodi, C.A.; Pertossi, F.R.; Marajofsky, A.

    1987-01-01

    The design of the 20% enriched U 3 O 8 powder production plant was based on laboratory level experiments. The UF 6 hydrolysis, ADU precipitation, U 3 O 8 conversion processes were used. The equipment, controls and confinement were set not only by the processes but also by safety requirements according to the kind and physical form of the uranium compounds in each stage and criticality considerations. This paper describes the installation, set up and operation of the plant during production. (Author)

  15. Results of Cesar II critical facility with low enriched fuel balls

    Energy Technology Data Exchange (ETDEWEB)

    Langlet, G; Guerange, J; Laponche, B; Morier, F; Neef, R D; Bock, H J; Kring, F J; Scherer, W

    1972-06-15

    The Cesar facility has been transformed to load in its center a pebble bed fuel. This new Cesar assembly is called Cesar II. The program for the measurements with HTR type fuel balls is managed under a cooperation between physicists of CEA/CADARACHE and KFA/JUELICH. A description of the measuring zones of Cesar II and of the experimental results is given.

  16. Progress in qualifying low-enriched U-Mo dispersion fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Hayes, S.L.; Meyer, M.K.

    2001-01-01

    The U.S. Reduced Enrichment for Research and Test Reactors program is working to qualify dispersions of U-Mo alloys in aluminum with fuel-meat densities of 8 to 9 gU cm -3 . Post irradiation examinations of the small fuel plates irradiated in the Advanced Test Reactor during the high-temperature RERTR-3 tests are virtually complete, and analysis of the large quantity of data obtained is underway. We have observed that the swelling of the fuel plates is stable and modest and that the swelling is dominated by the temperature-dependent interaction of the U-Mo fuel and the aluminum matrix. In order to extract detailed information about the behavior of these fuels from the data, a complex fuel-plate thermal model is being developed to account for the effects of the changing fission rate and thermal conductivity of the fuel meat during irradiation. This paper summarizes the empirical results of the post irradiation examinations and the preliminary results of the model development. In addition, the schedule for irradiation of full-sized elements in the HFR-Petten is briefly discussed. (author)

  17. Development of low enrichment technologies for high density fuels and for isotope production targets

    International Nuclear Information System (INIS)

    Taboada, Horacio; Gonzalez, Alfredo G.

    2005-01-01

    Since more than twenty years ago, CNEA has carried out RERTR activities. Main goals are to convert the RA 6 reactor core from HEU to LEU, to get a comprehensive understanding of U-Mo/Al compounds phase formation in dispersed and monolithic fuels, to develop possible solutions to VHD dispersed and monolithic fuels technical problems, and to optimize techniques to recover U from silicide scrap samples. The future plans include: 1) Completion the RA 6 reactor conversion to LEU; 2) Qualification by irradiation of the promising solutions found for the high density fuels; 3) Irradiation of mini plates and full scale fuel assemblies at the RA 3 reactor and at higher flux and temperature reactors; 4) Optimization of LEU target and radiochemical techniques for radioisotope production. (author) [es

  18. The Ford Nuclear Reactor demonstration project for the evaluation and analysis of low enrichment fuel

    International Nuclear Information System (INIS)

    Kerr, W.; King, J.S.; Lee, J.C.; Martin, W.R.; Wehe, D.K.

    1991-07-01

    The whole-core LEU fuel demonstration project at the University of Michigan was begun in 1979 as part of the Reduced Enrichment Research and Test Reactor (RERTR) Program at Argonne National Laboratory. An LEU fuel design was selected which would produce minimum perturbations in the neutronic, operations, and safety characteristics of the 2-MW Ford Nuclear Reactor (FNR). Initial criticality with a full LEU core on December 8, 1981, was followed by low- and full-power testing of the fresh LEU core, transitional operation with mixed HEU-LEU configurations, and establishment of full LEU equilibrium core operation. The transition from the HEU to the LEU configurations was achieved with negligible impact on experimental utilization and safe operation of the reactor. 78 refs., 74 figs., 84 tabs

  19. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    Energy Technology Data Exchange (ETDEWEB)

    Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  20. Parametric study of the low-enriched uranium integrated Fort-Saint-Vrain element; comparative evaluation with the interacting tubular element

    International Nuclear Information System (INIS)

    Cerles, J.M.; Carvallo, G.; Vallepin, C.

    1971-11-01

    This paper presents a study of the influence of the different geometric and neutronic parameters on the calculation of the cycle with low-enriched uranium in a Fort-Saint-Vrain type brick. The study is divided in two parts: a stage of physics, essentially neutronics; an economical part where the costs are taken into account. At the level of studies of neutronics and costs, a parallel comparison is developed between the brick Fort-Saint-Vrain and the interacting tubular element, and even thorium. 6 refs. 29 figs [fr

  1. Kinetics parameter measurements on RSG-GAS, a low-enriched fuel reactor

    International Nuclear Information System (INIS)

    Jujuratisbela, U; Arbie, B; Pinem, S.; Tukiran; Suparlina, L.; Singh, O.P.

    1995-01-01

    Kinetics parameter measurements, such as reactivity worths of control rods and fuel elements, beam tube void reactivity, power reactivity coefficient and xenon poisoning reactivity have been performed on different cores of Reaktor Serba Guna G.A. Siwabessy (RSG-GAS). In parallel, a programme was also initiated to measure the other kinetics parameters like effective delayed neutron life time, prompt neutron decay constant, validation of period reactivity relationship and zero power frequency response function. The paper provides the results of these measurements. (author)

  2. Spatial correction factors for YALINA Booster facility loaded with medium and low enriched fuels

    International Nuclear Information System (INIS)

    Talamo, A.; Gohar, Y.; Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C.

    2012-01-01

    The Bell and Glasstone spatial correction factor is used in analyses of subcritical assemblies to correct the experimental reactivity as function of the detector position. Besides the detector position, several other parameters affect the correction factor: the energy weighting function of the detector, the detector size, the energy-angle distribution of source neutrons, and the reactivity of the subcritical assembly. This work focuses on the dependency of the correction factor on the detector material and it investigates the YALINA Booster subcritical assembly loaded with medium (36%) and low (10%) enriched fuels. (authors)

  3. Spatial correction factors for YALINA Booster facility loaded with medium and low enriched fuels

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, A.; Gohar, Y. [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C. [Joint Inst. for Power and Nuclear Research-Sosny, 99 Academician A.K.Krasin Str, Minsk 220109 (Belarus)

    2012-07-01

    The Bell and Glasstone spatial correction factor is used in analyses of subcritical assemblies to correct the experimental reactivity as function of the detector position. Besides the detector position, several other parameters affect the correction factor: the energy weighting function of the detector, the detector size, the energy-angle distribution of source neutrons, and the reactivity of the subcritical assembly. This work focuses on the dependency of the correction factor on the detector material and it investigates the YALINA Booster subcritical assembly loaded with medium (36%) and low (10%) enriched fuels. (authors)

  4. Safe use of the Institute of Nuclear Physics reactor with low enriched fuel

    International Nuclear Information System (INIS)

    Baytelesov, S.A.; Dosimbaev, A.A.; Koblik, Yu.N.; Salikhbaev, U.S.; Khalikov, U.A.; Yuldashev, B.S.

    2006-01-01

    Full text: The requirements for safe exploitation of reactor do not accept boiling of water on the surface of fuel elements. At determination of safe thermal regime of reactor (permissible level of power) the regime of the most heat-stressed fuel assembly (FA) in the active core was analyzed. By using ASTRA code [1] the heat-stressed sector is determined by most heat-stressed FA. In calculations the power of reactor was selected so that stock factor prior to the water boiling on the FA surface was not less than 1.45. Besides, in calculations the value of maximal energy density in examined FA is decreased by 10 %. As the part of the energy generated in the FA cores will be lost in constructional materials of the active zone and on the reflector. The stocks of safety before occurrence of instability of flow in gaps between of FA and before crisis of heat exchange are also analyzed. Further, by using the MCNP-4C code [2], densities of fast (E > 0,821 MeV) and thermal flows (E < 0,625 eV) of neutrons were calculated for those experimental channels where the irradiation of samples would be carried out. (author)

  5. HEATHYD, Steady-State Thermal Hydraulic Analysis of Low-Enriched U Fuel Reactor

    International Nuclear Information System (INIS)

    NABBI, R.

    1989-01-01

    1 - Description of program or function: HEATHYD is a code for the steady-state heat transfer calculation of research nuclear reactors with forced convection. It models heat transfer and coolant flow for assemblies of parallel fuel plates of MTR type with any axial power distribution. The thermodynamic model accounts for single phase cooling and sub- cooled boiling condition using the transition criterion of Bergeles-Rosenow. In addition to the calculation of the channel flow velocities and coolant pressure drops, HEATHYD calculates axial distribution of the coolant and clad-surface temperatures. Safety margins to the critical heat flux as a result of burnout condition or flow instability are determined. 2 - Method of solution: Applying the finite difference method, HEATHYD solves the equations of heat conduction and heat transfer to the coolant. For the physical properties of the coolant as a function of the coolant temperature polynomials of degree 6 are used. Depending on the coolant condition, different correlations for the heat transfer coefficient can be applied. The analysis of the critical cooling conditions resulting in burnout or flow instability, is performed according to the correlations developed by Mirshak/ Labuntsov and Forgan/Whittle

  6. Comparison of low enriched uranium (UAlx-Al and U-Ni) targets with different geometries for the production of molybdenum-99 in the RMB (Brazilian multipurpose reactor)

    International Nuclear Information System (INIS)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da; Angelo, Gabriel; Fedorenko, Giuliana G.; Nishiyama, Pedro J.B. de O.

    2011-01-01

    The Brazilian Multipurpose Reactor (RMB), now in the conception design phase, is being designed in Brazil to attend the demand of radiopharmaceuticals in the country and conduct researches in various areas. The new reactor, planned for 30 MW, will replace the IEA-R1 reactor of IPEN-CNEN/SP. Low enriched uranium ( 235 U) UAl x dispersed in Al (plate geometry) and metallic uranium foil targets (plate and cylinder geometries) are being considered for production of Molybdenum-99 ( 99 Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of 99 Mo for these targets in the RMB. For the neutronic calculations were utilized the computer codes Hammer-Technion, Citation and Scale and for the thermal-hydraulics calculations were utilized the computer code MTRCR-IEAR1 and ANSYS CFX. (author)

  7. Delayed Fission Product Gamma-Ray Transmission Through Low Enriched UO2 Fuel Pin Lattices in Air

    Energy Technology Data Exchange (ETDEWEB)

    Trumbull, TH [Rensselaer Polytechnic Inst., Troy, NY (United States)

    2004-10-18

    The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray dose on the outside of the lattice as a function of radial position, the so-called “channeling” effect, was analyzed. Techniques for performing experimental measurements and data reduction at Rensselaer Polytechnic Institute’s Reactor Critical Facility (RCF) were derived. An experimental apparatus was constructed to hold the arrangements of fuel pins for the measurements. A gamma-ray spectroscopy system consisting of a sodium-iodide scintillation detector was used to collect data. Measurements were made with and without a collimator installed. A point-kernel transport code was developed to map the radial dependence of the gamma-ray flux. Input files for the Monte Carlo code, MCNP, were also developed to accurately model the experimental measurements. The results of the calculations were compared to the experimental measurements. In order to determine the delayed fission-product gamma-ray source for the calculations, a technique was developed using a previously written code, DELBG and the reactor state-point data obtained during the experimental measurements. Calculations were performed demonstrating the effects of material homogenization on the gamma-ray transmission through the fuel pin lattice.Homogeneous and heterogeneous calculations were performed for all RCF fuel pin lattices as well as for a typical commercial pressurized water reactor fuel bundle. The results of the study demonstrated the effectiveness of the experimental measurements to isolate the channeling effect of delayed fission-product gamma-rays through lattices of RCF fuel pins

  8. Uranium plutonium oxide fuels

    International Nuclear Information System (INIS)

    Cox, C.M.; Leggett, R.D.; Weber, E.T.

    1981-01-01

    Uranium plutonium oxide is the principal fuel material for liquid metal fast breeder reactors (LMFBR's) throughout the world. Development of this material has been a reasonably straightforward evolution from the UO 2 used routinely in the light water reactor (LWR's); but, because of the lower neutron capture cross sections and much lower coolant pressures in the sodium cooled LMFBR's, the fuel is operated to much higher discharge exposures than that of a LWR. A typical LMFBR fuel assembly is shown. Depending on the required power output and the configuration of the reactor, some 70 to 400 such fuel assemblies are clustered to form the core. There is a wide variation in cross section and length of the assemblies where the increasing size reflects a chronological increase in plant size and power output as well as considerations of decreasing the net fuel cycle cost. Design and performance characteristics are described

  9. Oscillator measurements of the reactivity changes resulting from the irradiation of low enrichment particulate fuel in the Dragon reactor

    International Nuclear Information System (INIS)

    Burbidge, B.L.H.; Franklin, B.M.; Small, V.G.

    1983-01-01

    This Report describes a series of experiments carried out as a joint UKAEA/CEA/DRAGON project to determine the reactivity changes of low-enrichment particulate fuel samples following their irradiation in the DRAGON reactor to various levels up to approximately 60,000 MWD/Te. The samples are described, together with the method of measurement of reactivity in the Winfrith reactor HECTOR, which was an extension of the well-known Oscillator Technique to yield simultaneously overall reactivity changes and changes in macroscopic absorption cross-sections. Measurements were carried out at room temperature in two reactor spectra; a thermal spectrum and one typical of an HTR type reactor. The resultant reactivity changes are presented together with the relevant sample burn-ups as determined by #betta#-scanning methods and, in some cases, by rigorous chemical analysis. The results of supporting measurements are also reported, carried out to characterise the neutron spectra in which the oscillator measurements were made and to determine the neutron flux distributions in the HECTOR reactor. (author)

  10. Progress in developing processes for converting 99Mo production from high- to low-enriched uranium--1998

    International Nuclear Information System (INIS)

    Conner, C.

    1998-01-01

    During 1998, the emphasis of our activities was focused mainly on target fabrication. Successful conversion requires a reliable irradiation target; the target being developed uses thin foils of uranium metal, which can be removed from the target hardware for dissolution and processing. This paper describes successes in (1) improving our method for heat-treating the uranium foil to produce a random-small grain structure, (2) improving electrodeposition of zinc and nickel fission-fragment barriers onto the foil, and (3) showing that these fission fragment barriers should be stable during transport of the targets following irradiation. A method was also developed for quantitatively electrodepositing uranium and plutonium contaminants in the 99 Mo. Progress was also made in broadening international cooperation in our development activities

  11. 78 FR 21100 - Low Enriched Uranium From France: Final Results of the Expedited Second Sunset Review of the...

    Science.gov (United States)

    2013-04-09

    ... received no response from the respondent interested parties, i.e., French uranium producers and exporters... Centralized Electronic Service System (IA ACCESS). IA ACCESS is available to registered users at http... the Internet at http://trade.gov/ia/ . The signed Decision Memorandum and electronic versions of the...

  12. Validation of KENO V.a. and two cross-section libraries for criticality calculations of low-enriched uranium systems

    International Nuclear Information System (INIS)

    Easter, M.E.

    1985-07-01

    The SCALE code system, utilizing the Monte Carlo computer code KENO V.a, was employed to calculate 37 critical experiments. The critical assemblies had 235 U enrichments of 5% or less and cover a variety of geometries and materials. Values of k/sub eff/ were calculated using two different results using either of the cross-section libraries. The 16-energy-group Hansen-Roach and the 27-energy-group ENDF/B-IV cross-section libraries, available in SCALE, were used in this validation study, and both give good results for the experiments considered. It is concluded that the code and cross sections are adequate for low-enriched uranium systems and that reliable criticality safety calculations can be made for such systems provided the limits of validated applicability are not exceeded

  13. Postirradiation analysis of experimental uranium-silicide dispersion fuel plates

    International Nuclear Information System (INIS)

    Hofman, G.L.; Neimark, L.A.

    1985-01-01

    Low-enriched uranium silicide dispersion fuel plates were irradiated to maximum burnups of 96% of 235 U. Fuel plates containing 33 v/o U 3 Si and U 3 Si 2 behaved very well up to this burnup. Plates containing 33 v/o U 3 Si-Al pillowed between 90 and 96% burnup of the fissile atoms. More highly loaded U 3 Si-Al plates, up to 50 v/o were found to pillow at lower burnups. Plates containing 40 v/o U 3 Si showed an increase swelling rate around 85% burnup. 5 refs., 10 figs

  14. Basic research on high-uranium density fuels for research and test reactors

    International Nuclear Information System (INIS)

    Ugajin, M.; Itoh, A.; Akabori, M.

    1992-01-01

    High-uranium density fuels, uranium silicides (U 3 Si 2 , U 3 Si) and U 6 Me-type uranium alloys (Me = Fe, Mn, Ni), were prepared and examined metallurgically as low-enriched uranium (LEU) fuels for research and test reactors. Miniature aluminum-dispersion plate-type fuel (miniplate) and aluminum-clad disk-type fuel specimens were fabricated and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Fuel-aluminum compatibility tests were conducted to elucidate the extent of reaction and to identify reaction products. The relative stability of the fuels in an aluminum matrix was established at 350degC or above. Experiments were also performed to predict the chemical form of the solid fission-products in the uranium silicide (U 3 Si 2 ) simulating a high burnup anticipated for reactor service. (author)

  15. Optimization of neutronic characteristics of U3Si2 low enrichment fuel elements for a new design of IEA-R1 reactor core

    International Nuclear Information System (INIS)

    Mai, L.A.; Maiorino, J.R.; Gouvea, E.A.

    1989-01-01

    This work shows a study of neutronic optimization of U 3 Si 2 -Al low enrichment fuel element. This study has a goal to propose a optimized Core to be used in the research reactor IEA-R1. The external dimensions of the fuel element were maintained as constraints and the loss of reactivity along fuel life-time was defined as 'objective function', and it has been minimized by varying the fuel element dimensions. Cell calculations were made with HAMMER-TECH /3/ Code, for burnups up to 50% of U-235 initial mass. The Computer values of the objective function for several combinations of fuel element dimensions were fitted by a surface using the SAS system /9/, and it has been minimized by a Harwell subroutine /10/. (author) [pt

  16. Research and design calculation of multipurpose critical assembly using moderated light water and low enriched fuel from 1.6 to 5.0% U-235

    International Nuclear Information System (INIS)

    Nguyen Kien Cuong; Vo Doan Hai Dang; Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Minh Tuan; Nguyen Manh Hung; Pham Quang Huy; Tran Quoc Duong; Tran Tri Vien

    2015-01-01

    Basing on the idea in ??using fuel of nuclear power plants such as PWR (AP-1000) and VVER-1000 with light water as moderation, design calculation of critical assembly was performed to confirm the possibility of using these fuels. Designed critical assembly has simple structure consisting of low enriched fuel from 1.6% to 5% U-235; water has functions as cooling, biological protection and control. Critical assembly is operated at nominal power 100 W with fuel pitch about 2.0 cm. Applications of the critical assembly are quite abundant in basic research, education and training with low investment cost compare with research reactor and easy in operation. So critical assembly can be used for university or training centre for nuclear engineering training. Main objectives of the project are: design calculation in neutronics, thermal hydraulics and safety analysis for critical configuration benchmarks using low enriched fuel; design in mechanical and auxiliary systems for critical assembly; determine technical specifications and estimate construction, installation cost of critical assembly. The process of design, fabrication, installation and construction of critical assembly will be considered with different implementation phases and localization capabilities in installation of critical assembly is highly feasibility. Cost estimation of construction and installation of critical assembly was implemented and showed that investment cost for critical assembly is much lower than research reactor and most of components, systems of critical assembly can be localized with current technique quality of the country. (author)

  17. In-Situ Measurements of Low Enrichment Uranium Holdup Process Gas Piping at K-25 - Paper for Waste Management Symposia 2010 East Tennessee Technology Park Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    Rasmussen, B.

    2010-01-01

    This document is the final version of a paper submitted to the Waste Management Symposia, Phoenix, 2010, abstract BJC/OR-3280. The primary document from which this paper was condensed is In-Situ Measurement of Low Enrichment Uranium Holdup in Process Gas Piping at K-25 Using NaI/HMS4 Gamma Detection Systems, BJC/OR-3355. This work explores the sufficiency and limitations of the Holdup Measurement System 4 (HJVIS4) software algorithms applied to measurements of low enriched uranium holdup in gaseous diffusion process gas piping. HMS4 has been used extensively during the decommissioning and demolition project of the K-25 building for U-235 holdup quantification. The HMS4 software is an integral part of one of the primary nondestructive assay (NDA) systems which was successfully tested and qualified for holdup deposit quantification in the process gas piping of the K-25 building. The initial qualification focused on the measurement of highly enriched UO 2 F 2 deposits. The purpose of this work was to determine if that qualification could be extended to include the quantification of holdup in UO 2 F 2 deposits of lower enrichment. Sample field data are presented to provide evidence in support of the theoretical foundation. The HMS4 algorithms were investigated in detail and found to sufficiently compensate for UO 2 F 2 source self-attenuation effects, over the range of expected enrichment (4-40%), in the North and East Wings of the K-25 building. The limitations of the HMS4 algorithms were explored for a described set of conditions with respect to area source measurements of low enriched UO 2 F 2 deposits when used in conjunction with a 1 inch by 1/2 inch sodium iodide (NaI) scintillation detector. The theoretical limitations of HMS4, based on the expected conditions in the process gas system of the K-25 building, are related back to the required data quality objectives (DQO) for the NBA measurement system established for the K-25 demolition project. The combined

  18. Low enriched uranium UAl{sub X}-Al targets for the production of Molybdenum-99 in the IEA-R1 and RMB reactors

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2011-07-01

    The IEA-R1 reactor of IPEN/CNEN-SP in Brazil is a pool type research reactor cooled and moderated by demineralized water and having Beryllium and Graphite as reflectors. In 1997 the reactor received the operating licensing for 5 MW. A new research reactor is being planned in Brazil to replace the IEA-R1 reactor. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Low enriched uranium (LEU) (<20% {sup 235}U) UAl{sub x} dispersed in Al targets are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed, respectively, to compare the production of {sup 99}Mo for these targets in IEA-R1 reactor and RMB and to determine the temperatures achieved in the UAl{sub x}-Al targets during irradiation. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations was utilized the computer code MTRCR-IEAR1. (author)

  19. Acceptable standard format and content for the fundamental nuclear material control (FNMC) plan required for low-enriched uranium facilities. Revision 2

    International Nuclear Information System (INIS)

    Joy, D.R.

    1995-12-01

    This report documents a standard format suggested by the NRC for use in preparing fundamental nuclear material control (FNMC) plans as required by the Low Enriched Uranium Reform Amendments (10CFR 74.31). This report also describes the necessary contents of a comprehensive plan and provides example acceptance criteria which are intended to communicate acceptable means of achieving the performance capabilities of the Reform Amendments. By using the suggested format, the licensee or applicant will minimize administrative problems associated with the submittal, review and approval of the FNMC plan. Preparation of the plan in accordance with this format Will assist the NRC in evaluating the plan and in standardizing the review and licensing process. However, conformance with this guidance is not required by the NRC. A license applicant who employs a format that provides a equal level of completeness and detail may use their own format. This document is also intended for providing guidance to licensees when making revisions to their FNMC plan

  20. Trace metal assay of uranium silicide fuel

    International Nuclear Information System (INIS)

    Kulkarni, M.J.; Argekar, A.A.; Thulasidas, S.K.; Dhawale, B.A.; Rajeswari, B.; Adya, V.C.; Purohit, P.J.; Neelam, G.; Bangia, T.R.; Page, A.G.; Sastry, M.D.; Iyer, R.H.

    1994-01-01

    A comprehensive trace metal assay of uranium silicide, a fuel for nuclear research reactors that employs low-enrichment uranium, is carried out by atomic spectrometry. Of the list of specification elements, 21 metallic elements are determined by a direct current (dc) arc carrier distillation technique; the rare earths yttrium and zirconium are chemically separated from the major matrix followed by a dc arc/inductively coupled argon plasma (ICP) excitation technique in atomic emission spectrometry (AES); silver is determined by electrothermal atomization-atomic absorption spectrometry (ETA-AAS) without prior chemical separation of the major matrix. Gamma radioactive tracers are used to check the recovery of rare earths during the chemical separation procedure. The detection limits for trace metallics vary in the 0.1- to 40-ppm range. The precision of the determinations as evaluated from the analysis of the synthetic sample with intermediate range analyte concentration is better than 25% relative standard deviation (RSD) for most of the elements employing dc arc-AES, while that for silver determination by ETS-AAS is 10% RSD. The precision of the determinations for four crucially important rare earths by ICP-AES is better than 3% RSD

  1. Development of low enriched uranium target plates by thermo-mechanical processing of UAl2–Al matrix for production of 99Mo in Pakistan

    International Nuclear Information System (INIS)

    Ali, Kanwar Liaqat; Khan, Akhlaque Ahmad; Mushtaq, Ahmad; Imtiaz, Farhan; Ziai, Maratab Ali; Gulzar, Amir; Farooq, Muhammad; Hussain, Nazar; Ahmed, Nisar; Pervez, Shahid; Zaidi, Jamshed Hussain

    2013-01-01

    Uranium aluminide predominated with UAl 2 phase was prepared by arc-melting procedures and comminuted to required particle size. UAl 2 and Al powders were blended and compacted to achieve LEU fuel density of 2.17 g/cm 3 . The picture-frame technique was used to clad the dispersions (UAl 2 –Al) with aluminum. A few target plates were fabricated by thermo-mechanical processing (hot rolling and annealing) of UAl 2 –Al matrix contained in roll billet of Al. The fabricated plates were characterized by destructive and some of non-destructive testing techniques and then annealed to achieve required phase of uranium aluminide for proper dissolution in basic media

  2. Post-irradiation analysis of low enriched U-Mo/Al dispersions fuel miniplate tests, RERTR 4 and 5

    International Nuclear Information System (INIS)

    Hofman, G.L.; Finlay, M.R.; Kim, Y.S.

    2005-01-01

    Interpretation of the post irradiation data of U-Mo/Al dispersion fuel mini plates irradiated in the Advanced Test Reactor to a maximum U-235 burn up of 80% are presented. The analyses addresses fuel swelling and porosity formation as these fuel performance issues relate to fuel fabrication and irradiation parameters. Specifically, mechanisms involved in the formation of porosity observed in the U-Mo/Al interaction phase are discussed and, means of mitigating or eliminating this irradiation phenomenon are offered. (author)

  3. Uranium - the nuclear fuel

    International Nuclear Information System (INIS)

    Smith, E.E.N.

    1976-01-01

    A brief history is presented of Canadian uranium exploration, production, and sales. Statistics show that Canada is a good customer for its own uranium due to a rapidly expanding nuclear power program. Due to an average 10 year lag between commencement of exploration and production, and with current producers sold out through 1985, it is imperative that exploration efforts be increased. (E.C.B.)

  4. Communication dated 19 May 2011 received from the Resident Representative of the United Kingdom of Great Britain and Northern Ireland to the Agency regarding Assurance of Supply of Enrichment Services and Low Enriched Uranium for Use in Nuclear Power Plants

    International Nuclear Information System (INIS)

    2011-01-01

    The Secretariat has received a letter dated 19 May 2011 from the Resident Representative of the United Kingdom of Great Britain and Northern Ireland to the Agency, attaching the Proposal for the Assurance of Supply of Enrichment Services and Low Enriched Uranium for Use in Nuclear Power Plants, as described in document GOV/2011/10. As requested by the Resident Representative, the letter and its attachment are circulated herewith for information of all Member States

  5. Communication dated 19 May 2011 received from the Resident Representative of the United Kingdom of Great Britain and Northern Ireland to the Agency regarding Assurance of Supply of Enrichment Services and Low Enriched Uranium for Use in Nuclear Power Plants

    International Nuclear Information System (INIS)

    2011-01-01

    The Secretariat has received a letter dated 19 May 2011 from the Resident Representative of the United Kingdom of Great Britain and Northern Ireland to the Agency, attaching the Proposal for the Assurance of Supply of Enrichment Services and Low Enriched Uranium for Use in Nuclear Power Plants, as described in document GOV/2011/10. As requested by the Resident Representative, the letter and its attachment are circulated herewith for information of all Member States [es

  6. Feasibility of Producing Molybdenum-99 on a Small Scale Using Fission of Low Enriched Uranium or Neutron Activation of Natural Molybdenum

    International Nuclear Information System (INIS)

    2015-01-01

    This publication documents the work performed within the IAEA coordinated research project (CRP) on Developing Techniques for Small Scale Indigenous Molybdenum-99 Production Using LEU Fission or Neutron Activation. The project allowed participating institutions to receive training and information on aspects necessary for starting production of molybdenum-99 ( 99 Mo) on a small scale, that is, to become national level producers of this medical isotope. Stable production of 99Mo is one of the most pressing issues facing the nuclear community at present, because the medical isotope technetium-99m ( 99m Tc), which decays from 99 Mo, is one of the most widely used radionuclides in diagnostic imaging and treatment around the world. In the past five years, there have been widespread shortages of 99 Mo owing to the limited number of producers, many of which use ageing facilities. To assist in stabilizing the production of 99Mo, and to promote the use of production methods that do not rely on the use of highly enriched uranium (HEU), the IAEA initiated the abovementioned CRP on small scale 99Mo production using low enriched uranium (LEU) fission or neutron activation methods. The intention was to enable participating institutions to gain the knowledge necessary to become national level producers of 99Mo in the event of further global shortages. Some of the institutions that participated in the CRP have continued their work on 99 Mo production, and are enlisting the assistance of other CRP members and the IAEA’s technical cooperation programme to set up a small scale production capability. In total, the CRP was active for six years, and concluded in December 2011. During the CRP, fourteen IAEA Member States took part; four research coordination meetings were held, and four workshops were held on operational aspects of 99 Mo production, LEU target fabrication and waste management. Most participants carried out work related to the entire production process, from target

  7. Evaluation of core physics analysis methods for conversion of the INL advanced test reactor to low-enrichment fuel

    International Nuclear Information System (INIS)

    DeHart, M. D.; Chang, G. S.

    2012-01-01

    Computational neutronics studies to support the possible conversion of the ATR to LEU are underway. Simultaneously, INL is engaged in a physics methods upgrade project to put into place modern computational neutronics tools for future support of ATR fuel cycle and experiment analysis. A number of experimental measurements have been performed in the ATRC in support of the methods upgrade project, and are being used to validate the new core physics methods. The current computational neutronics work is focused on performance of scoping calculations for the ATR core loaded with a candidate LEU fuel design. This will serve as independent confirmation of analyses that have been performed previously, and will evaluate some of the new computational methods for analysis of a candidate LEU fuel for ATR. (authors)

  8. Accelerator-driven sub-critical research facility with low-enriched fuel in lead matrix: Neutron flux calculation

    Directory of Open Access Journals (Sweden)

    Avramović Ivana

    2007-01-01

    Full Text Available The H5B is a concept of an accelerator-driven sub-critical research facility (ADSRF being developed over the last couple of years at the Vinča Institute of Nuclear Sciences, Belgrade, Serbia. Using well-known computer codes, the MCNPX and MCNP, this paper deals with the results of a tar get study and neutron flux calculations in the sub-critical core. The neutron source is generated by an interaction of a proton or deuteron beam with the target placed inside the sub-critical core. The results of the total neutron flux density escaping the target and calculations of neutron yields for different target materials are also given here. Neutrons escaping the target volume with the group spectra (first step are used to specify a neutron source for further numerical simulations of the neutron flux density in the sub-critical core (second step. The results of the calculations of the neutron effective multiplication factor keff and neutron generation time L for the ADSRF model have also been presented. Neutron spectra calculations for an ADSRF with an uranium tar get (highest values of the neutron yield for the selected sub-critical core cells for both beams have also been presented in this paper.

  9. MTR (Materials Testing Reactors) cores fuel management. Application of a low enrichment reactor for the equilibrium and transitory core calculation

    International Nuclear Information System (INIS)

    Relloso, J.M.

    1990-01-01

    This work describes a methodology to define the equilibrium core and a MTR (Materials Testing Reactors) type reactor's fuel management upon multiple boundary conditions, such as: end cycle and permitted maximum reactivities, burn-up extraction and maximun number of movements by rechange. The methodology proposed allows to determine the best options through conceptual relations, prior to a detailed calculation with the core code, reducing the test number with these codes and minimizing in this way CPU cost. The way to better systematized search of transient cores from the first one to the equilibrium one is presented. (Author) [es

  10. Metallic uranium as fuel for fast reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de

    1988-01-01

    This paper presents a first overview of the use of metallic uranium and its alloys as an option for fuel for rapid reactors. Aspects are discussed concerning uranium alloys which present high solubility in the gamma phase. (author)

  11. Uranium dioxide Caramel fuel

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    The work performed in France on Caramel fuels for research reactors reflects the reality of a program based on non proliferation criteria, as they have already appeared several years ago. This work actually includes the following different aspects: identification of the non proliferation criterion defining this action; determination of the economical and technical goals to be reached; realization of research and development studies finalized in a full scale demonstration; transposition to an industrial and commercial level

  12. Critical experiment program of heterogeneous core composed for LWR fuel rods and low enriched uranyl nitrate solution

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori; Yamamoto, Toshihiro; Watanabe, Shouichi; Nakamura, Takemi

    2003-01-01

    In order to stimulate the criticality characteristics of a dissolver in a reprocessing plant, a critical experiment program of heterogeneous cores is under going at a Static Critical Experimental Facility, STACY in Japan Atomic Energy Research Institute, JAERI. The experimental system is composed of 5w/o enriched PWR-type fuel rod array immersed in 6w/o enriched uranyl nitrate solution. First series of experiments are basic benchmark experiments on fundamental critical data in order to validate criticality calculation codes for 'general-form system' classified in the Japanese Criticality Safety Handbook, JCSHB. Second series of experiments are concerning the neutron absorber effects of fission products related to the burn-up credit Level-2. For demonstrating the reactivity effects of fission products, reactivity effects of natural elements such as Sm, Nd, Eu and 103 Rh, 133 Cs, solved in the nitrate solution are to be measured. The objective of third series of experiments is to validate the effect of gadolinium as a soluble neutron poison. Properties of temperature coefficients and kinetic parameters are also studied, since these parameters are important to evaluate the transient behavior of the criticality accident. (author)

  13. Operation of Nuclear Fuel Based on Reprocessed Uranium for VVER-type Reactors in Competitive Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Troyanov, V.; Molchanov, V.; Tuzov, A. [TVEL Corporation, 49 Kashirskoe shosse, Moscow 115409 (Russian Federation); Semchenkov, Yu.; Lizorkin, M. [RRC ' Kurchatov Institute' (Russian Federation); Vasilchenko, I.; Lushin, V. [OKB ' Gidropress' (Russian Federation)

    2009-06-15

    Current nuclear fuel cycle of Russian nuclear power involves reprocessed low-enriched uranium in nuclear fuel production for some NPP units with VVER-type LWR. This paper discusses design and performance characteristics of commercial nuclear fuel based on natural and reprocessed uranium. It presents the review of results of commercial operation of nuclear fuel based on reprocessed uranium on Russian NPPs-unit No.2 of Kola NPP and unit No.2 of Kalinin NPP. The results of calculation and experimental validation of safe fuel operation including necessary isotope composition conformed to regulation requirements and results of pilot fuel operation are also considered. Meeting the customer requirements the possibility of high burn-up achieving was demonstrated. In addition the paper compares the characteristics of nuclear fuel cycles with maximum length based on reprocessed and natural uranium considering relevant 5% enrichment limitation and necessity of {sup 236}U compensation. The expedience of uranium-235 enrichment increasing over 5% is discussed with the aim to implement longer fuel cycles. (authors)

  14. Comparison of low enriched uranium (UAl{sub x}-Al and U-Ni) targets with different geometries for the production of molybdenum-99 in the RMB (Brazilian multipurpose reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da; Angelo, Gabriel; Fedorenko, Giuliana G., E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil)

    2011-07-01

    The Brazilian Multipurpose Reactor (RMB), now in the conception design phase, is being designed in Brazil to attend the demand of radiopharmaceuticals in the country and conduct researches in various areas. The new reactor, planned for 30 MW, will replace the IEA-R1 reactor of IPEN-CNEN/SP. Low enriched uranium (<20% {sup 235}U) UAl{sub x} dispersed in Al (plate geometry) and metallic uranium foil targets (plate and cylinder geometries) are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of {sup 99}Mo for these targets in the RMB. For the neutronic calculations were utilized the computer codes Hammer-Technion, Citation and Scale and for the thermal-hydraulics calculations were utilized the computer code MTRCR-IEAR1 and ANSYS CFX. (author)

  15. Production of molybdenum-99 by heterogeneous and homogeneous uranium fueled reactors

    International Nuclear Information System (INIS)

    Carlin, G.E.; Bonin, H.W.

    2012-01-01

    The use of radioisotopes for various procedures in the health care industry has become one of the most important practices in medicine. At the forefront of the medical isotope list is molybdenum-99 and its daughter isotope technetium-99m, which encompass over 80% of radiopharmaceutical procedures. Fission of uranium-235 to produce molybdenum-99 is the most widely used method for producing this radioisotope. The heterogeneous reactor and the aqueous homogeneous reactor are looked at here with emphasis on the use of low enriched uranium as the fuel source. Methods of technetium-99m generation and its medical use are also reviewed. (author)

  16. Production of molybdenum-99 by heterogeneous and homogeneous uranium fueled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carlin, G.E.; Bonin, H.W., E-mail: george.carlin@rmc.ca, E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

    2012-07-01

    The use of radioisotopes for various procedures in the health care industry has become one of the most important practices in medicine. At the forefront of the medical isotope list is molybdenum-99 and its daughter isotope technetium-99m, which encompass over 80% of radiopharmaceutical procedures. Fission of uranium-235 to produce molybdenum-99 is the most widely used method for producing this radioisotope. The heterogeneous reactor and the aqueous homogeneous reactor are looked at here with emphasis on the use of low enriched uranium as the fuel source. Methods of technetium-99m generation and its medical use are also reviewed. (author)

  17. Fuel powder production from ductile uranium alloys

    International Nuclear Information System (INIS)

    Clark, C.R.; Meyer, M.K.

    1998-01-01

    Metallic uranium alloys are candidate materials for use as the fuel phase in very-high-density LEU dispersion fuels. These ductile alloys cannot be converted to powder form by the processes routinely used for oxides or intermetallics. Three methods of powder production from uranium alloys have been investigated within the US-RERTR program. These processes are grinding, cryogenic milling, and hydride-dehydride. In addition, a gas atomization process was investigated using gold as a surrogate for uranium. (author)

  18. Slightly enriched uranium fuel for a PHWR

    International Nuclear Information System (INIS)

    Notari, C.; Marajofsky, A.

    1997-01-01

    An improved fuel element design for a PHWR using slightly enriched uranium fuel is presented. It maintains the general geometric disposition of the currently used in the argentine NPP's reactors, replacing the outer ring of rods by rods containing annular pellets. Power density reduction is achieved with modest burnup losses and the void volume in the pellets can be used to balance these two opposite effects. The results show that with this new design, the fuel can be operated at higher powers without violating thermohydraulic limits and this means an improvement in fuel management flexibility, particularly in the transition from natural uranium to slightly enriched uranium cycle. (author)

  19. A Preliminary Study on the Reuse of the Recovered Uranium from the Spent CANDU Fuel Using Pyroprocessing

    International Nuclear Information System (INIS)

    Park, C. J.; Na, S. H.; Yang, J. H.; Kang, K. H.; Lee, J. W.

    2009-01-01

    During the pyroprocessing, most of the uranium is gathered in metallic form around a solid cathode during an electro-refining process, which is composed of about 94 weight percent of the spent fuel. In the previous study, a feasibility study has been done to reuse the recovered uranium for the CANDU reactor fuel following the traditional DUPIC (direct use of spent pressurized water reactor fuel into CANDU reactor) fuel fabrication process. However, the weight percent of U-235 in the recovered uranium is about 1 wt% and it is sufficiently re-utilized in a heavy water reactor which uses a natural uranium fuel. The reuse of recovered uranium will bring not only a huge economic profit and saving of uranium resources but also an alleviation of the burden on the management and the disposal of the spent fuel. The research on recycling of recovered uranium was carried out 10 years ago and most of the recovered uranium was assumed to be imported from abroad at that time. The preliminary results showed there is the sufficient possibility to recycle recovered uranium in terms of a reactor's characteristics as well as the fuel performance. However, the spent CANDU fuel is another issue in the storage and disposal problem. At present, most countries are considering that the spent CANDU fuel is disposed directly due to the low enrichment (∼0.5 wt%) of the discharge fissile content and lots of fission products. If mixing the spent CANDU fuel and the spent PWR fuel, the estimated uranium fissile enrichment will be about 0.6 wt% ∼ 1.0 wt% depending on the mixing ratio, which is sufficiently reusable in a CANDU reactor. Therefore, this paper deals with a feasibility study on the recovered uranium of the mixed spent fuel from the pyroprocessing. With the various mixing ratios between the PWR spent fuel and the CANDU spent fuel, a reactor characteristics including the safety parameters of the CANDU reactor was evaluated

  20. Uranium production in thorium/denatured uranium fueled PWRs

    International Nuclear Information System (INIS)

    Arthur, W.B.

    1977-01-01

    Uranium-232 buildup in a thorium/denatured uranium fueled pressurized water reactor, PWR(Th), was studied using a modified version of the spectrum-dependent zero dimensional depletion code, LEOPARD. The generic Combustion Engineering System 80 reactor design was selected as the reactor model for the calculations. Reactors fueled with either enriched natural uranium and self-generated recycled uranium or uranium from a thorium breeder and self-generated recycled uranium were considered. For enriched natural uranium, concentrations of 232 U varied from about 135 ppM ( 232 U/U weight basis) in the zeroth generation to about 260 ppM ( 232 U/U weight basis) at the end of the fifth generation. For the case in which thorium breeder fuel (with its relatively high 232 U concentration) was used as reactor makeup fuel, concentrations of 232 U varied from 441 ppM ( 232 U/U weight basis) at discharge from the first generation to about 512 ppM ( 232 U/U weight basis) at the end of the fifth generation. Concentrations in freshly fabricated fuel for this later case were 20 to 35% higher than the discharge concentration. These concentrations are low when compared to those of other thorium fueled reactor types (HTGR and MSBR) because of the relatively high 238 U concentration added to the fuel as a denaturant. Excellent agreement was found between calculated and existing experimental values. Nevertheless, caution is urged in the use of these values because experimental results are very limited, and the relevant nuclear data, especially for 231 Pa and 232 U, are not of high quality

  1. Agreement Between the International Atomic Energy Agency, the Government of Jamaica and the Government of the United States of America for Assistance in Securing Low Enriched Uranium for a Research Reactor

    International Nuclear Information System (INIS)

    2014-01-01

    The text of the Agreement between the International Atomic Energy Agency, the Government of Jamaica and the Government of the United States of America for Assistance in Securing Low Enriched Uranium for a Research Reactor is reproduced in this document for the information of all Members of the Agency. The Agency's Board of Governors approved the text of the Agreement on 6 March 2013. The Agreement was signed by the authorized representatives of Jamaica on 25 November 2013, the United States on 2 May 2013 and the Director General of the IAEA on 16 December 2013. Pursuant to the Article XI of the Agreement, the Agreement entered into force on 16 December 2013, upon signature by the Director General of the IAEA and by the authorized representatives of Jamaica and the United States [fr

  2. Agreement Between the International Atomic Energy Agency, the Government of Jamaica and the Government of the United States of America for Assistance in Securing Low Enriched Uranium for a Research Reactor

    International Nuclear Information System (INIS)

    2014-01-01

    The text of the Agreement between the International Atomic Energy Agency, the Government of Jamaica and the Government of the United States of America for Assistance in Securing Low Enriched Uranium for a Research Reactor is reproduced in this document for the information of all Members of the Agency. The Agency's Board of Governors approved the text of the Agreement on 6 March 2013. The Agreement was signed by the authorized representatives of Jamaica on 25 November 2013, the United States on 2 May 2013 and the Director General of the IAEA on 16 December 2013. Pursuant to the Article XI of the Agreement, the Agreement entered into force on 16 December 2013, upon signature by the Director General of the IAEA and by the authorized representatives of Jamaica and the United States

  3. Agreement Between the International Atomic Energy Agency, the Government of Jamaica and the Government of the United States of America for Assistance in Securing Low Enriched Uranium for a Research Reactor

    International Nuclear Information System (INIS)

    2014-01-01

    The text of the Agreement between the International Atomic Energy Agency, the Government of Jamaica and the Government of the United States of America for Assistance in Securing Low Enriched Uranium for a Research Reactor is reproduced in this document for the information of all Members of the Agency. The Agency's Board of Governors approved the text of the Agreement on 6 March 2013. The Agreement was signed by the authorized representatives of Jamaica on 25 November 2013, the United States on 2 May 2013 and the Director General of the IAEA on 16 December 2013. Pursuant to the Article XI of the Agreement, the Agreement entered into force on 16 December 2013, upon signature by the Director General of the IAEA and by the authorized representatives of Jamaica and the United States [es

  4. Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

    International Nuclear Information System (INIS)

    Sinha, V.P.; Prasad, G.J.; Hegde, P.V.; Keswani, R.; Basak, C.B.; Pal, S.; Mishra, G.P.

    2009-01-01

    Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and γ-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes

  5. Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, V.P. [Metallic Fuels Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)], E-mail: vedsinha@barc.gov.in; Prasad, G.J.; Hegde, P.V.; Keswani, R.; Basak, C.B.; Pal, S.; Mishra, G.P. [Metallic Fuels Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2009-04-03

    Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and {gamma}-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes.

  6. Recommendations to the NRC on acceptable standard format and content for the Fundamental Nuclear Material Control (FNMC) Plan required for low-enriched uranium enrichment facilities

    International Nuclear Information System (INIS)

    Moran, B.W.; Belew, W.L.; Hammond, G.A.; Brenner, L.M.

    1991-11-01

    A new section, 10 CFR 74.33, has been added to the material control and accounting (MC ampersand A) requirements of 10 CFR Part 74. This new section pertains to US Nuclear Regulatory Commission (NRC)-licensed uranium enrichment facilities that are authorized to produce and to possess more than one effective kilogram of special nuclear material (SNM) of low strategic significance. The new section is patterned after 10 CFR 74.31, which pertains to NRC licensees (other than production or utilization facilities licensed pursuant to 10 CFR Part 50 and 70 and waste disposal facilities) that are authorized to possess and use more than one effective kilogram of unencapsulated SNM of low strategic significance. Because enrichment facilities have the potential capability of producing SNM of moderate strategic significance and also strategic SNM, certain performance objectives and MC ampersand A system capabilities are required in 10 CFR 74.33 that are not contained in 10 CFR 74.31. This document recommends to the NRC information that the licensee or applicant should provide in the fundamental nuclear material control (FNMC) plan. This document also describes methods that should be acceptable for compliance with the general performance objectives. While this document is intended to cover various uranium enrichment technologies, the primary focus at this time is gas centrifuge and gaseous diffusion

  7. Licensed fuel facility status report

    International Nuclear Information System (INIS)

    1990-04-01

    NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related NRC investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  8. Licensed fuel facility status report

    International Nuclear Information System (INIS)

    Joy, D.; Brown, C.

    1993-04-01

    NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related NRC investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  9. Research Establishment progress report 1978 - uranium fuel cycle

    International Nuclear Information System (INIS)

    1978-12-01

    A report of research programs continuing in the following areas is presented: mining and treatment of uranium ores, uranium enrichment, waste treatment, reprocessing and the uranium fuel cycle. Staff responsible for each project are indicated

  10. Fossile fuel and uranium resources

    International Nuclear Information System (INIS)

    Gorkum, A.A. van.

    1975-01-01

    The world's resources of coal, lignite, oil, natural gas, shale oil and uranium are reviewed. These quantities depend on the prices which make new resources exploitable. Uranium resources are given exclusively for the USSR, Eastern Europe and China. Their value in terms of energy depends heavily on the reactor type used. All figures given are estimated to be conservative

  11. Uranium: the nuclear fuel. [Canada

    Energy Technology Data Exchange (ETDEWEB)

    Smith, E E.N. [Eldorado Nuclear Ltd., Ottawa, Ontario (Canada)

    1976-05-01

    A brief history is presented of Canadian uranium exploration, production, and sales. Statistics show that Canada is a good customer for its own uranium due to a rapidly expanding nuclear power program. Due to an average 10 year lag between commencement of exploration and production, and with current producers sold out through 1985, it is imperative that exploration efforts be increased.

  12. Status of development and irradiation performance of advanced proliferation resistant MTR fuel at NUKEM

    International Nuclear Information System (INIS)

    Hrovat, M.; Hassel, H.-W.; Wehner, E.

    1985-01-01

    This paper describes the current status of development and irradiation performance of fuel elements for Material Test and Research (MTR) Reactors with Medium Enriched Uranium (MEU, ≤ 45 % 235-U) and Low Enriched Uranium (LEU, ≤ 20 % 235-U). (author)

  13. Postirradiation examination of high-U-loaded, low-enriched U3O8, UAl2, and U3Si test fuel plates

    International Nuclear Information System (INIS)

    Gomez, J.; Morando, R.; Perez, E.E.; Giorsetti, D.R.; Copeland, G.L.; Hofman, G.L.; Snelgrove, J.L.

    1985-01-01

    The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded, low-enriched U 3 O 8 , UAl 2 and U 3 Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examined, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the RERTR Program. Postirradiation examination of these plates showed satisfactory performance for the oxides, aluminides and silicides (except for the highest-loaded U 3 Si plate) with the only indication of detrimental behavior being the slight bowing of some plates at about 80% burnup

  14. Postirradiation examination of high-U-loaded, low-enriched U3O8, UAl2, and U3Si test fuel plates

    International Nuclear Information System (INIS)

    Gomez, J.; Morando, R.; Perez, E.E.; Giorsetti, D.R.; Copeland, G.L.; Hofman, G.L.; Snelgrove, J.L.

    1985-01-01

    The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded, low-enriched U 3 O 8 , UAl 2 and U 3 Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examined, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the REM Program. Postirradiation examination of these plates showed satisfactory performance for the oxides, aluminides and silicides (except for the highest-loaded U 3 Si plate) with the only indication of detrimental behavior being the slight bowing of some plates at about 80% burnup. (author)

  15. Postirradiation examination of high-U-loaded low-enriched U3O8, UAl2, and U3Si test fuel plates

    International Nuclear Information System (INIS)

    Gomez, J.; Morando, R.; Perez, E.E.; Giorsetti, D.R.; Copeland, G.L.; Hofmann, G.; Snelgrove, J.L.

    1984-01-01

    The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded low-enriched U 3 O 8 , UAl 2 and U 3 Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examinated, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the RERTR Program. Postirradiation examination of these plates showed satisfactory poerformance for the oxides, aluminides and silicides (except for the highest-loaded U 3 Si plate) with the only indication of detrimental behavior during the slight bowing of some plates at about 80% burnup

  16. Development of low enriched uranium target plates by thermo-mechanical processing of UAl{sub 2}–Al matrix for production of {sup 99}Mo in Pakistan

    Energy Technology Data Exchange (ETDEWEB)

    Ali, Kanwar Liaqat; Khan, Akhlaque Ahmad [Pakistan Institute of Nuclear Science and Technology (PINSTECH) P.O Nilore, Islamabad (Pakistan); Mushtaq, Ahmad, E-mail: amushtaq1@hotmail.com [Pakistan Institute of Nuclear Science and Technology (PINSTECH) P.O Nilore, Islamabad (Pakistan); Imtiaz, Farhan; Ziai, Maratab Ali; Gulzar, Amir; Farooq, Muhammad; Hussain, Nazar; Ahmed, Nisar; Pervez, Shahid; Zaidi, Jamshed Hussain [Pakistan Institute of Nuclear Science and Technology (PINSTECH) P.O Nilore, Islamabad (Pakistan)

    2013-02-15

    Uranium aluminide predominated with UAl{sub 2} phase was prepared by arc-melting procedures and comminuted to required particle size. UAl{sub 2} and Al powders were blended and compacted to achieve LEU fuel density of 2.17 g/cm{sup 3}. The picture-frame technique was used to clad the dispersions (UAl{sub 2}–Al) with aluminum. A few target plates were fabricated by thermo-mechanical processing (hot rolling and annealing) of UAl{sub 2}–Al matrix contained in roll billet of Al. The fabricated plates were characterized by destructive and some of non-destructive testing techniques and then annealed to achieve required phase of uranium aluminide for proper dissolution in basic media.

  17. Uranium Fuel Plant. Applicants environmental report

    International Nuclear Information System (INIS)

    1975-05-01

    The Uranium Fuel Plant, located at the Cimarron Facility, was constructed in 1964 with operations commencing in 1965 in accordance with License No. SNM-928, Docket No. 70-925. The plant has been in continuous operation since the issuance of the initial license and currently possesses contracts extending through 1978, for the production of nuclear fuels. The Uranium Plant is operated in conjunction with the Plutonium Facility, each sharing common utilities and sanitary wastes disposal systems. The operation has had little or no detrimental ecological impact on the area. For the operation of the Uranium Fuel Fabrication Plant, initial equipment provided for the production of UO 2 , UF 4 , uranium metal and recovery of scrap materials. In 1968, the plant was expanded by increasing the UO 2 and pellet facilities by the installation of another complete production line for the production of fuel pellets. In 1969, fabrication facilities were added for the production of fuel elements. Equipment initially installed for the recovery of fully enriched scrap has not been used since the last work was done in 1970. Economically, the plant has benefited the Logan County area, with approximately 104 new jobs with an annual payroll of approximately $1.3 million. In addition, $142,000 is annually paid in taxes to state, local and federal governments, and local purchases amount to approximately $1.3 million. This was all in land that was previously used for pasture land, with a maximum value of approximately 37,000 dollars. Environmental effects of plant operation have been minimal. A monitoring and measurement program is maintained in order to ensure that the ecology of the immediate area is not affected by plant operations

  18. Vapor corrosion of aluminum cladding alloys and aluminum-uranium fuel materials in storage environments

    International Nuclear Information System (INIS)

    Lam, P.; Sindelar, R.L.; Peacock, H.B. Jr.

    1997-04-01

    An experimental investigation of the effects of vapor environments on the corrosion of aluminum spent nuclear fuel (A1 SNF) has been performed. Aluminum cladding alloys and aluminum-uranium fuel alloys have been exposed to environments of air/water vapor/ionizing radiation and characterized for applications to degradation mode analysis for interim dry and repository storage systems. Models have been developed to allow predictions of the corrosion response under conditions of unlimited corrodant species. Threshold levels of water vapor under which corrosion does not occur have been identified through tests under conditions of limited corrodant species. Coupons of aluminum 1100, 5052, and 6061, the US equivalent of cladding alloys used to manufacture foreign research reactor fuels, and several aluminum-uranium alloys (aluminum-10, 18, and 33 wt% uranium) were exposed to various controlled vapor environments in air within the following ranges of conditions: Temperature -- 80 to 200 C; Relative Humidity -- 0 to 100% using atmospheric condensate water and using added nitric acid to simulate radiolysis effects; and Gamma Radiation -- none and 1.8 x 10 6 R/hr. The results of this work are part of the body of information needed for understanding the degradation of the A1 SNF waste form in a direct disposal system in the federal repository. It will provide the basis for data input to the ongoing performance assessment and criticality safety analyses. Additional testing of uranium-aluminum fuel materials at uranium contents typical of high enriched and low enriched fuels is being initiated to provide the data needed for the development of empirical models

  19. Corrosion testing of uranium silicide fuel specimens

    International Nuclear Information System (INIS)

    Bourns, W.T.

    1968-09-01

    U 3 Si is the most promising high density natural uranium fuel for water-cooled power reactors. Power reactors fuelled with this material are expected to produce cheaper electricity than those fuelled with uranium dioxide. Corrosion tests in 300 o C water preceded extensive in-reactor performance tests of fuel elements and bundles. Proper heat-treatment of U-3.9 wt% Si gives a U 3 5i specimen which corrodes at less than 2 mg/cm 2 h in 300 o C water. This is an order of magnitude lower than the maximum corrosion rate tolerable in a water-cooled reactor. U 3 Si in a defected unbonded Zircaloy-2 sheath showed only a slow uniform sheath expansion in 300 o C water. All tests were done under isothermal conditions in an out-reactor loop. (author)

  20. Corrosion testing of uranium silicide fuel specimens

    Energy Technology Data Exchange (ETDEWEB)

    Bourns, W T

    1968-09-15

    U{sub 3}Si is the most promising high density natural uranium fuel for water-cooled power reactors. Power reactors fuelled with this material are expected to produce cheaper electricity than those fuelled with uranium dioxide. Corrosion tests in 300{sup o}C water preceded extensive in-reactor performance tests of fuel elements and bundles. Proper heat-treatment of U-3.9 wt% Si gives a U{sub 3}5i specimen which corrodes at less than 2 mg/cm{sup 2} h in 300{sup o}C water. This is an order of magnitude lower than the maximum corrosion rate tolerable in a water-cooled reactor. U{sub 3}Si in a defected unbonded Zircaloy-2 sheath showed only a slow uniform sheath expansion in 300{sup o}C water. All tests were done under isothermal conditions in an out-reactor loop. (author)

  1. Present state and problems of uranium fuel fabrication businesses

    International Nuclear Information System (INIS)

    Yuki, Akio

    1981-01-01

    The businesses of uranium fuel fabrication converting uranium hexafluoride to uranium dioxide powder and forming fuel assemblies are the field of most advanced industrialization among nuclear fuel cycle industries in Japan. At present, five plants of four companies engage in this business, and their yearly sales exceeded 20 billion yen. All companies are planning the augmentation of installation capacity to meet the growth of nuclear power generation. The companies of uranium fuel fabrication make the nuclear fuel of the specifications specified by reactor manufacturers as the subcontractors. In addition to initially loaded fuel, the fuel for replacement is required, therefore the demand of uranium fuel is relatively stable. As for the safety of enriched uranium flowing through the farbicating processes, the prevention of inhaling uranium powder by workers and the precaution against criticality are necessary. Also the safeguard measures are imposed so as not to convert enriched uranium to other purposes than peacefull ones. The strict quality control and many times of inspections are carried out to insure the soundness of nuclear fuel. The growth of the business of uranium fuel fabrication and the regulation of the businesses by laws are described. As the problems for the future, the reduction of fabrication cost, the promotion of research and development and others are pointed out. (Kako, I.)

  2. Possibilities for recycling of weapon-grade uranium and plutonium and its peaceful use as reactor fuel

    International Nuclear Information System (INIS)

    Floeter, W.

    2000-01-01

    At present 90% of the energy production is based on fossil fuels. Since March 1999, however, the peaceful use of weapon-grade uranium as reactor fuel is being discussed politically. Partners of this discussion is a group of some private western companies on one side and a state-owned company of the Russian Federation (GUS) on the other. Main topic of the deal besides the winning of electrical energy is the useful disposal of the surplus on weapon-grade material of both leading nations. According to the deal, about 160,000 t of Russian uranium, expressed as natural uranium U 3 O 8 , would be processed during the next 15 years. Proven processes would be applied. Those methods are being already used in Russian facilities at low capacity rates. There are shortages in the production of low enriched uranium (LEU), because of the low capacity rates in the old facilities. The capacity should be increased by a factor of ten, but there is not enough money available in Russia for financing the remodeling of the plants. Financing should therefore probably be provided by the western clients of this deal. The limited amount of uranium produced could be furnised to the uranium market without major difficulties for the present suppliers of natural uranium. The discussions regarding the security of the details of the deal - however - are not yet finalized. (orig.) [de

  3. Processing of irradiated, enriched uranium fuels at the Savannah River Plant

    Energy Technology Data Exchange (ETDEWEB)

    Hyder, M L; Perkins, W C; Thompson, M C; Burney, G A; Russell, E R; Holcomb, H P; Landon, L F

    1979-04-01

    Uranium fuels containing /sup 235/U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of /sup 238/Pu is high enough to make its recovery desirable. Most of the /sup 238/Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, /sup 239/Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse.

  4. Processing of irradiated, enriched uranium fuels at the Savannah River Plant

    International Nuclear Information System (INIS)

    Hyder, M.L.; Perkins, W.C.; Thompson, M.C.; Burney, G.A.; Russell, E.R.; Holcomb, H.P.; Landon, L.F.

    1979-04-01

    Uranium fuels containing 235 U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of 238 Pu is high enough to make its recovery desirable. Most of the 238 Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, 239 Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse

  5. Potential health hazard of nuclear fuel waste and uranium ore

    International Nuclear Information System (INIS)

    Mehta, K.; Sherman, G.R.; King, S.G.

    1991-06-01

    The variation of the radioactivity of nuclear fuel waste (used fuel and fuel reprocessing waste) with time, and the potential health hazard (or inherent radiotoxicity) resulting from its ingestion are estimated for CANDU (Canada Deuterium Uranium) natural-uranium reactors. Four groups of radionuclides in the nuclear fuel waste are considered: actinides, fission products, activation products of zircaloy, and activation products of fuel impurities. Contributions from each of these groups to the radioactivity and to the potential health hazard are compared and discussed. The potential health hazard resulting from used fuel is then compared with that of uranium ore, mine tailings and refined uranium (fresh fuel) on the basis of equivalent amounts of uranium. The computer code HAZARD, specifically developed for these computations, is described

  6. Status report on the irradiation testing and post-irradiation examination of low-enriched U3O8-Al and UAlsub(x)-Al fuel element by the Netherlands Energy Research Foundation (ECN)

    International Nuclear Information System (INIS)

    Pruimboom, H.; Lijbrink, E.; Otterdijk, K. von; Swanenburg de Veye, R.J.

    1984-01-01

    Within the framework of the RERTR-programme four low-enriched (20%) MTR-type fuel elements have been irradiated in the High Flux Reactor at Petten (The Netherlands) and are presently subjected to postirradiation examination. Two of the elements contain UAlsub(x)-Al and two contain U 3 O 8 -Al fuel. The test irradiation has been completed up to the target burn-up values of 50% and 75% respectively. An extensive surveillance programme carried out during the test period has confirmed the excellent in-reactor behaviour of both types. Post-irradiation examination of the 50% burn-up test elements, comprising of dimensional measurements, burn-up determination, fuel metallography and blister testing, has sofar confirmed the irradiation experiences. Good agreement between calculated and measured power and burn-up characteristics has been found. A survey of the test element characteristics, their irradiation history, the irradiation tests and the preliminary PIE results is given in the paper. (author)

  7. Uranium

    International Nuclear Information System (INIS)

    Hamdoun, N.A.

    2007-01-01

    The article includes a historical preface about uranium, discovery of portability of sequential fission of uranium, uranium existence, basic raw materials, secondary raw materials, uranium's physical and chemical properties, uranium extraction, nuclear fuel cycle, logistics and estimation of the amount of uranium reserves, producing countries of concentrated uranium oxides and percentage of the world's total production, civilian and military uses of uranium. The use of depleted uranium in the Gulf War, the Balkans and Iraq has caused political and environmental effects which are complex, raising problems and questions about the effects that nuclear compounds left on human health and environment.

  8. Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

    Energy Technology Data Exchange (ETDEWEB)

    K. J. Allen; I. Bolshinsky; L. L. Biro; M. E. Budu; N. V. Zamfir; M. Dragusin

    2010-07-01

    Romania safely air shipped 23.7 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel from the VVR S research reactor at Magurele, Romania, to the Russian Federation in June 2009. This was the world’s first air shipment of spent nuclear fuel transported in a Type B(U) cask under existing international laws without special exceptions for the air transport licenses. This shipment was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in cooperation with the Romania National Commission for Nuclear Activities Control (CNCAN), the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), and the Russian Federation State Corporation Rosatom. The shipment was transported by truck to and from the respective commercial airports in Romania and the Russian Federation and stored at a secure nuclear facility in Russia where it will be converted into low enriched uranium. With this shipment, Romania became the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the work, equipment, and approvals that were required to complete this spent fuel air shipment.

  9. Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

    International Nuclear Information System (INIS)

    Allen, K.J.; Bolshinsky, I.; Biro, L.L.; Budu, M.E.; Zamfir, N.V.; Dragusin, M.

    2010-01-01

    Romania safely air shipped 23.7 kilograms of Russian-origin highly enriched uranium (HEU) spent nuclear fuel from the VVR-S research reactor at Magurele, Romania, to the Russian Federation in June 2009. This was the world's first air shipment of spent nuclear fuel transported in a Type B(U) cask under existing international laws without special exceptions for the air transport licenses. This shipment was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in cooperation with the Romania National Commission for Nuclear Activities Control (CNCAN), the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), and the Russian Federation State Corporation Rosatom. The shipment was transported by truck to and from the respective commercial airports in Romania and the Russian Federation and stored at a secure nuclear facility in Russia where it will be converted into low enriched uranium. With this shipment, Romania became the 3. country under the RRRFR program and the 14. country under the GTRI program to remove all HEU. This paper describes the work, equipment, and approvals that were required to complete this spent fuel air shipment. (authors)

  10. Loading ion exchange resins with uranium for HTGR fuel kernels

    International Nuclear Information System (INIS)

    Notz, K.J.; Greene, C.W.

    1976-12-01

    Uranium-loaded ion exchange beads provide an excellent starting material in the production of uranium carbide microspheres for nuclear fuel applications. Both strong-acid (sulfonate) and weak-acid (carboxylate) resins can be fully loaded with uranium from a uranyl nitrate solution utilizing either a batch method or a continuous column technique

  11. Research on using depleted uranium as nuclear fuel for HWR

    International Nuclear Information System (INIS)

    Zhang Jiahua; Chen Zhicheng; Bao Borong

    1999-01-01

    The purpose of our work is to find a way for application of depleted uranium in CANDU reactor by using MOX nuclear fuel of depleted U and Pu instead of natural uranium. From preliminary evaluation and calculation, it was shown that MOX nuclear fuel consisting of depleted uranium enrichment tailings (0.25% 235 U) and plutonium (their ratio 99.5%:0.5%) could replace natural uranium in CANDU reactor to sustain chain reaction. The prospects of application of depleted uranium in nuclear energy field are also discussed

  12. Use of enriched uranium as a fuel in CANDU reactors

    International Nuclear Information System (INIS)

    Zech, H.J.

    1976-08-01

    The use of slightly enriched uranium as a fuel in CANDU-reactors is studied in a simple parametric way. The results show the possibility of 1) about 30% savings in natural uranium consumption 2) about 35% increase in the utilization of the natural uranium 3) a decrease in fuelling costs to about 70 - 80% of the normal case of natural uranium fuelling. (orig.) [de

  13. Licensed fuel facility status report: Inventory difference data, July 1, 1990--June 30, 1991

    International Nuclear Information System (INIS)

    1992-03-01

    NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related NRC investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  14. Licensed fuel facility status report: Inventory difference data, July 1987-December 1987

    International Nuclear Information System (INIS)

    1988-09-01

    NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  15. Licensed fuel facility status report. Inventory difference data, January-June 1985. Volume 6, No. 1

    International Nuclear Information System (INIS)

    1986-02-01

    NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  16. Licensed-fuel-facility status report: inventory difference data, July 1981-December 1981

    International Nuclear Information System (INIS)

    1982-10-01

    NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  17. Licensed fuel facility status report. Inventory difference data, July-December 1985. Volume 6, No. 2

    International Nuclear Information System (INIS)

    1986-08-01

    NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  18. Licensed fuel facility status report. Inventory difference data, January-June 1983. Volume 4, No. 1

    International Nuclear Information System (INIS)

    1984-03-01

    NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, or uranium-233

  19. Licensed fuel facility status report. Inventory difference data, January-June 1982

    International Nuclear Information System (INIS)

    1983-02-01

    NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  20. Licensed fuel facility status report. Inventory difference data, July 1983-December 1983. Volume 4, No. 2

    International Nuclear Information System (INIS)

    1984-08-01

    NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  1. Licensed fuel facility status report. Inventory difference data, January-June 1984. Volume 5, No. 1

    International Nuclear Information System (INIS)

    1985-04-01

    NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or Uranium-233

  2. Licensed fuel facility status report: Inventory difference data, January 1986-June 1986

    International Nuclear Information System (INIS)

    1987-02-01

    NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  3. Licensed fuel facility status report: Inventory difference data, January 1987-June 1987

    International Nuclear Information System (INIS)

    1988-03-01

    NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  4. Licensed fuel facility status report. Volume 5, No. 2. Inventory difference data, July 1984-December 1984

    International Nuclear Information System (INIS)

    1985-10-01

    NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  5. Conversion and start up of Tehran Research Reactor with LEU fuel

    International Nuclear Information System (INIS)

    Zaker, M.

    2004-01-01

    The MW Tehran Research Reactor, Highly Enriched Uranium (HEU) fuel has been converted to Low Enriched Uranium (LEU) fuel using U 3 0 8 -Al with less than 20% enriched uranium. Measured value of excess reactivity, control rod worth and other parameters indicate good agreement with computational predictions. (author)

  6. Effect of kinetic parameters on simultaneous ramp reactivity insertion plus beam tube flooding accident in a typical low enriched U{sub 3}Si{sub 2}-Al fuel-based material testing reactor-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nasir, Rubina; Mirza, Nasir M. [Dept. of, Physics, Air University, Islamabad (Pakistan); Mirza, Sikander M. [Dept. of, Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences, Post Office Nilore, Islamabad (Pakistan)

    2017-06-15

    This work looks at the effect of changes in kinetic parameters on simultaneous reactivity insertions and beam tube flooding in a typical material testing reactor-type research reactor with low enriched high density (U{sub 3}Si{sub 2}-Al) fuel. Using a modified PARET code, various ramp reactivity insertions (from $0.1/0.5 s to $1.3/0.5 s) plus beam tube flooding ($0.5/0.25 s) accidents under uncontrolled conditions were analyzed to find their effects on peak power, net reactivity, and temperature. Then, the effects of changes in kinetic parameters including the Doppler coefficient, prompt neutron lifetime, and delayed neutron fractions on simultaneous reactivity insertion and beam tube flooding accidents were analyzed. Results show that the power peak values are significantly sensitive to the Doppler coefficient of the system in coupled accidents. The material testing reactor-type system under such a coupled accident is not very sensitive to changes in the prompt neutron life time; the core under such a coupled transient is not very sensitive to changes in the effective delayed neutron fraction.

  7. Security of supply of uranium as nuclear fuel

    International Nuclear Information System (INIS)

    Guzman Gomez-Selles, L.

    2011-01-01

    When we talk about Sustainability related to nuclear fuel, the first concern that comes to our mind is about the possibility of having guarantees on the uranium supply for a sufficient period of time. In this paper we are going to analyze the last Reserves data published by the OCD's Red Book and also how the Reserve concept in fully linked to the uranium price. Additionally, it is demonstrated how the uranium Security of supply is guaranteed for, at least, the next 100 years. finally, some comments are made regarding other sources of nuclear fuel as it is the uranium coming from the phosphates or the thorium. (Author)

  8. Licensed-fuel-facility status report: inventory difference data, July 1982-December 1982. Vol. 3, No. 2

    International Nuclear Information System (INIS)

    1983-07-01

    NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  9. Critical review of uranium resources and production capability to 2020

    International Nuclear Information System (INIS)

    Underhill, D.H.

    2002-01-01

    Even with a modest forecast of nuclear power growth for the next 25 years, it is expected that the world uranium requirements will increase. This analysis indicates uranium mine production will continue to be the primary supply of requirements through 2020. Secondary supplies, such as low enriched uranium blended from highly enriched uranium, reprocessing of spent fuel would have to make-up the remaining balance, although the contribution of US and Russian strategic stockpiles is not well known at this time. (author)

  10. The uranium-plutonium breeder reactor fuel cycle

    International Nuclear Information System (INIS)

    Salmon, A.; Allardice, R.H.

    1979-01-01

    All power-producing systems have an associated fuel cycle covering the history of the fuel from its source to its eventual sink. Most, if not all, of the processes of extraction, preparation, generation, reprocessing, waste treatment and transportation are involved. With thermal nuclear reactors more than one fuel cycle is possible, however it is probable that the uranium-plutonium fuel cycle will become predominant; in this cycle the fuel is mined, usually enriched, fabricated, used and then reprocessed. The useful components of the fuel, the uranium and the plutonium, are then available for further use, the waste products are treated and disposed of safely. This particular thermal reactor fuel cycle is essential if the fast breeder reactor (FBR) using plutonium as its major fuel is to be used in a power-producing system, because it provides the necessary initial plutonium to get the system started. In this paper the authors only consider the FBR using plutonium as its major fuel, at present it is the type envisaged in all, current national plans for FBR power systems. The corresponding fuel cycle, the uranium-plutonium breeder reactor fuel cycle, is basically the same as the thermal reactor fuel cycle - the fuel is used and then reprocessed to separate the useful components from the waste products, the useful uranium and plutonium are used again and the waste disposed of safely. However the details of the cycle are significantly different from those of the thermal reactor cycle. (Auth.)

  11. RECOVERY OF URANIUM FROM ZIRCONIUM-URANIUM NUCLEAR FUELS

    Science.gov (United States)

    Gens, T.A.

    1962-07-10

    An improvement was made in a process of recovering uranium from a uranium-zirconium composition which was hydrochlorinated with gsseous hydrogen chloride at a temperature of from 350 to 800 deg C resulting in volatilization of the zirconium, as zirconium tetrachloride, and the formation of a uranium containing nitric acid insoluble residue. The improvement consists of reacting the nitric acid insoluble hydrochlorination residue with gaseous carbon tetrachloride at a temperature in the range 550 to 600 deg C, and thereafter recovering the resulting uranium chloride vapors. (AEC)

  12. Cost update: Technology, safety, and costs of decommissioning a reference uranium fuel fabrication plant

    International Nuclear Information System (INIS)

    Miles, T.L.; Liu, Y.

    1994-06-01

    The cost estimates originally developed in NUREG/CR-1266 for commissioning a reference low-enrichment uranium fuel fabrication plant are updated from 1978 to early 1993 dollars. During this time, the costs for labor and materials increased approximately at the rate of inflation, the cost of energy increased more slowly than the rate of inflation, and the cost of low-level radioactive waste disposal increased much more rapidly than the rate of inflation. The results of the analysis indicate that the estimated costs for the immediate dismantlement and decontamination for unrestricted facility release (DECON) of the reference plant have increased from the mid-1978 value of $3.57 million to $8.08 million in 1993 with in-compact low-level radioactive waste disposal at the US Ecoloay facility near Richland, Washington. The cost estimate rises to $19.62 million with out-of-compact radioactive waste disposal at the Chem-Nuclear facility near Barnwell, South Carolina. A methodology and a formula are presented for estimating the cost of decommissioning the reference uranium fuel fabrication plant at some future time, based on these early 1993 cost estimates. The formula contains essentially the same elements as the formula given in 10 CFR 50.75 for escalating the decommissioning costs for nuclear power reactors to some future time

  13. Uranium silicide activities at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Noel, W.W.; Freim, J.B.

    1983-01-01

    Babcock and Wilcox, Naval Nuclear Fuel Division (NNFD) in conjunction with Argonne National Laboratory (ANL) is actively involved in the Reduced Enrichment Research Test Reactor (RERTR) Program to produce low enriched fuel elements for research reactors. B and W and ANL have undertaken a joint effort in which NNFD will fabricate two low enriched uranium (LEU), Oak Ridge Reactor (ORR) elements with uranium silicide fuel furnished by ANL. These elements are being fabricated for irradiation testing at Oak Ridge National Laboratory (ORNL). Concurrently with this program, NNFD is developing and implementing the uranium silicide and uranium aluminide fuel fabrication technology. NNFD is fabricating the uranium silicide ORR elements in a two-phase program, Development and Production. To summarize: 1. Full size fuel plates can be made with U 3 SiAl but the fabricator must prevent oxidation of the compact prior to hot roll bonding; 2. Providing the ANL U 3 Si x irradiation results are successful, NNFD plans to provide two ORR elements during February 1983; 3. NNFD is developing and implementing U 3 Si x and UAI x fuel fabrication technology to be operational in 1983; 4. NNFD can supply U 3 O 8 high enriched uranium (HEU) or low enriched uranium (LEU) research reactor elements; 5. NNFD is capable of providing high quality, cost competitive LEU or HEU research reactor elements to meet the needs of the customer

  14. Natural uranium fueled light water moderated breeding hybrid power reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.; Levin, P.

    The feasibility of fission-fusion hybrid reactors based on breeding light water thermal fission systems is investigated. The emphasis is on fuel-self-sufficient (FSS) hybrid power reactors that are fueled with natural uranium. Other LWHRs considered include FSS-LWHRs that are fueled with spent fuel from LWRs, and LWHRs which are to supplement LWRs to provide a tandem LWR-LWHR power economy that is fuel-self-sufficient

  15. Features of spherical uranium-graphite HTGR fuel elements control

    International Nuclear Information System (INIS)

    Kreindlin, I.I.; Oleynikov, P.P.; Shtan, A.S.

    1985-01-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described

  16. Features of spherical uranium-graphite HTGR fuel elements control

    Energy Technology Data Exchange (ETDEWEB)

    Kreindlin, I I; Oleynikov, P P; Shtan, A S

    1985-07-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described.

  17. Process for preparing sintered uranium dioxide nuclear fuel

    International Nuclear Information System (INIS)

    Carter, R.E.

    1975-01-01

    Uranium dioxide is prepared for use as fuel in nuclear reactors by sintering it to the desired density at a temperature less than 1300 0 C in a chemically controlled gas atmosphere comprised of at least two gases which in equilibrium provide an oxygen partial pressure sufficient to maintain the uranium dioxide composition at an oxygen/uranium ratio of at least 2.005 at the sintering temperature. 7 Claims, No Drawings

  18. Possibilities of using metal uranium fuel in heavy water reactors

    International Nuclear Information System (INIS)

    Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.

    1965-11-01

    There are serious economic reasons for using metal uranium in heavy water reactors, because of its high density, i.e. high conversion factor, and low cost of fuel elements production. Most important disadvantages are swelling at high burnup and corrosion risk. Some design concepts and application of improved uranium obtained by alloying are promising for achievement of satisfactory stability of metal uranium under reactor operation conditions [sr

  19. Measurement of enriched uranium and uranium-aluminum fuel materials with the AWCC

    International Nuclear Information System (INIS)

    Krick, M.S.; Menlove, H.O.; Zick, J.; Ikonomou, P.

    1985-05-01

    The active well coincidence counter (AWCC) was calibrated at the Chalk River Nuclear Laboratories (CRNL) for the assay of 93%-enriched fuel materials in three categories: (1) uranium-aluminum billets, (2) uranium-aluminum fuel elements, and (3) uranium metal pieces. The AWCC was a standard instrument supplied to the International Atomic Energy Agency under the International Safeguards Project Office Task A.51. Excellent agreement was obtained between the CRNL measurements and previous Los Alamos National Laboratory measurements on similar mockup fuel material. Calibration curves were obtained for each sample category. 2 refs., 8 figs., 15 tabs

  20. EPA requirements for the uranium fuel cycle

    International Nuclear Information System (INIS)

    Dunster, H.J.

    1975-01-01

    The draft Environmental Statement issued by the Environmental Protection Agency (EPA) in the United States in preparation for Proposed Rulemaking Action concerning 'Environmental radiation protection requirements for normal operations of activities in the uranium fuel cycle' is summarized and discussed. The standards proposed by the EPA limit the annual dose equivalents to any member of the public, and also the releases of radionuclides to the 'general environment' for each gigawatt year of electrical energy produced. These standards were based on cost effectiveness arguements and levels and correspond to the ICRP recommendation to keep all exposures as low as reasonably achievable, economic and social factors being taken into account. They should be clearly distinguished from dose limits, although the EPA does not make this at all clear. The EPA seems to have shown an unexpected lack of understanding of the recommendations of ICRP Publication 9 (1965) and an apparent unawareness of ICRP Publication 22 (1973), and has therefore wrongly presented the new standards as a significant change in policy. The EPA has reviewed the information on the likely level of dose equivalents to members of the public and the likely cost reductions, thereby quantifying existing principles as applied to the fuel cycle as a whole. The EPA has stated that its proposals could be achieved as a cost in the region of Pound100,000 per death (or major genetic defect). It is pointed out that the EPA's use of the term 'waste' to exclude liquid and gaseous effluents may cause confusion. (U.K.)

  1. Possibility of using metal uranium fuel in heavy water reactors

    International Nuclear Information System (INIS)

    Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.

    1965-01-01

    The review of metal uranium properties including irradiation in the reactor core lead to the following conclusions. Using metal uranium in the heavy water reactors would be favourable from economic point of view for ita high density, i.e. high conversion factor and low cost of fuel elements fabrication. Most important constraint is swelling during burnup and corrosion

  2. The use of medium enriched uranium fuel for research reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The evaluation described in the present paper concerns the use of medium enriched uranium fuel for our research reactors. The underlying assumptions set up for the evaluation are as follows: (1) At first, the use of alternative fuel should not affect, even to a small extent, research and development programs in nuclear energy utilization, which were described in the previous paper. Hence the use of lower enrichment fuel should not cause any reduction in reactor performances. (2) The fuel cycle cost for operating research reactors with alternative fuel, excepting R and D cost for such fuel, should not increase beyond an acceptable limit. (3) The use of alternative fuel should be satisfactory with respect to non-proliferation purposes, to the almost same degree as the use of 20% enriched uranium fuel

  3. Natural uranium metallic fuel elements: fabrication and operating experience

    International Nuclear Information System (INIS)

    Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.

    1980-01-01

    The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)

  4. Once-through uranium thorium fuel cycle in CANDU reactors

    International Nuclear Information System (INIS)

    Ozdemir, S.; Cubukcu, E.

    2000-01-01

    In this study, the performance of the once-through uranium-thorium fuel cycle in CANDU reactors is investigated. (Th-U)O 2 is used as fuel in all fuel rod clusters where Th and U are mixed homogeneously. CANDU reactors have the advantage of being capable of employing various fuel cycle options because of its good neutron economy, continuous on line refueling ability and axial fuel replacement possibility. For lattice cell calculations transport code WIMS is used. WIMS cross-section library is modified to achieve precise lattice cell calculations. For various enrichments and Th-U mixtures, criticality, heavy element composition changes, diffusion coefficients and cross-sections are calculate. Reactor core is modeled by using the diffusion code CITATION. We conclude that an overall saving of 22% in natural uranium demand can be achieved with the use of Th cycle. However, slightly enriched U cycle still consumes less natural Uranium and is a lot less complicated. (author)

  5. Irradiation behavior of miniature experimental uranium silicide fuel plates

    International Nuclear Information System (INIS)

    Hofman, G.L.; Neimark, L.A.; Mattas, R.F.

    1983-01-01

    Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk from, on the order of 7 x 10 20 cm -3 , far short of the approximately 20 x 10 20 cm -3 goal established for the RERTR program. The purpose of the irradiation experiments on silicide fuels on the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix

  6. Fossil fuels, uranium, and the energy crisis

    Energy Technology Data Exchange (ETDEWEB)

    Playford, P E

    1977-01-01

    Relevant data on the world energy picture are presented to indicate present energy sources and resources, especially fossil fuels and the role of uranium in energy production, with some predictions for the future. World energy is presently being derived from petroleum (some 62%), coal (31%), hydropower (6%), and nuclear (1%). The fundamental cause of the present world energy crisis is attributed to the increase in consumption of petroleum over the past 20 yr, compared with the relatively small size and unequal distribution of the world's remaining reserves. The reserves/production ratio for petroleum has fallen steadily from a general level of 60 to 80 yr from 1920 to 1955, to about 31 yr today. New oil is becoming harder and more expensive to find and produce, the size of discoveries is declining. There is no reason to believe that this trend will be substantially altered, and production is expected to begin to decline between 1985 and 1990. Gas resources also are expected to fall short after the mid-1980s. Coal reserves are enormous, but their full utilization is doubtful because of economic and environmental problems. Tar sands and oil shale resources are potentially major sources of oil, and they are expected to become more competitive with petroleum as higher oil prices occur.

  7. Microstructural evolution of a uranium-10 wt.% molybdenum alloy for nuclear reactor fuels

    International Nuclear Information System (INIS)

    Clarke, A.J.; Clarke, K.D.; McCabe, R.J.; Necker, C.T.; Papin, P.A.; Field, R.D.; Kelly, A.M.; Tucker, T.J.; Forsyth, R.T.; Dickerson, P.O.; Foley, J.C.; Swenson, H.; Aikin, R.M.; Dombrowski, D.E.

    2015-01-01

    Low-enriched uranium-10 wt.% molybdenum (LEU-10wt.%Mo) is of interest for the fabrication of monolithic fuels to replace highly-enriched uranium (HEU) dispersion fuels in high performance research and test reactors around the world. In this work, depleted uranium-10 wt.%Mo (DU-10wt.%Mo) is used to simulate the solidification and microstructural evolution of LEU-10wt.%Mo. Electron backscatter diffraction (EBSD) and complementary electron probe microanalysis (EPMA) reveal significant microsegregation present in the metastable γ-phase after solidification. Homogenization is performed at 800 and 1000 °C for times ranging from 1 to 32 h to explore the time–temperature combinations that will reduce the extent of microsegregation, as regions of higher and lower Mo content may influence local mechanical properties and provide preferred regions for γ-phase decomposition. We show for the first time that EBSD can be used to qualitatively assess microstructural evolution in DU-10wt.%Mo after homogenization treatments. Complementary EPMA is used to quantitatively confirm this finding. Homogenization at 1000 °C for 2–4 h may the regions that contain 8 wt.% Mo or lower, whereas homogenization at 1000 °C for longer than 8 h effectively saturates Mo chemical homogeneity, but results in substantial grain growth. The appropriate homogenization time will depend upon additional microstructural considerations, such as grain growth and intended subsequent processing. Higher carbon LEU-10wt.%Mo generally contains more inclusions within the grains and at grain boundaries after solidification. The effect of these inclusions on microstructural evolution (e.g. grain growth) during homogenization and as potential γ-phase decomposition nucleation sites is unclear, but likely requires additional study.

  8. Microstructural evolution of a uranium-10 wt.% molybdenum alloy for nuclear reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    Clarke, A.J., E-mail: aclarke@lanl.gov; Clarke, K.D.; McCabe, R.J.; Necker, C.T.; Papin, P.A.; Field, R.D.; Kelly, A.M.; Tucker, T.J.; Forsyth, R.T.; Dickerson, P.O.; Foley, J.C.; Swenson, H.; Aikin, R.M.; Dombrowski, D.E.

    2015-10-15

    Low-enriched uranium-10 wt.% molybdenum (LEU-10wt.%Mo) is of interest for the fabrication of monolithic fuels to replace highly-enriched uranium (HEU) dispersion fuels in high performance research and test reactors around the world. In this work, depleted uranium-10 wt.%Mo (DU-10wt.%Mo) is used to simulate the solidification and microstructural evolution of LEU-10wt.%Mo. Electron backscatter diffraction (EBSD) and complementary electron probe microanalysis (EPMA) reveal significant microsegregation present in the metastable γ-phase after solidification. Homogenization is performed at 800 and 1000 °C for times ranging from 1 to 32 h to explore the time–temperature combinations that will reduce the extent of microsegregation, as regions of higher and lower Mo content may influence local mechanical properties and provide preferred regions for γ-phase decomposition. We show for the first time that EBSD can be used to qualitatively assess microstructural evolution in DU-10wt.%Mo after homogenization treatments. Complementary EPMA is used to quantitatively confirm this finding. Homogenization at 1000 °C for 2–4 h may the regions that contain 8 wt.% Mo or lower, whereas homogenization at 1000 °C for longer than 8 h effectively saturates Mo chemical homogeneity, but results in substantial grain growth. The appropriate homogenization time will depend upon additional microstructural considerations, such as grain growth and intended subsequent processing. Higher carbon LEU-10wt.%Mo generally contains more inclusions within the grains and at grain boundaries after solidification. The effect of these inclusions on microstructural evolution (e.g. grain growth) during homogenization and as potential γ-phase decomposition nucleation sites is unclear, but likely requires additional study.

  9. Soil Characterization by Large Scale Sampling of Soil Mixed with Buried Construction Debris at a Former Uranium Fuel Fabrication Facility

    International Nuclear Information System (INIS)

    Nardi, A.J.; Lamantia, L.

    2009-01-01

    Recent soil excavation activities on a site identified the presence of buried uranium contaminated building construction debris. The site previously was the location of a low enriched uranium fuel fabrication facility. This resulted in the collection of excavated materials from the two locations where contaminated subsurface debris was identified. The excavated material was temporarily stored in two piles on the site until a determination could be made as to the appropriate disposition of the material. Characterization of the excavated material was undertaken in a manner that involved the collection of large scale samples of the excavated material in 1 cubic meter Super Sacks. Twenty bags were filled with excavated material that consisted of the mixture of both the construction debris and the associated soil. In order to obtain information on the level of activity associated with the construction debris, ten additional bags were filled with construction debris that had been separated, to the extent possible, from the associated soil. Radiological surveys were conducted of the resulting bags of collected materials and the soil associated with the waste mixture. The 30 large samples, collected as bags, were counted using an In-Situ Object Counting System (ISOCS) unit to determine the average concentration of U-235 present in each bag. The soil fraction was sampled by the collection of 40 samples of soil for analysis in an on-site laboratory. A fraction of these samples were also sent to an off-site laboratory for additional analysis. This project provided the necessary soil characterization information to allow consideration of alternate options for disposition of the material. The identified contaminant was verified to be low enriched uranium. Concentrations of uranium in the waste were found to be lower than the calculated site specific derived concentration guideline levels (DCGLs) but higher than the NRC's screening values. The methods and results are presented

  10. Impact of fuel fabrication and fuel management technologies on uranium management

    International Nuclear Information System (INIS)

    Arnsberger, P.L.; Stucker, D.L.

    1994-01-01

    Uranium utilization in commercial pressurized water reactors is a complex function of original NSSS design, utility energy requirements, fuel assembly design, fuel fabrication materials and fuel fabrication materials and fuel management optimization. Fuel design and fabrication technologies have reacted to the resulting market forcing functions with a combination of design and material changes. The technologies employed have included ever-increasing fuel discharge burnup, non-parasitic structural materials, burnable absorbers, and fissile material core zoning schemes (both in the axial and radial direction). The result of these technological advances has improved uranium utilization by roughly sixty percent from the infancy days of nuclear power to present fuel management. Fuel management optimization technologies have also been developed in recent years which provide fuel utilization improvements due to core loading pattern optimization. This paper describes the development and impact of technology advances upon uranium utilization in modern pressurized water reactors. 10 refs., 3 tabs., 10 figs

  11. Turkey's regulatory plans for high enriched to low enriched conversion of TR-2 reactor core

    International Nuclear Information System (INIS)

    Guelol Oezdere, Oya

    2003-01-01

    Turkey is a developing country and has three nuclear facilities two of which are research reactors and one pilot fuel production plant. One of the two research reactors is TR-2 which is located in Cekmece site in Istanbul. TR-2 Reactor's core is composed of both high enriched and low enriched fuel and from high enriched to low enriched core conversion project will take place in year 2005. This paper presents the plans for drafting regulations on the safety analysis report updates for high enriched to low enriched core conversion of TR-2 reactor, the present regulatory structure of Turkey and licensing activities of nuclear facilities. (author)

  12. PHWR fuel fabrication with imported uranium - procedures and processes

    International Nuclear Information System (INIS)

    Rao, R.V.R.L.V.; Rameswara Rao, A.; Hemantha Rao, G.V.S.; Jayaraj, R.N.

    2010-01-01

    Following the 123 agreement and subsequent agreements with IAEA & NSG, Government of India has entered into bilateral agreements with different countries for nuclear trade. Department of Atomic Energy (DAE), Government of India, has entered into contract with few countries for supply of uranium material for use in the safeguarded PHWRs. Nuclear Fuel Complex (NFC), an industrial unit of DAE, established in the early seventies, is engaged in the production of Nuclear Fuel and Zircaloy items required for Nuclear Power Reactors operating in the country. NFC has placed one of its fuel fabrication facilities (NFC, Block-A, INE-) under safeguards. DAE has opted to procure uranium material in the form of ore concentrate and fuel pellets. Uranium ore concentrate was procured as per the ASTM specifications. Since no international standards are available for PHWR fuel pellets, Specifications have to be finalized based on the present fabrication and operating experience. The process steps have to be modified and fine tuned for handling the imported uranium material especially for ore concentrate. Different transportation methods are to be employed for transportation of uranium material to the facility. Cost of the uranium material imported and the recoveries at various stages of fuel fabrication have impact on the fuel pricing and in turn the unit energy costs. Similarly the operating procedures have to be modified for safeguards inspections by IAEA. NFC has successfully manufactured and supplied fuel bundles for the three 220 MWe safeguarded PHWRs. The paper describes various issues encountered while manufacturing fuel bundles with different types of nuclear material. (author)

  13. Advances of the low enriched uranium utilization project in CNA-1 during 1998 and 1999; Avances del proyecto de utilizacion de uranio levemente enriquecido en la CNA-I en 1998 y 1999

    Energy Technology Data Exchange (ETDEWEB)

    Fink, Jose M; Higa, Manabu; Sidelnik, Jorge I [Nucleoelectrica Argentina SA (NASA), Buenos Aires (Argentina); Perez, Ramon A [Nucleoelectrica Argentina SA (NASA), Lima (Argentina). Central Nuclear Atucha I; Casario, Jose A; Alvarez, Luis A [Comision Nacional de Energia Atomica, San Martin (Argentina). Centro Atomico Constituyentes

    1999-07-01

    In this work, a general description of advances of the Enriched Fuel Introduction Project in CNA-1 and the main tasks performed during 1998 and 1999 are presented. The program is being satisfactorily developed and during that period the number of slightly enriched fuels (LEU) introduced had significantly increased in relation to previous years. At present, there are 181 LEU fuel elements in the core and 125 LEU fuel elements have been extracted. The number of full power burnt fuel elements per day decreased from 1.31 FE/dpp in 1994 (when all fuel was natural) to 0.92 in 1998 and 0.83 in 1999, reaching the predicted value for homogeneous LEU core of 0.7. The cost of burnt fuel in 1998 was 25% lower that if only natural fuel would have been used. (author)

  14. Nuclear fuel banks

    International Nuclear Information System (INIS)

    Anon.

    2010-01-01

    In december 2010 IAEA gave its agreement for the creation of a nuclear fuel bank. This bank will allow IAEA to help member countries that renounce to their own uranium enrichment capacities. This bank located on one or several member countries will belong to IAEA and will be managed by IAEA and its reserve of low enriched uranium will be sufficient to fabricate the fuel for the first load of a 1000 MW PWR. Fund raising has been successful and the running of the bank will have no financial impact on the regular budget of the IAEA. Russia has announced the creation of the first nuclear fuel bank. This bank will be located on the Angarsk site (Siberia) and will be managed by IAEA and will own 120 tonnes of low-enriched uranium fuel (between 2 and 4.95%), this kind of fuel is used in most Russian nuclear power plants. (A.C.)

  15. Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

    International Nuclear Information System (INIS)

    Clegg, L.J.; Coady, J.R.

    1977-01-01

    The computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 1 presents these data for unirradiated fuel, uranium ore and uranium mill tailings. In Part 2 they have been computed for fuel irradiated to levels of burnup ranging from 140 GJ/kg U to 1150 GJ/kg U. (author)

  16. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    International Nuclear Information System (INIS)

    Souza, J.A.B.; Durazzo, M.

    2010-01-01

    IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 gU/cm 3 by using the U 3 Si 2 -Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 gU/cm 3 for the U 3 Si 2 -Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian-Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)

  17. Fuel Cycle Impacts of Uranium-Plutonium Co-extraction

    International Nuclear Information System (INIS)

    Taiwo, Temitope; Szakaly, Frank; Kim, Taek-Kyum; Hill, Robert

    2008-01-01

    A systematic investigation of the impacts of uranium and plutonium co-extraction during fuel separations on reactor performance and fuel cycle has been performed. Proliferation indicators, critical mass and radiation source levels of the separation products or fabricated fuel, were also evaluated. Using LWR-spent-uranium-based MOX fuel instead of natural-uranium-based fuel in a PWR MOX core requires a higher initial plutonium content (∼1%), and results in higher Np-237 content (factor of 5) in the spent fuel, and less consumption of Pu-238 (20%) and Am-241 (14%), indicating a reduction in the effective repository space utilization. Additionally, minor actinides continue to accumulate in the fuel cycle, and thus a separate solution is required for them. Differences were found to be quite smaller (∼0.4% in initial transuranics) between the equilibrium cycles of advanced fast reactor cores using spent and depleted uranium for make-up, in additional to transuranics. The critical masses of the co-extraction products were found to be higher than for weapons-grade plutonium (WG-Pu) and the decay heat and radiation sources of the materials (products) were also found to be generally higher than for WG-Pu in the transuranics content range of 10% to 100% in the heavy-metal. (authors)

  18. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Jose Antonio Batista de; Durazzo, Michelangelo, E-mail: jasouza@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 g U/c m3 by using the U{sub 3}Si{sub 2}-Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 g U/c m3 for the U{sub 3}Si{sub 2}-Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian- Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)

  19. Detailed analysis of uranium silicide dispersion fuel swelling

    International Nuclear Information System (INIS)

    Hofmann, G.L.; Ryu, Woo-Seog

    1991-01-01

    Swelling of U 3 Si and U 3 Si 2 is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and micro structural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide dispersion fuel. (orig.)

  20. Detailed analysis of uranium silicide dispersion fuel swelling

    International Nuclear Information System (INIS)

    Hofman, G.L.; Ryu, Woo-Seog.

    1989-01-01

    Swelling of U 3 Si and U 3 Si 2 is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and microstructural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide disperson fuel. 5 refs., 10 figs

  1. Statistical evaluation of an interlaboratory comparison for the determination of uranium by potentiometric titration

    International Nuclear Information System (INIS)

    Ketema, D.J.; Harry, R.J.S.; Zijp, W.L.

    1990-09-01

    Upon request of the ESARDA working group 'Low enriched uranium conversion - and fuel fabrication plants' an interlaboratory comparison was organized, to assess the precision and accuracy concerning the determination of uranium by the potentiometric titration method. This report presents the results of a statistical evaluation on the data of the first phase of this exercise. (author). 9 refs.; 5 figs.; 24 tabs

  2. Development of metal uranium fuel and testing of construction materials (I-VI); Part I

    International Nuclear Information System (INIS)

    Mihajlovic, A.

    1965-11-01

    This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors

  3. Economic analysis of thorium-uranium fuel cycle introduced into PWRs

    International Nuclear Information System (INIS)

    Fan Li; Sun Qian

    2014-01-01

    Using PWR of Daya Bay Unit l as the reference reactor, a validated computer code was used to calculate the fuel cycle costs for uranium fuel cycle and thorium-uranium fuel cycle over the following 20 0perational years respectively. The calculation results show that the thorium-uranium fuel cycle is economically competitive with the uranium fuel cycle when reprocessing mode is adopted. For thorium-uranium fuel cycle, if the price of natural uranium is higher than 120 $ /pound U_3O_8, the fuel cycle cost of the direct disposal mode is greater than that of the reprocessing mode. Therefore, when the uranium price may maintain a high level long-termly, adopting reprocessing mode will benefit the economic advantage for the thorium-uranium fuel cycle introduced into PWRs. (authors)

  4. Uranium for Nuclear Power: Resources, Mining and Transformation to Fuel

    International Nuclear Information System (INIS)

    Hore-Lacy, Ian

    2016-01-01

    Uranium for Nuclear Power: Resources, Mining and Transformation to Fuel discusses the nuclear industry and its dependence on a steady supply of competitively priced uranium as a key factor in its long-term sustainability. A better understanding of uranium ore geology and advances in exploration and mining methods will facilitate the discovery and exploitation of new uranium deposits. The practice of efficient, safe, environmentally-benign exploration, mining and milling technologies, and effective site decommissioning and remediation are also fundamental to the public image of nuclear power. This book provides a comprehensive review of developments in these areas: • Provides researchers in academia and industry with an authoritative overview of the front end of the nuclear fuel cycle • Presents a comprehensive and systematic coverage of geology, mining, and conversion to fuel, alternative fuel sources, and the environmental and social aspects • Written by leading experts in the field of nuclear power, uranium mining, milling, and geological exploration who highlight the best practices needed to ensure environmental safety

  5. Uranium

    International Nuclear Information System (INIS)

    Poty, B.; Cuney, M.; Bruneton, P.; Virlogeux, D.; Capus, G.

    2010-01-01

    With the worldwide revival of nuclear energy comes the question of uranium reserves. For more than 20 years, nuclear energy has been neglected and uranium prospecting has been practically abandoned. Therefore, present day production covers only 70% of needs and stocks are decreasing. Production is to double by 2030 which represents a huge industrial challenge. The FBR-type reactors technology, which allows to consume the whole uranium content of the fuel, is developing in several countries and will ensure the long-term development of nuclear fission. However, the implementation of these reactors (the generation 4) will be progressive during the second half of the 21. century. For this reason an active search for uranium ores will be necessary during the whole 21. century to ensure the fueling of light water reactors which are huge uranium consumers. This dossier covers all the aspects of natural uranium production: mineralogy, geochemistry, types of deposits, world distribution of deposits with a particular attention given to French deposits, the exploitation of which is abandoned today. Finally, exploitation, ore processing and the economical aspects are presented. Contents: 1 - the uranium element and its minerals: from uranium discovery to its industrial utilization, the main uranium minerals (minerals with tetravalent uranium, minerals with hexavalent uranium); 2 - uranium in the Earth's crust and its geochemical properties: distribution (in sedimentary rocks, in magmatic rocks, in metamorphic rocks, in soils and vegetation), geochemistry (uranium solubility and valence in magmas, uranium speciation in aqueous solution, solubility of the main uranium minerals in aqueous solution, uranium mobilization and precipitation); 3 - geology of the main types of uranium deposits: economical criteria for a deposit, structural diversity of deposits, classification, world distribution of deposits, distribution of deposits with time, superficial deposits, uranium

  6. Improving the neutronic characteristics of a boiling water reactor by using uranium zirconium hydride fuel instead of uranium dioxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, Ahmed Abdelghafar [Higher Technological Institute, Ramadan (Egypt)

    2016-06-15

    The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO{sub 2}) and uranium zirconium hydride (UZrH{sub 1.6}) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO{sub 2} contains 8 × 8 fuel rods while that fueled with UZrH{sub 1.6} contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH{sub 1.6} fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

  7. Fuel cycle parameters for strategy studies

    International Nuclear Information System (INIS)

    Archinoff, G.H.

    1979-05-01

    This report summarizes seven fuel cycle parameters (efficiency, specific power, burnup, equilibrium net fissile feed, equilibrium net fissile surplus, first charge fissile content, and whether or not fuel reprocessing is required) to be used in long-term strategy analyses of fuel cycles based on natural UO 2 , low enriched uranium, mixed oxides, plutonium topped thorium, uranium topped thorium, and the fast breeder oxide cycle. (LL)

  8. Choice and utilization of slightly enriched uranium fuel for high performance research reactors

    International Nuclear Information System (INIS)

    Cerles, J.M.; Schwartz, J.P.

    1978-01-01

    Problems relating to the replacement of highly enriched (90% or 93% U 235 ) uranium fuel: by moderately enriched (20% or 40% in U 235 ) metallic uranium fuel and slightly enriched (3% or 8% in U 235 ) uranium oxide fuel are discussed

  9. Separation and recovery method for depleted uranium from spent fuel

    International Nuclear Information System (INIS)

    Imoto, Yoshie; Fujita, Reiko.

    1993-01-01

    Spent oxide fuels are reduced in a molten salt of CaCl 2 -CaF 2 to convert them into metals, then melted in an Fe-U bath disposed in an electrolytic refining vessel and brought into contact with molten Mg, to extract transuranium elements and rare earth elements contained in the Fe-U bath as metals in the molten Mg. Then molten Mg is removed and the residue is brought into contact with KCl-LiCl molten salt and electrolyzed using the Fe-U as an anode. Then, uranium is recovered by deposition on an iron cathode disposed in chloride electrolytes of the electrolytic refining vessel. Uranium and transuranium elements can be thus separated and, for example, depleted uranium for use in blanket fuels can be recovered easily. This can greatly reduce the temporary storage amount of depleted uranium, to eliminate requirement for a large-scaled facility used exclusively for storing uranium and long time management for uranium. (T.M.)

  10. Optimization of fuel cycle strategies with constraints on uranium availability

    International Nuclear Information System (INIS)

    Silvennoinen, P.; Vira, J.; Westerberg, R.

    1982-01-01

    Optimization of nuclear reactor and fuel cycle strategies is studied under the influence of reduced availability of uranium. The analysis is separated in two distinct steps. First, the global situation is considered within given high and low projections of the installed capacity up to the year 2025. Uranium is regarded as an exhaustible resource whose production cost would increase proportionally to increasing cumulative exploitation. Based on the estimates obtained for the uranium cost, a global strategy is derived by splitting the installed capacity between light water reactor (LWR) once-through, LWR recycle, and fast breeder reactor (FBR) alternatives. In the second phase, the nuclear program of an individual utility is optimized within the constraints imposed from the global scenario. Results from the global scenarios indicate that in a reference case the uranium price would triple by the year 2000, and the price escalation would continue throughout the planning period. In a pessimistic growth scenario where the global nuclear capacity would not exceed 600 GW(electric) in 2025, the uranium price would almost double by 2000. In both global scenarios, FBRs would be introduced, in the reference case after 2000 and in the pessimistic case after 2010. In spite of the increases in the uranium prices, the levelized power production cost would increase only by 45% up to 2025 in the utility case provided that the plutonium is incinerated as a substitute fuel

  11. Fuel loading and homogeneity analysis of HFIR design fuel plates loaded with uranium silicide fuel

    International Nuclear Information System (INIS)

    Blumenfeld, P.E.

    1995-08-01

    Twelve nuclear reactor fuel plates were analyzed for fuel loading and fuel loading homogeneity by measuring the attenuation of a collimated X-ray beam as it passed through the plates. The plates were identical to those used by the High Flux Isotope Reactor (HFIR) but were loaded with uranium silicide rather than with HFIR's uranium oxide fuel. Systematic deviations from nominal fuel loading were observed as higher loading near the center of the plates and underloading near the radial edges. These deviations were within those allowed by HFIR specifications. The report begins with a brief background on the thermal-hydraulic uncertainty analysis for the Advanced Neutron Source (ANS) Reactor that motivated a statistical description of fuel loading and homogeneity. The body of the report addresses the homogeneity measurement techniques employed, the numerical correction required to account for a difference in fuel types, and the statistical analysis of the resulting data. This statistical analysis pertains to local variation in fuel loading, as well as to ''hot segment'' analysis of narrow axial regions along the plate and ''hot streak'' analysis, the cumulative effect of hot segment loading variation. The data for all twelve plates were compiled and divided into 20 regions for analysis, with each region represented by a mean and a standard deviation to report percent deviation from nominal fuel loading. The central regions of the plates showed mean values of about +3% deviation, while the edge regions showed mean values of about -7% deviation. The data within these regions roughly approximated random samplings from normal distributions, although the chi-square (χ 2 ) test for goodness of fit to normal distributions was not satisfied

  12. Irradiation performance of uranium-molybdenum alloy dispersion fuels

    International Nuclear Information System (INIS)

    Almeida, Cirila Tacconi de

    2005-01-01

    The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm 3 were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm 3 showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

  13. Criticality safety considerations for MSRE fuel drain tank uranium aggregation

    International Nuclear Information System (INIS)

    Hollenbach, D.F.; Hopper, C.M.

    1997-01-01

    This paper presents the results of a preliminary criticality safety study of some potential effects of uranium reduction and aggregation in the Molten Salt Reactor Experiment (MSRE) fuel drain tanks (FDTs) during salt removal operations. Since the salt was transferred to the FDTs in 1969, radiological and chemical reactions have been converting the uranium and fluorine in the salt to UF 6 and free fluorine. Significant amounts of uranium (at least 3 kg) and fluorine have migrated out of the FDTs and into the off-gas system (OGS) and the auxiliary charcoal bed (ACB). The loss of uranium and fluorine from the salt changes the chemical properties of the salt sufficiently to possibly allow the reduction of the UF 4 in the salt to uranium metal as the salt is remelted prior to removal. It has been postulated that up to 9 kg of the maximum 19.4 kg of uranium in one FDT could be reduced to metal and concentrated. This study shows that criticality becomes a concern when more than 5 kg of uranium concentrates to over 8 wt% of the salt in a favorable geometry

  14. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    International Nuclear Information System (INIS)

    Souza, Jose Antonio Batista de

    2011-01-01

    IPEN-CNEN/SP developed the technology to produce the dispersion type fuel elements for research reactors and made it available for routine production. Today, the fuel produced in IPEN-CNEN/SP is limited to the uranium concentration of 3.0 gU/cm 3 for U 3 Si 2 -Al dispersion-based and 2.3 gU/cm 3 for U 3 O 8 -Al dispersion. The increase of uranium concentration in fuel plates enables the reactivity of the reactor core reactivity to be higher and extends the fuel life. Concerning technology, it is possible to increase the uranium concentration in the fuel meat up to the limit of 4.8 gU/cm 3 in U 3 Si 2 -Al dispersion and 3.2 gU/cm 3 U 3 O 8 -Al dispersion. These dispersions are well qualified worldwide. This work aims to develop the manufacturing process of both fuel meats with high uranium concentrations, by redefining the manufacturing procedures currently adopted in the Nuclear Fuel Center of IPEN-CNEN/SP. Based on the results, it was concluded that to achieve the desired concentration, it is necessary to make some changes in the established procedures, such as in the particle size of the fuel powder and in the feeding process inside the matrix, before briquette pressing. These studies have also shown that the fuel plates, with a high concentration of U 3 Si 2 -Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U 3 O 8 -Al dispersion fuel plates with 3.2 gU/cm 3 showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U 3 Si 2 production at 4.8 gU/cm 3 , with enriched uranium. The produced plates were used to assemble the fuel element IEA-228, which was irradiated in order to check its performance in the IEA-R1 reactor at IPEN-CNEN/SP. These new fuels have potential to be used in the new Brazilian Multipurpose Reactor - RMB. (author)

  15. Uranium savings on a once through PWR fuel cycle

    International Nuclear Information System (INIS)

    Cupo, J.V.

    1980-01-01

    A number of alternatives which have the greatest potential for near term savings with minimum plant and fuel modifications have been examined at Westinghouse as part of continued internal assessment and part of NASAP study conducted for DOE pertaining to uranium utilization in a once through PWR fuel cycle. The alternatives which could be retrofitted to existing reactors were examined in more detail in the evaluation since they would have the greater near term impact on U savings

  16. The high temperature reactor and its fuel cycle options

    International Nuclear Information System (INIS)

    1979-07-01

    The status of the HTR system in the Federal Republic of Germany as well as the consecutive steps and the probable cost of further development are presented. The considerations are based on a recycling Th/highly enriched uranium (HEU) fuel cycle which has been chosen as the main line of the German HTR R and D efforts. Alternative fuel cycles such as medium-enriched uranium (MEU) and low-enriched uranium (LEU) are discussed as well

  17. Assuaging Nuclear Energy Risks: The Angarsk International Uranium Enrichment Center

    International Nuclear Information System (INIS)

    Myers, Astasia

    2011-01-01

    The recent nuclear renaissance has motivated many countries, especially developing nations, to plan and build nuclear power reactors. However, domestic low enriched uranium demands may trigger nations to construct indigenous enrichment facilities, which could be redirected to fabricate high enriched uranium for nuclear weapons. The potential advantages of establishing multinational uranium enrichment sites are numerous including increased low enrichment uranium access with decreased nuclear proliferation risks. While multinational nuclear initiatives have been discussed, Russia is the first nation to actualize this concept with their Angarsk International Uranium Enrichment Center (IUEC). This paper provides an overview of the historical and modern context of the multinational nuclear fuel cycle as well as the evolution of Russia's IUEC, which exemplifies how international fuel cycle cooperation is an alternative to domestic facilities.

  18. DUCTILE URANIUM FUEL FOR NUCLEAR REACTORS AND METHOD OF MAKING

    Science.gov (United States)

    Zegler, S.T.

    1963-11-01

    The fabrication process for a ductile nuclear fuel alloy consisting of uranium, fissium, and from 0.25 to 1.0 wt% of silicon or aluminum or from 0.25 to 2 wt% of titanium or yttrium is presented. (AEC)

  19. Evaluation of plutonium, uranium, and thorium use in power reactor fuel cycles

    International Nuclear Information System (INIS)

    Kasten, P.R.; Homan, F.J.

    1977-01-01

    The increased cost of uranium and separative work has increased the attractiveness of plutonium use in both uranium and thorium fuel cycles in thermal reactors. A technology, fuel utilization, and economic evaluation is given for uranium and thorium fuel cycles in various reactor types, along with the use of plutonium and 238 U. Reactors considered are LWRs, HWRs, LWBRs, HTGRs, and FBRs. Key technology factors are fuel irradiation performance and associated physical property values. Key economic factors are unit costs for fuel fabrication and reprocessing, and for refabrication of recycle fuels; consistent cost estimates are utilized. In thermal reactors, the irradiation performance of ceramic fuels appears to be satisfactory. At present costs for uranium ore and separative work, recycle of plutonium with thorium rather than uranium is preferable from fuel utilization and economic viewpoints. Further, the unit recovery cost of plutonium is lower from LWR fuels than from natural-uranium HWR fuels; use of LWR product permits plutonium/thorium fueling to compete with uranium cycles. Converting uranium cycles to thorium cycles increases the energy which can be extracted from a given uranium resource. Thus, additional fuel utilization improvement can be obtained by fueling all thermal reactors with thorium, but this requires use of highly enriched uranium; use of 235 U with thorium is most economic in HTGRs followed by HWRs and then LWRs. Marked improvement in long-term fuel utilization can be obtained through high thorium loadings and short fuel cycle irradiations as in the LWBR, but this imposes significant economic penalties. Similar operating modes are possible in HWRs and HTGRs. In fast reactors, use of the plutonium-uranium cycle gives advantageous fuel resource utilization in both LMFBRs and GCFRs; use of the thorium cycle provides more negative core reactivity coefficients and more flexibility relative to use of recycle fuels containing uranium of less than 20

  20. Reduction of uranium in disposal conditions of spent nuclear fuel

    International Nuclear Information System (INIS)

    Myllykylae, E.

    2008-02-01

    This literature study is a summary of publications, in which the reduction of uranium by iron has been investigated in anaerobic groundwater conditions or in aqueous solution in general. The basics of the reduction phenomena and the oxidation states, complexes and solubilities of uranium and iron in groundwaters are discussed as an introduction to the subject, as well as, the Finnish disposal concept of spent nuclear fuel. The spent fuel itself mainly (∼96 %) consists of a sparingly soluble uranium(IV) dioxide, UO 2 (s), which is stable phase in the anticipated reducing disposal conditions. If spent fuel gets in contact with groundwater, oxidizing conditions might be induced by the radiolysis of water, or by the intrusion of oxidizing glacial melting water. Under these conditions, the oxidation and dissolution of uranium dioxide to more soluble U(VI) species could occur. This could lead to the mobilization of uranium and other components of spent fuel matrix including fission products and transuranium elements. The reduction of uranium back to oxidation state U(IV) can be considered as a favourable immobilization mechanism in a long-term, leading to precipitation due to the low solubility of U(IV) species. The cast iron insert of the disposal canister and its anaerobic corrosion products are the most important reductants under disposal conditions, but dissolved ferrous iron may also function as reductant. Other iron sources in the buffer or near-field rock, are also considered as possible reductants. The reduction of uranium is a very challenging phenomenon to investigate. The experimental studies need e.g. well-controlled anoxic conditions and measurements of oxidation states. Reduction and other simultaneous phenomena are difficult to distinghuish. The groundwater conditions (pH, Eh and ions) influence on the prevailing complexes of U and Fe and on forming corrosion products of iron and, thus they determine also the redox chemistry. The partial reduction of

  1. Critical experiments on low enriched uranyl nitrate solution with STACY

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori

    1996-01-01

    As the STACY started steady operations, systematic criticality data on low enriched uranyl nitrate solution system could be accumulated. Main experimental parameters for the cylindrical tank of 60 cm in diameter were uranium concentration and the reflector condition. Basic data on a simple geometry will be helpful for the validation of the standard criticality safety codes, and for evaluating the safety margin included in the criticality designs. Experiments on the reactivity effects of structural materials such as borated concrete and polyethylene are on schedule next year as the second series of experiments using 10 wt% enriched uranyl solution. Furthermore, neutron interacting experiments with two slab tanks will be performed to investigate the fundamental properties of neutron interaction effects between core tanks. These data will be useful for making more reasonable calculation models and for evaluating the safety margin in the criticality designs for the multiple unit system. (J.P.N.)

  2. Challenges in the front end of the uranium fuel cycle

    International Nuclear Information System (INIS)

    Seitz, Ken

    2010-01-01

    The long-term fundamentals for nuclear remain strong. Climate change and clean air concerns remain high on the agenda of national energy policies, as both developing and developed economies pursue a strategy of energy diversity and energy security. A global industry of 435 reactors is expected to grow to more than 639 reactors within the next 20 years with the potential for even more rapid expansion. This nuclear generating capacity relies on an international fuel cycle that can ensure stable and secure supply for decades to come. As the first step in the fuel cycle, the uranium industry has received various price signals over the past 5 decades, from the birth of an industry with strong demand and stock pile building and the associated robust pricing and new production stimulation, to an industry in decline and a period marked by liquidation of large inventories, to the recent resurgence of nuclear and the associated uranium price signals. In many ways, understanding the current uranium environment and the outlook for the industry requires some understanding of these phases of nuclear. The global nuclear fleet today needs about 65,000 tonnes of uranium per year to meet reactor feed requirements. Primary production meets about two thirds of this requirement while the remainder is drawn from secondary supply. Secondary supply can essentially be described as stockpiles of previously produced uranium. However, secondary supplies are finite and more primary production will be needed. From a long-term perspective, there is no question that there are sufficient uranium resources to support the nuclear industry for many years to come. The IAEA's 'Red Book' estimates that more than 5 million tonnes of known resources could potentially be developed at today's prices. This is enough to supply the global reactor fleet for almost 80 years at current usage rates. Recently higher uranium prices have resulted in some production increases although the rate of growth has been held

  3. Irradiation behavior of experimental miniature uranium silicide fuel plates

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Neimark, L.A.; Mattas, R.F.

    1983-01-01

    Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk form, on the order of 7 x 10 20 cm -3 , far short of he approximately 20 x 10 20 cm -3 goal established for the RERTR Program. The purpose of the irradiation experiments on silicide fuels in the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix. The first group of experimental 'mini' fuel plates have recently reached the program's goal burnup and are in various stages of examination. Although the results to date indicate some limitations, it appears that within the range of parameters examined thus far the uranium silicide dispersion holds promise for satisfying most of the needs of the RERTR Program. The twelve experimental silicide dispersion fuel plates that were irradiated to approximately their goal exposure show the 30-vol % U 3 Si-Al plates to be in a stage of relatively rapid fission-gas-driven swelling at a fission density of 2 x 10 20 cm -3 . This fuel swelling will likely result in unacceptably large plate-thickness increases. The U 3 Si plates appear to be superior in this respect; however, they, too, are starting to move into the rapid fuel-swelling stage. Analysis of the currently available post irradiation data indicates that a 40-vol % dispersed fuel may offer an acceptable margin to the onset of unstable thickness changes at exposures of 2 x 10 21 fission/cm 3 . The interdiffusion between fuel and matrix

  4. High-uranium-loaded U3O8--Al fuel element development program

    International Nuclear Information System (INIS)

    Martin, M.M.

    1978-01-01

    The High-Uranium-Loaded U 3 O 8 --Al Fuel Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages

  5. Improved locations of reactivity devices in future CANDU reactors fuelled with natural uranium or enriched fuels

    International Nuclear Information System (INIS)

    Boczar, P.G.; Van Dyk, M.T.

    1987-02-01

    A new configuration of reactivity devices is proposed for future CANDU reactors which improves the core characteristics with enriched fuels, while still allowing the use of natural uranium fuel. Physics calculations for this new configuration are presented for four fuel types: natural uranium, mixed plutonium - uranium oxide (MOX) having a burnup of 21 MWd/kg, and slightly enriched uranium (SEU) having burnups of either 21 or 31 MWd/kg

  6. Status of the atomized uranium silicide fuel development at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C.K.; Kim, K.H.; Park, H.D.; Kuk, I.H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-08-01

    While developing KMRR fuel fabrication technology an atomizing technique has been applied in order to eliminate the difficulties relating to the tough property of U{sub 3}Si and to take advantage of the rapid solidification effect of atomization. The comparison between the conventionally comminuted powder dispersion fuel and the atomized powder dispersion fuel has been made. As the result, the processes, uranium silicide powdering and heat treatment for U{sub 3}Si transformation, become simplified. The workability, the thermal conductivity and the thermal compatibility of fuel meat have been investigated and found to be improved due to the spherical shape of atomized powder. In this presentation the overall developments of atomized U{sub 3}Si dispersion fuel and the planned activities for applying the atomizing technique to the real fuel fabrication are described.

  7. Radiation protection training at uranium hexafluoride and fuel fabrication plants

    International Nuclear Information System (INIS)

    Brodsky, A.; Soong, A.L.; Bell, J.

    1985-05-01

    This report provides general information and references useful for establishing or operating radiation safety training programs in plants that manufacture nuclear fuels, or process uranium compounds that are used in the manufacture of nuclear fuels. In addition to a brief summary of the principles of effective management of radiation safety training, the report also contains an appendix that provides a comprehensive checklist of scientific, safety, and management topics, from which appropriate topics may be selected in preparing training outlines for various job categories or tasks pertaining to the uranium nuclear fuels industry. The report is designed for use by radiation safety training professionals who have the experience to utilize the report to not only select the appropriate topics, but also to tailor the specific details and depth of coverage of each training session to match both employee and management needs of a particular industrial operation. 26 refs., 3 tabs

  8. Phenomenology of uranium-plutonium homogenization in nuclear fuels

    International Nuclear Information System (INIS)

    Marin, J.M.

    1988-01-01

    The uranium and plutonium cations distribution in mixed oxide fuels (U 1-y Pu y )O 2 with y ≤ 0.1 has been studied in laboratory with industrial fabrication methods. Our experiences has showed a slow cations migration. In the substoichiometry (UPu)O 2-x the diffusion is in connection with the plutonium valence which is an indicator of the oxidoreduction state of the crystal lattice. The plutonium valence is in connection with the oxygen ion deficit in order to compensate the electrical charge. The oxygen ratio of the solid depends of the oxygen partial pressure prevailing at the time of product elaboration but it can be modified by impurities. These impurities permit to increase or decrease the fuel characteristics and performances. An homogeneity analysis methodology is proposed, its objective is to classify the mixed oxide fuels according to the uranium and plutonium ions distribution [fr

  9. Determination of uranium traces in fuel cans of nuclear reactors

    International Nuclear Information System (INIS)

    Acosta L, C.E.; Benavides M, A.M.; Sanchez P, L.A.; Nava S, G.F.

    1997-01-01

    The objective of this work is to quantify the uranium content that as impurity can be found in zircon and zircaloy alloys which are used in the construction of fuel cans. The determination of this serves as a quality control measure due to that the increment of uranium content in alloy, diminishing the corrosion resistance. The fluorimetric method was used to do this determination. It is a very sensitive, reliable, rapid method also high reproducibility and repeatability as well as low detection limits (0.25 mg/kg). (Author)

  10. Transition from uranium to denatured uranium/thorium fuel in an existing PWR

    International Nuclear Information System (INIS)

    Walters, M.A.

    1982-01-01

    The purpose of this research was to determine whether it is possible to make a gradual transition from uranium to denatured uranium/thorium (DUTH) fuel in an existing PWR by adding DUTH assemblies during each scheduled refueling and, if the transition is possible, to develop a general procedure for making it. The feasibility of the transition was established by identifying acceptable refueling schemes for a series of transition cores, and in the process, a method for identifying acceptable schemes evolved. The utility of the method was then demonstrated by applying it to a standard reactor operating under normal conditions. The vehicle used to examine proposed fuel mixtures and to select acceptable ones was a set of one-dimensional computer codes. The core was modeled as a set of five concentric fuel zones with a reflector. Fuel mixtures were proposed and the computer codes were used to determine whether a mixture was acceptable, i.e., whether it had the desired k-effective and flux and power distributions. The parameters allowed to vary in selection of proposed fuel mixtures were enrichment of fresh fuel assemblies, number of uranium and DUTH assemblies added during each refueling, and distribution of fuel in the core. Results of the research showed that a gradual transition is possible. Furthermore, there is a method that allows the identification of fuel mixtures that are likely to be acceptable. It requires the calculation of K-infinity for the entire proposed core and for some of its regions. These values of K-infinity and relationships developed in this research can be used to predict the flux distribution and the final k-effective for the proposed fuel mixture

  11. Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

    Science.gov (United States)

    Mcdonald, G. E.

    1973-01-01

    A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

  12. Uranium-plutonium fuel for fast reactors

    International Nuclear Information System (INIS)

    Antipov, S.A.; Astafiev, V.A.; Clouchenkov, A.E.; Gustchin, K.I.; Menshikova, T.S.

    1996-01-01

    Technology was established for fabrication of MOX fuel pellets from co-precipitated and mechanically blended mixed oxides. Both processes ensure the homogeneous structure of pellets readily dissolvable in nitric acid upon reprocessing. In order to increase the plutonium charge in a reactor-burner a process was tested for producing MOX fuel with higher content of plutonium and an inert diluent. It was shown that it is feasible to produce fuel having homogeneous structure and the content of plutonium up to 45% mass

  13. Uranium Resource Availability Analysis of Four Nuclear Fuel Cycle Options

    International Nuclear Information System (INIS)

    Youn, S. R.; Lee, S. H.; Jeong, M. S.; Kim, S. K.; Ko, W. I.

    2013-01-01

    Making the national policy regarding nuclear fuel cycle option, the policy should be established in ways that nuclear power generation can be maintained through the evaluation on the basis of the following aspects. To establish the national policy regarding nuclear fuel cycle option, that must begin with identification of a fuel cycle option that can be best suited for the country, and the evaluation work for that should be proceeded. Like all the policy decision, however, a certain nuclear fuel cycle option cannot be superior in all aspects of sustain ability, environment-friendliness, proliferation-resistance, economics, technologies, which make the comparison of the fuel cycle options very complicated. For such a purpose, this paper set up four different fuel cycle of nuclear power generation considering 2nd Comprehensive Nuclear Energy Promotion Plan(CNEPP), and analyzed material flow and features in steady state of all four of the fuel cycle options. As a result of an analysis on material flow of each nuclear fuel cycle, it was analyzed that Pyro-SFR recycling is most effective on U resource availability among four fuel cycle option. As shown in Figure 3, OT cycle required the most amount of U and Pyro-SFR recycle consumed the least amount of U. DUPIC recycling, PWR-MOX recycling, and Pyro-SFR recycling fuel cycle appeared to consumed 8.2%, 12.4%, 39.6% decreased amount of uranium respectively compared to OT cycle. Considering spent fuel can be recycled as potential energy resources, U and TRU taken up to be 96% is efficiently used. That is, application period of limited uranium natural resources can be extended, and it brings a great influence on stable use of nuclear energy

  14. Dual fuel gradients in uranium silicide plates

    Energy Technology Data Exchange (ETDEWEB)

    Pace, B.W. [Babock and Wilcox, Lynchburg, VA (United States)

    1997-08-01

    Babcock & Wilcox has been able to achieve dual gradient plates with good repeatability in small lots of U{sub 3}Si{sub 2} plates. Improvements in homogeneity and other processing parameters and techniques have allowed the development of contoured fuel within the cladding. The most difficult obstacles to overcome have been the ability to evaluate the bidirectional fuel loadings in comparison to the perfect loading model and the different methods of instilling the gradients in the early compact stage. The overriding conclusion is that to control the contour of the fuel, a known relationship between the compact, the frames and final core gradient must exist. Therefore, further development in the creation and control of dual gradients in fuel plates will involve arriving at a plausible gradient requirement and building the correct model between the compact configuration and the final contoured loading requirements.

  15. Preliminary concepts: coordinated safeguards for materials management in a thorium--uranium fuel reprocessing plant

    International Nuclear Information System (INIS)

    Hakkila, E.A.; Barnes, J.W.; Dayem, H.A.; Dietz, R.J.; Shipley, J.P.

    1978-10-01

    This report addresses preliminary concepts for coordinated safeguards materials management in a typical generic thorium--uranium-fueled light-water reactor (LWR) fuels reprocessing plant. The reference facility is designed to recover thorium and uranium from first-generation (denatured 235 U) startup fuels, first-recycle and equilibrium (denatured 233 U) thorium--uranium LWR fuels, and to recover the plutonium generated in the 238 U denaturant as well. 12 figures, 3 tables

  16. Chapter 1. General information about uranium. 1.10. Uranium application

    International Nuclear Information System (INIS)

    Khakimov, N.; Nazarov, Kh.M.; Mirsaidov, I.U.

    2011-01-01

    Full text: Metallic uranium or its compounds are used as nuclear fuel in nuclear reactors. A natural or low-enriched admixture of uranium isotopes is applied in stationery reactors of nuclear power plants, and products of a high enrichment degree are used in nuclear power plants or in reactors that operates with fast neutrons. 235 U is a source of nuclear energy in nuclear weapons. Depleted uranium is used as armour-piercing core in bombshells. 238 U serves as a source of secondary nuclear fuel - plutonium. (author)

  17. Chapter 1. General information about uranium. 1.10. Uranium application

    International Nuclear Information System (INIS)

    Khakimov, N.; Nazarov, Kh.M.; Mirsaidov, I.U.

    2012-01-01

    Full text: Metallic uranium or its compounds are used as nuclear fuel in nuclear reactors. A natural or low-enriched admixture of uranium isotopes is applied in stationery reactors of nuclear power plants, and products of a high enrichment degree are used in nuclear power plants or in reactors that operates with fast neutrons. 235 U is a source of nuclear energy in nuclear weapons. Depleted uranium is used as armour-piercing core in bombshells. 238 U serves as a source of secondary nuclear fuel - plutonium.

  18. Evaluation of bioassay program at uranium fuel fabrication plants

    International Nuclear Information System (INIS)

    Biggs, D.

    1981-03-01

    Results of a comprehensive study of urinalysis, lung burden and personal air sample measurements for workers at a uranium fuel fabrication plant are presented. Correlations between measurements were found and regression models used to explain the relationship between lung burden, daily intakes and urinary excretions of uranium. Assuming the ICRP lung model, the lung burden histories of ten workers were used to estimate the amounts in each of the long-term compartments of the lung. Estimates of the half lives of each compartment and of the maximum relative contributions to the urine from each compartment are given. These values were then used to predict urinary excretions from the long-term compartments for workers at another fuel fabrication plant. The standard error of estimate compared well with the daily variation in urinary excretion. (author)

  19. Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

    International Nuclear Information System (INIS)

    Clegg, L.J.; Coady, J.R.

    1977-01-01

    The two books of Volume 1 comprise the first in a three-volume series of compilations on the radioactive decay propertis of CANDU fuel and deal with the natural uranium fuel cycle. Succeeding volumes will deal with fuel cycles based on plutonium recycle and thorium. In Volume 1 which is divided into three parts, the computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 3 contains the data relating to the plutonium product and the high level wastes produced during fuel reprocessing. (author)

  20. Low enrichment Mo-99 target development program at ANSTO

    International Nuclear Information System (INIS)

    Donlevy, Therese M.; Anderson, Peter J.; Beattie, David; Braddock, Ben; Fulton, Scott; Godfrey, Robert; Law, Russell; McNiven, Scott; Sirkka, Pertti; Storr, Greg; Wassink, David; Wong, Alan; Yeoh, Guan

    2002-01-01

    The Australian Nuclear Science and Technology Organisation (ANSTO, formerly AAEC) has been producing fission product Mo-99 in HIFAR, from the irradiation of Low Enrichment Uranium (LEU) UO 2 targets, for nearly thirty years. Over this period, the U-235 enrichment has been increased in stages, from natural to 1.8% to 2.2%. The decision to provide Australia with a replacement research reactor (RRR) for HIFAR has created an ideal opportunity to review and improve the current Mo-99 production process from target design through to chemical processing and waste management options. ANSTO has entered into a collaboration with Argonne National Laboratory (RERTR) to develop a target using uranium metal foil with U-235 enrichment of less than 20% The initial focus has been to demonstrate use of LEU foil targets in HIFAR, using existing irradiation methodology. The current effort focussed on designing a target assembly with optimised thermohydraulic characteristics to accommodate larger LEU foils to meet Mo-99 production needs. The ultimate goal is to produce an LEU target suitable for use in the Replacement Research Reactor when it is commissioned in 2005. This paper reports our activities on: - The regulatory approval processes required in order to undertake irradiation of this new target; -Supporting calculations (neutronics, computational fluid dynamics) for safety submission; - Design challenges and changes to prototype irradiation; - Trial irradiation of LEU foil target in HIFAR; - Future target and rig development program at ANSTO. (author)

  1. Melting temperature of uranium - plutonium mixed oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, Tetsuya; Hirosawa, Takashi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-08-01

    Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960`s and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960`s and that some of the 1960`s data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO{sub 2} - PuO{sub 2} - PuO{sub 1.61} ideal solution model, and then formulized. (J.P.N.)

  2. Melting temperature of uranium - plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    Ishii, Tetsuya; Hirosawa, Takashi

    1997-08-01

    Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960's and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960's and that some of the 1960's data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO 2 - PuO 2 - PuO 1.61 ideal solution model, and then formulized. (J.P.N.)

  3. Uranium density reduction on fuel element side plates assessment

    International Nuclear Information System (INIS)

    Rios, Ilka A.; Andrade, Delvonei A.; Domingos, Douglas B.; Umbehaun, Pedro E.

    2011-01-01

    During operation of IEA-R1 research reactor, located at Instituto de Pesquisas Energeticas e Nucleares, IPEN - CNEN/SP, an abnormal oxidation on some fuel elements was noted. It was also verified, among the possible causes of the problem, that the most likely one was insufficient cooling of the elements in the core. One of the propositions to solve or minimize the problem is to reduce uranium density on fuel elements side plates. In this paper, the influence of this change on neutronic and thermal hydraulic parameters for IEA-R1 reactor is verified by simulations with the codes HAMMER and CITATION. Results are presented and discussed. (author)

  4. Uranium density reduction on fuel element side plates assessment

    Energy Technology Data Exchange (ETDEWEB)

    Rios, Ilka A. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Andrade, Delvonei A.; Domingos, Douglas B.; Umbehaun, Pedro E. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    During operation of IEA-R1 research reactor, located at Instituto de Pesquisas Energeticas e Nucleares, IPEN - CNEN/SP, an abnormal oxidation on some fuel elements was noted. It was also verified, among the possible causes of the problem, that the most likely one was insufficient cooling of the elements in the core. One of the propositions to solve or minimize the problem is to reduce uranium density on fuel elements side plates. In this paper, the influence of this change on neutronic and thermal hydraulic parameters for IEA-R1 reactor is verified by simulations with the codes HAMMER and CITATION. Results are presented and discussed. (author)

  5. Milling uranium silicide powder for dispersion nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, E.; Silva, D.G.; Souza, J.A.B.; Durazzo, M. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Riella, H.G. [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil)

    2009-07-01

    Full text: Uranium silicide (U3Si2) is presently considered the best fuel qualified so far in terms of uranium loading and performance. Stability of the U3Si2 fuel with uranium density of 4.8 g/cm3 was confirmed by burnup stability tests performed during the Reduced Enrichment for Research and Test Reactors (RERTR) program. This fuel was chosen to compose the first core of the new Brazilian Multipurpose Research Reactor (RMB), planned to be constructed in the next years. This new reactor will consume bigger quantities of U3Si2 powder, when compared with the small consumption of the IEA-R1 research reactor of IPEN-CNEN/SP, the unique MTR type research reactor operating in the country. At the present time, the milling operation of U3Si2 ingots is made manually. In order to increase the powder production capacity, the manual milling must be replaced by an automated procedure. This paper describes a new milling machine and procedure developed to produce U3Si2 powder with higher efficiency. (author)

  6. Uranium chloride extraction of transuranium elements from LWR fuel

    International Nuclear Information System (INIS)

    Miller, W.E.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Pierce, R.D.

    1992-01-01

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800 C to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein. 1 figure

  7. Licensed-fuel-facility status report, inventory difference data January 1981-June 1981

    International Nuclear Information System (INIS)

    1982-07-01

    NRC is committed to the periodic release of inventory difference data from the licensed fuel facilities after the agency has had an opportunity to review the data and has performed any related investigations associated with the data. Information included in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  8. Caramel, uranium oxide fuel plates for water cooled reactors

    International Nuclear Information System (INIS)

    Bussy, Pierre; Delafosse, Jacques; Lestiboudois, Guy; Cerles, J.-M.; Schwartz, J.-P.

    1979-01-01

    The fuel is composed of thin plates assembled parallel to each other to form bundles or assemblies. Each plate is composed of a pavement of uranium oxide pellets, insulated from each other by a zircaloy cladding. The 235 U enrichment does not exceed 8%. The range of uses for this fuel extends from electric power generating reactors to irradiation reactors for research work. A parametric study in test loops has made it possible to determine the operating limits of this thick fuel, without bursting. The resulting diagram gives the permissible power densities, with and without cycling for specific burn-ups beyond 50,000 MWd/t. The thinnest plates were also irradiated in total in the form of advance assemblies irradiated in the core of the OSIRIS pile prior to its transformation. This transformation and the operation of this reactor with a core of 'Caramel' elements is the main trial experiment of this fuel [fr

  9. RA-3 core with uranium silicide fuel elements

    International Nuclear Information System (INIS)

    Abbate, Maximo J.; Sbaffoni, Maria M.

    2000-01-01

    Following on with studies on uranium silicide fuel elements, this paper reports some comparisons between the use of standard ECN [U 3 O 8 ] fuel elements and type P-06 [from U 3 Si 2 ] fuel elements in the RA-3 core.The first results showed that the calculated overall mean burn up is in agreement with that reported for the facility, which gives more confidence to the successive ones. Comparing the mentioned cores, the silicide one presents several advantages such as: -) a mean burn up increase of 18 %; -) an extraction burn up increase of 20 %; -) 37.4 % increase in full power days, for mean burn up. All this is meritorious for this fuel. Moreover, grouped and homogenized libraries were prepared for CITVAP code that will be used for planning experiments and other bidimensional studies. Preliminary calculations were also performed. (author)

  10. Radionuclide compositions of spent fuel and high level waste for the uranium and plutonium fuelled PWR

    International Nuclear Information System (INIS)

    Fairclough, M.P.; Tymons, B.J.

    1985-06-01

    The activities of a selection of radionuclides are presented for three types of reactor fuel of interest in radioactive waste management. The fuel types are for a uranium 'burning' PWR, a plutonium 'burning' PWR using plutonium recycled from spent uranium fuel and a plutonium 'burning' PWR using plutonium which has undergone multiple recycle. (author)

  11. Automation of potentiometric titration for the determination of uranium in nuclear fuel materials

    International Nuclear Information System (INIS)

    Kelkar, Anoop; Pandey, Ashish; Kapoor, Y.S.; Kumar, Manish; Singh, Mamta; Fulzele, Ajeet; Prakash, Amrit; Afzal, Mohd; Panakkal, J.P.

    2010-01-01

    Advanced Fuel Fabrication Facility is fabricating various types of mixed oxide fuels, namely for PHWR, BWR, FBTR and PFBR. Precise determination of uranium in MOX fuel sample is important to get desired burn up in the reactor. The modified Davies and Gray method is routinely used for the potentiometric titration of uranium

  12. Irradiation behavior of uranium-molybdenum dispersion fuel: Fuel performance data from RERTR-1 and RERTR-2

    International Nuclear Information System (INIS)

    Meyer, M.K.; Clark, C.R.; Hayes, S.L.; Strain, R.V.; Hofman, G.L.; Snelgrove, J.L.; Park, J.M.; Kim, K.H.

    1999-01-01

    This paper presents quantitative data on the irradiation behavior of uranium-molybdenum fuels from the low temperature RERTR-1 and -2 experiments. Fuel swelling measurements of U-Mo fuels at ∼40% and ∼70% burnup are presented. The rate of fuel-matrix interaction layer growth is estimated. Microstructures of fuel in the pre- and postirradiation condition were compared. Based on these data, a qualitative picture of the evolution of the U-Mo fuel microstructure during irradiation has been developed. Estimates of uranium-molybdenum fuel swelling and fuel-matrix interaction under high-power research reactor operating conditions are presented. (author)

  13. Linking fuel design features ampersand plant management to uranium, SWU savings

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This article, contributed by Scott Garrett, Manager of Planning and Uranium Operations for Siemens Power Corporation in Bellevue, Washington, explores the impact of advances in fuel design and fuel management strategies on uranium utilization in the United States. Nuclear plant operators are deriving substantial benefits from these changes, including longer fuel cycle lengths, increased burnup, and added capacity - and experiencing cost savings in both uranium and enrichment services at the same time

  14. Acceptance criteria for the low enriched uranium reform amendments

    International Nuclear Information System (INIS)

    Emeigh, C.W.; Gundersen, G.E.; Withee, C.J.

    1984-05-01

    Revisions have been made to the material control and accounting requirements for NRC licensees authorized to possess and use more than one effective kilogram of special nuclear material of low strategic significance to have material control and accounting systems able to (1) confirm the presence of special nuclear material, (2) resolve indications of missing material, and (3) aid in the investigation and recovery of missing material. This document presents criteria that can be used to aid in judging the acceptability of licensee plans that would be submitted to the NRC for implementing these capabilities. General performance objectives, system capabilities, and recordkeeping are addressed

  15. Temperature coefficients in the Dragon low-enriched power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1972-05-15

    The temperature coefficient of the fuel and of the moderator have been evaluated for the Dragon HTR design for different stages in reactor life, initial core, end of no-refuelling period and equilibrium conditions. The investigation has shown the low-enriched HTR to have a strong, positive moderator coefficient. In some cases and for special operating conditions, even leading to a positive total temperature coefficient. This does not imply, however, that the HTR is an unsafe reactor system. By adequate design of the control system, safe and reliable operating characteristics can be achieved. This has already been proved satisfactory through many years of operation of other graphite moderated systems, such as the Magnox stations.

  16. Behaviour of irradiated uranium silicide fuel revisited

    International Nuclear Information System (INIS)

    Finlay, M. Ross; Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

    2002-01-01

    Irradiated U 3 Si 2 dispersion fuels demonstrate very low levels of swelling, even at extremely high burn-up. This behaviour is attributed to the stability of fission gas bubbles that develop during irradiation. The bubbles remain uniformly distributed throughout the fuel and show no obvious signs of coalescence. Close examination of high burn-up samples during the U 3 Si 2 qualification program revealed a bimodal distribution of fission gas bubbles. Those observations suggested that an underlying microstructure was responsible for the behaviour. An irradiation induced recrystallisation model was developed that relied on the presence of sufficient grain boundary surface to trap and pin fission gas bubbles and prevent coalescence. However, more recent work has revealed that the U 3 Si 2 becomes amorphous almost instantaneously upon irradiation. Consequently, the recrystallisation model does not adequately explain the nucleation and growth of fission gas bubbles in U 3 Si 2 . Whilst it appears to work well within the range of measured data, it cannot be relied on to extrapolate beyond that range since it is not mechanistically valid. A review of the mini-plates irradiated in the Oak Ridge Research Reactor from the U 3 Si 2 qualification program has been performed. This has yielded a new understanding of U 3 Si 2 behaviour under irradiation. (author)

  17. Comparison of the Environment, Health, And Safety Characteristics of Advanced Thorium- Uranium and Uranium-Plutonium Fuel Cycles

    Science.gov (United States)

    Ault, Timothy M.

    The environment, health, and safety properties of thorium-uranium-based (''thorium'') fuel cycles are estimated and compared to those of analogous uranium-plutonium-based (''uranium'') fuel cycle options. A structured assessment methodology for assessing and comparing fuel cycle is refined and applied to several reference fuel cycle options. Resource recovery as a measure of environmental sustainability for thorium is explored in depth in terms of resource availability, chemical processing requirements, and radiological impacts. A review of available experience and recent practices indicates that near-term thorium recovery will occur as a by-product of mining for other commodities, particularly titanium. The characterization of actively-mined global titanium, uranium, rare earth element, and iron deposits reveals that by-product thorium recovery would be sufficient to satisfy even the most intensive nuclear demand for thorium at least six times over. Chemical flowsheet analysis indicates that the consumption of strong acids and bases associated with thorium resource recovery is 3-4 times larger than for uranium recovery, with the comparison of other chemical types being less distinct. Radiologically, thorium recovery imparts about one order of magnitude larger of a collective occupational dose than uranium recovery. Moving to the entire fuel cycle, four fuel cycle options are compared: a limited-recycle (''modified-open'') uranium fuel cycle, a modified-open thorium fuel cycle, a full-recycle (''closed'') uranium fuel cycle, and a closed thorium fuel cycle. A combination of existing data and calculations using SCALE are used to develop material balances for the four fuel cycle options. The fuel cycle options are compared on the bases of resource sustainability, waste management (both low- and high-level waste, including used nuclear fuel), and occupational radiological impacts. At steady-state, occupational doses somewhat favor the closed thorium option while low

  18. Future fuel cycles

    International Nuclear Information System (INIS)

    Archinoff, G.H.

    1980-01-01

    A fuel cycle must offer both financial and resource savings if it is to be considered for introduction into Ontario's nuclear system. The most promising alternative CANDU fuel cycles are examined in the context of both of these factors over a wide range of installed capacity growth rates and economic assumptions, in order to determine which fuel cycle, or cycles, should be introduced, and when. It is concluded that the optimum path for the long term begins with the prompt introduction of the low-enriched-uranium fuel cycle. For a wide range of conditions, this cycle remains the optimum throughout the very long term. Conditions of rapid nuclear growth and very high uranium price escalation rates warrant the supersedure of the low-enriched-uranium cycle by either a plutonium-topped thorium cycle or plutonium recycle, beginning between 2010 and 2025. It is also found that the uranium resource position is sound in terms of both known resources and production capability. Moreover, introduction of the low-enriched-uranium fuel cycle and 1250 MWe reactor units will assure the economic viability of nuclear power until at least 2020, even if uranium prices increase at a rate of 3.5% above inflation. The interrelationship between these two conclusions lies in the tremendous incentive for exploration which will occur if the real uranium price escalation rate is high. From a competitive viewpoint, nuclear power can withstand increases in the price of uranium. However, such increases will likely further expand the resource base, making nuclear an even more reliable energy source. (auth)

  19. Low content uranium alloys for nuclear fuels

    International Nuclear Information System (INIS)

    Aubert, H.; Laniesse, J.

    1964-01-01

    A description is given of the structure and the properties of low content alloys containing from 0.1 to 0.5 per cent by weight of Al, Fe, Cr, Si, Mo or a combination of these elements. A study of the kinetics and of the mode of transformation has made it possible to choose the most satisfactory thermal treatment. An attempt has been made to prepare alloys suitable for an economical industrial development having a small α grain structure without marked preferential orientation, with very fine and stable precipitates as well as a high creep-resistance. The physical properties and the mechanical strength of these alloys are given for temperatures of 20 to 600 deg C. These alloys proved very satisfactory when irradiated in the form of normal size fuel elements. (authors) [fr

  20. Penetrate-leach dissolution of zirconium-clad uranium and uranium dioxide fuels

    International Nuclear Information System (INIS)

    Harmon, H.D.

    1975-01-01

    A new decladding-dissolution process was developed for zirconium-clad uranium metal and UO 2 fuels. The proposed penetrate-leach process consists of penetrating the zirconium cladding with Alniflex solution (2M HF--1M HNO 3 --1M Al(NO 3 ) 3 --0.1M K 2 Cr 2 O 7 ) and of leaching the exposed core with 10M HNO 3 . Undissolved cladding pieces are discarded as solid waste. Periodic HF and HNO 3 additions, efficient agitation, and in-line zirconium analyses are required for successful control of ZrF 4 and/or AlF 3 precipitation during the cladding-penetration step. Preliminary solvent extraction studies indicated complete recovery of uranium with 30 vol. percent tributyl phosphate (TBP) from both Alniflex solution and blended Alniflex-HNO 3 leach solutions. With 7.5 vol. percent TBP, high extractant/feed flow ratios and low scrub flows are required for satisfactory uranium recovery from Alniflex solution. Modified waste-handling procedures may be required for Alniflex waste, because it cannot be evaporated before neutralization and large quantities of solids are generated on neutralization. The effect of unstable UZr 3 (epsilon phase of uranium-zirconium system) on the safety of penetrate-leach dissolution was investigated

  1. IAEA Activities on Uranium Resources and Production, and Databases for the Nuclear Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Ganguly, C.; Slezak, J. [Divison of Nuclear Fuel Cycle and Waste Technology, International Atomic Energy Agency, Vienna (Austria)

    2014-05-15

    In recent years rising expectation for nuclear power has led to a significant increase in the demand for uranium and in turn dramatic increases in uranium exploration, mining and ore processing activities worldwide. Several new countries, often with limited experience, have also embarked on these activities. The ultimate goal of the uranium raw material industry is to provide an adequate supply of uranium that can be delivered to the market place at a competitive price by environmentally sound, mining and milling practices. The IAEA’s programme on uranium raw material encompass all aspects of uranium geology and deposits, exploration, resources, supply and demand, uranium mining and ore processing, environmental issues in the uranium production cycle and databases for the uranium fuel cycle. Radiological safety and environmental protection are major challenges in uranium mines and mills and their remediation. The IAEA has revived its programme for the Uranium Production Site Appraisal Team (UPSAT) to assist Member States to improve operational and safety performances at uranium mines and mill sites. The present paper summarizes the ongoing activities of IAEA on uranium raw material, highlighting the status of global uranium resources, their supply and demand, the IAEA database on world uranium deposit (UDEPO) and nuclear fuel cycle information system (NFCIS), recent IAEA Technical Meetings (TM) and related ongoing Technical Cooperation (TC) projects. (author)

  2. Fabrication of Uranium Oxycarbide Kernels for HTR Fuel

    International Nuclear Information System (INIS)

    Barnes, Charles; Richardson, Clay; Nagley, Scott; Hunn, John; Shaber, Eric

    2010-01-01

    Babcock and Wilcox (B and W) has been producing high quality uranium oxycarbide (UCO) kernels for Advanced Gas Reactor (AGR) fuel tests at the Idaho National Laboratory. In 2005, 350-(micro)m, 19.7% 235U-enriched UCO kernels were produced for the AGR-1 test fuel. Following coating of these kernels and forming the coated-particles into compacts, this fuel was irradiated in the Advanced Test Reactor (ATR) from December 2006 until November 2009. B and W produced 425-(micro)m, 14% enriched UCO kernels in 2008, and these kernels were used to produce fuel for the AGR-2 experiment that was inserted in ATR in 2010. B and W also produced 500-(micro)m, 9.6% enriched UO2 kernels for the AGR-2 experiments. Kernels of the same size and enrichment as AGR-1 were also produced for the AGR-3/4 experiment. In addition to fabricating enriched UCO and UO2 kernels, B and W has produced more than 100 kg of natural uranium UCO kernels which are being used in coating development tests. Successive lots of kernels have demonstrated consistent high quality and also allowed for fabrication process improvements. Improvements in kernel forming were made subsequent to AGR-1 kernel production. Following fabrication of AGR-2 kernels, incremental increases in sintering furnace charge size have been demonstrated. Recently small scale sintering tests using a small development furnace equipped with a residual gas analyzer (RGA) has increased understanding of how kernel sintering parameters affect sintered kernel properties. The steps taken to increase throughput and process knowledge have reduced kernel production costs. Studies have been performed of additional modifications toward the goal of increasing capacity of the current fabrication line to use for production of first core fuel for the Next Generation Nuclear Plant (NGNP) and providing a basis for the design of a full scale fuel fabrication facility.

  3. Method for the chemical reprocessing of irradiated nuclear fuels, in particular nuclear fuels containing uranium

    International Nuclear Information System (INIS)

    Koch, G.

    1976-01-01

    In the chemical processing of irradiated uranium-containing nuclear fuels which are hydrolyzed with aqueous nitric acid, a suggestion is made to use as quaternary ammonium nitrate trialkyl-methyl ammonium nitrates as extracting agent, in which the sum of C atoms is greater than 16. In the illustrated examples, tricaprylmethylammonium nitrate, trilaurylmethylammonium nitrate and tridecylmethylammonium nitrate are named. (HPH/LH) [de

  4. Failure mechanisms for compacted uranium oxide fuel cores

    International Nuclear Information System (INIS)

    Berghaus, D.G.; Peacock, H.B.

    1980-01-01

    Tension, compression, and shear tests were performed on test specimens of aluminum-clad, compacted powder fuel cores to determine failure mechanisms of the core material. The core, which consists of 70% uranium oxide in an aluminum matrix, frequently fails during post-extrusion drawing. Tests were conducted to various strain levels up to failure of the core. Sections were made of tested specimens to microscopically study initiation of failure. Two failure modes wee observed. Tensile failure mode is initiated by prior tensile failure of uranium oxide particles with the separation path strongly influenced by the arrangement of particles. Delamination mode consists of the separation of laminae formed during extrusion of tubes. Separation proceeds from fine cracks formed parallel to the laminae. Tensile failure mode was experienced in tension and shear tests. Delamination mode was produced in compression tests

  5. Ternary carbide uranium fuels for advanced reactor design applications

    International Nuclear Information System (INIS)

    Knight, Travis; Anghaie, Samim

    1999-01-01

    Solid-solution mixed uranium/refractory metal carbides such as the pseudo-ternary carbide, (U, Zr, Nb)C, hold significant promise for advanced reactor design applications because of their high thermal conductivity and high melting point (typically greater than 3200 K). Additionally, because of their thermochemical stability in a hot-hydrogen environment, pseudo-ternary carbides have been investigated for potential space nuclear power and propulsion applications. However, their stability with regard to sodium and improved resistance to attack by water over uranium carbide portends their usefulness as a fuel for advanced terrestrial reactors. An investigation into processing techniques was conducted in order to produce a series of (U, Zr, Nb)C samples for characterization and testing. Samples with densities ranging from 91% to 95% of theoretical density were produced by cold pressing and sintering the mixed constituent carbides at temperatures as high as 2650 K. (author)

  6. Fabrication routes for Thorium and Uranium233 based AHWR fuel

    International Nuclear Information System (INIS)

    Danny, K.M.; Saraswat, Anupam; Chakraborty, S.; Somayajulu, P.S.; Kumar, Arun

    2011-01-01

    India's economic growth is on a fast growth track. The growth in population and economy is creating huge demand for energy which has to be met with environmentally benign technologies. Nuclear Energy is best suited to meet this demand without causing undue environmental impact. Considering the large thorium reserves in India, the future nuclear power program will be based on Thorium- Uranium 233 fuel cycle. The major characteristic of thorium as the fuel of future comes from its superior fuel utilization. 233 U produced in a reactor is always contaminated with 232 U. This 232 U undergoes a decay to produce 228 Th and it is followed by decay chain including 212 Bi and 208 Tl. Both 212 Bi and 208 Tl are hard gamma emitters ranging from 0.6 MeV-1.6 MeV and 2.6 MeV respectively, which necessitates its handling in hot cell. The average concentration of 232 U is expected to exceed 1000 ppm after a burn-up of 24,000 MWD/t. Work related to developing the fuel fabrication technology including automation and remotization needed for 233 U based fuels is in progress. Various process for fuel fabrication have been developed i.e. Coated Agglomerate Pelletisation (CAP), impregnation technique (Pellet/Gel), Sol Gel Micro-sphere Pelletisation (SGMP) apart from Powder to Pellet (POP) route. This paper describes each process with respect to its advantages, disadvantages and its amenability to automation and remotisation. (author)

  7. Biamperometric estimation of uranium in input KMP samples of spent fuel reprocessing plant: field experience

    International Nuclear Information System (INIS)

    Gurba, P.B.; Dhakras, S.P.; Chaugule, G.A.; Venugopal, A.K.; Singh, R.K.; Bajpai, D.D.; Nair, P.R.; Xavier, Mary; Aggarwal, S.K.

    2000-01-01

    Feasibility of simple, precise and accurate biamperometric determination of uranium at about 0.1 mg level was earlier established using simulated uranium standards. To evaluate the usefulness of this method for accurate determination of uranium in spent fuel dissolver solution samples, analytical work was carried out

  8. Mixing of Al into uranium silicides reactor fuels

    International Nuclear Information System (INIS)

    Ding, F.R.; Birtcher, R.C.; Kestel, B.J.; Baldo, P.M.

    1996-11-01

    SEM observations have shown that irradiation induced interaction of the aluminum cladding with uranium silicide reactor fuels strongly affects both fission gas and fuel swelling behaviors during fuel burn-up. The authors have used ion beam mixing, by 1.5 MeV Kr, to study this phenomena. RBS and the 27 Al(p, γ) 28 Si resonance nuclear reaction were used to measure radiation induced mixing of Al into U 3 Si and U 3 Si 2 after irradiation at 300 C. Initially U mixes into the Al layer and Al mixes into the U 3 Si. At a low dose, the Al layer is converted into UAl 4 type compound while near the interface the phase U(Al .93 Si .07 ) 3 grows. Under irradiation, Al diffuses out of the UAl 4 surface layer, and the lower density ternary, which is stable under irradiation, is the final product. Al mixing into U 3 Si 2 is slower than in U 3 Si, but after high dose irradiation the Al concentration extends much farther into the bulk. In both systems Al mixing and diffusion is controlled by phase formation and growth. The Al mixing rates into the two alloys are similar to that of Al into pure uranium where similar aluminide phases are formed

  9. Fuel balance in nuclear power with fast reactors without a uranium blanket

    International Nuclear Information System (INIS)

    Naumov, V.V.; Orlov, V.V.; Smirnov, V.S.

    1994-01-01

    General aspects related to replacing the uranium blanket of a lead-cooled fast reactor burning uranium-plutonium nitride fuel with a more efficient lead reflector are briefly discussed in the article. A study is very briefly summarized, which showed that a breeding ratio of about 1 and electric power of about 300 MW were achievable. A nuclear fuel balance is performed to estimate the increased consumption of uranium to produce power and the gains achievable by eliminating the uranium blanket. Elimination of the uranium blanket has the advantages of simplifying and improving the fast reactor and eliminating the production of weapons quality plutonium. 3 figs

  10. Uranium dioxide and beryllium oxide enhanced thermal conductivity nuclear fuel development

    International Nuclear Information System (INIS)

    Andrade, Antonio Santos; Ferreira, Ricardo Alberto Neto

    2007-01-01

    The uranium dioxide is the most used substance as nuclear reactor fuel for presenting many advantages such as: high stability even when it is in contact with water in high temperatures, high fusion point, and high capacity to retain fission products. The conventional fuel is made with ceramic sintered pellets of uranium dioxide stacked inside fuel rods, and presents disadvantages because its low thermal conductivity causes large and dangerous temperature gradients. Besides, the thermal conductivity decreases further as the fuel burns, what limits a pellet operational lifetime. This research developed a new kind of fuel pellets fabricated with uranium dioxide kernels and beryllium oxide filling the empty spaces between them. This fuel has a great advantage because of its higher thermal conductivity in relation to the conventional fuel. Pellets of this kind were produced, and had their thermophysical properties measured by the flash laser method, to compare with the thermal conductivity of the conventional uranium dioxide nuclear fuel. (author) (author)

  11. Uranium-fuel thermal reactor benchmark testing of CENDL-3

    International Nuclear Information System (INIS)

    Liu Ping

    2001-01-01

    CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)

  12. Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

    Energy Technology Data Exchange (ETDEWEB)

    Heidet, F.; Kim, T.; Grandy, C. (Nuclear Engineering Division)

    2012-07-30

    Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium is more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the

  13. Extending the world's uranium resources through advanced CANDU fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    De Vuono, Tony; Yee, Frank; Aleyaseen, Val; Kuran, Sermet; Cottrell, Catherine

    2010-09-15

    The growing demand for nuclear power will encourage many countries to undertake initiatives to ensure a self-reliant fuel source supply. Uranium is currently the only fuel utilized in nuclear reactors. There are increasing concerns that primary uranium sources will not be enough to meet future needs. AECL has developed a fuel cycle vision that incorporates other sources of advanced fuels to be adaptable to its CANDU technology.

  14. Kinetic and thermodynamic bases to resolve issues regarding conditioning of uranium metal fuels

    International Nuclear Information System (INIS)

    Johnson, A.B.; Ballinger, R.G.; Simpson, K.A.

    1994-12-01

    Numerous uranium - bearing fuels are corroding in fuel storage pools in several countries. At facilities where reprocessing is no longer available, dry storage is being evaluated to preclude aqueous corrosion that is ongoing. It is essential that thermodynamic and kinetic factors are accounted for in transitions of corroding uranium-bearing fuels to dry storage. This paper addresses a process that has been proposed to move Hanford N-Reactor fuel from wet storage to dry storage

  15. Development of a recovery process of scraps resulting from the manufacture of metallic uranium fuels

    International Nuclear Information System (INIS)

    Camilo, Ruth L.; Kuada, Terezinha A.; Forbicini, Christina A.L.G.O.; Cohen, Victor H.; Araujo, Bertha F.; Lobao, Afonso S.T.

    1996-01-01

    The study of the dissolution of natural metallic uranium fuel samples with aluminium cladding is presented, in order to obtain optimized conditions for the system. The aluminium cladding was dissolved in an alkaline solution of Na OH/Na NO 3 and the metallic uranium with HNO 3 . A fumeless dissolution with total recovery of nitrous gases was achieved. The main purpose of this project was the recovery of uranium from scraps resulting from the manufacture of the metallic uranium fuel or other non specified fuels. (author)

  16. Radiological and environmental safety aspects of uranium fuel fabrication plants at Nuclear Fuel Complex, Hyderabad

    International Nuclear Information System (INIS)

    Viswanathan, S.; Surya Rao, B.; Lakshmanan, A.R.; Krishna Rao, T.

    1991-01-01

    Nuclear Fuel Complex, Hyderabad manufactures uranium dioxide fuel assemblies for PHWRs and BWRs operating in India. Starting materials are magnesium diuranate received from UCIL, Jaduguda and imported UF. Both of these are converted to UO 2 pellets by identical chemical processes and mechanical compacting. Since the uranium handled here is free of daughter product activities, external radiation is not a problem. Inhalation of airborne U compounds is one of the main source of exposure. Engineered protective measures like enclosures around U bearing powder handling equipment and local exhausts reduce worker's exposure. Installation of pre-filters, wet rotoclones and electrostatic precipitators in the ventillation system reduces the release of U into the environment. The criticality hazard in handling slightly enriched uranium is very low due to the built-in control based on geometry and inventory. Where airborne uranium is significant, workers are provided with protective respirators. The workers are regularly monitored for external exposure and also for internal exposure. The environmental releases from the NFC facility is well controlled. Soil, water and air from the NFC environment are routinely collected and analysed for all the possible pollutants. The paper describes the Health Physics experience during the last five years on occupational exposures and on environmental surveillance which reveals the high quality of safety observed in our nuclear fuel fabricating installations. (author). 4 refs., 6 tabs

  17. Minimization of waste from uranium purification, enrichment and fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-10-01

    As any industry, nuclear industry generates a diverse range of waste which has to be managed in a safe manner to be acceptable to the public and the environment. The cost of waste management, the risks to the public and employees, and the detriment to the environment are dependent on the quantity and radioactive content of the waste generated. Waste minimization is a necessary activity needed to reduce the impact from nuclear fuel cycle operations and it is included in the national policy of some countries. In recognition of the importance of the subject, the IAEA has decided to review the current status of the work aimed at waste minimization in the nuclear fuel cycle. The waste minimization issues related to the back end of the nuclear fuel cycle are covered in Technical Reports Series No. 377 'Minimization of Radioactive Waste from Nuclear Power Plants and the Back End of the Nuclear Fuel Cycle' published in 1995. The present report deals with the front end of the nuclear fuel cycle, including existing options, approaches, developments and some specific considerations to be taken into account in decision making on waste minimization. It has been recognized that, in comparison with the back end of the nuclear fuel cycle, much less information is available, and this report should be considered as a first attempt to analyse waste minimization practices and opportunities in uranium purification, conversion, enrichment and fuel fabrication. Although mining and milling is an important part of the front end of the nuclear fuel cycle, these activities are excluded from consideration since relevant activities are covered in other IAEA publications.

  18. Minimization of waste from uranium purification, enrichment and fuel fabrication

    International Nuclear Information System (INIS)

    1999-10-01

    As any industry, nuclear industry generates a diverse range of waste which has to be managed in a safe manner to be acceptable to the public and the environment. The cost of waste management, the risks to the public and employees, and the detriment to the environment are dependent on the quantity and radioactive content of the waste generated. Waste minimization is a necessary activity needed to reduce the impact from nuclear fuel cycle operations and it is included in the national policy of some countries. In recognition of the importance of the subject, the IAEA has decided to review the current status of the work aimed at waste minimization in the nuclear fuel cycle. The waste minimization issues related to the back end of the nuclear fuel cycle are covered in Technical Reports Series No. 377 'Minimization of Radioactive Waste from Nuclear Power Plants and the Back End of the Nuclear Fuel Cycle' published in 1995. The present report deals with the front end of the nuclear fuel cycle, including existing options, approaches, developments and some specific considerations to be taken into account in decision making on waste minimization. It has been recognized that, in comparison with the back end of the nuclear fuel cycle, much less information is available, and this report should be considered as a first attempt to analyse waste minimization practices and opportunities in uranium purification, conversion, enrichment and fuel fabrication. Although mining and milling is an important part of the front end of the nuclear fuel cycle, these activities are excluded from consideration since relevant activities are covered in other IAEA publications

  19. Postirradiation examination of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Strain, R.V.

    1998-01-01

    Two irradiation test vehicles, designated RERTR-2, were inserted into the Advanced Test reactor in Idaho in August 1997. These tests were designed to obtain irradiation performance information on a variety of potential new, high-density uranium alloy dispersion fuels, including U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru and U-10Mo-0.05Sn: the intermetallic compounds U 2 Mo and U-10Mo-0.-5Sn; the intermetallic compounds U 2 Mo and U 3 Si 2 were also included in the fuel test matrix. These fuels are included in the experiments as microplates (76 mm x 22 mm x 1.3mm outer dimensions) with a nominal fuel volume loading of 25% and irradiated at relatively low temperature (∼100 deg C). RERTR-1 and RERTR-2 were discharged from the reactor in November 1997 and July 1998, respectively at calculated peak fuel burnups of 45 and 71 at %-U 235 Both experiments are currently under examination at the Alpha Gamma Hot Cell Facility at Argonne National Laboratory in Chicago. This paper presents the postirradiation examination results available to date from these experiments. (author)

  20. Metallography of pitted aluminum-clad, depleted uranium fuel

    International Nuclear Information System (INIS)

    Nelson, D.Z.; Howell, J.P.

    1994-01-01

    The storage of aluminum-clad fuel and target materials in the L-Disassembly Basin at the Savannah River Site for more than 5 years has resulted in extensive pitting corrosion of these materials. In many cases the pitting corrosion of the aluminum clad has penetrated in the uranium metal core, resulting in the release of plutonium, uranium, cesium-137, and other fission product activity to the basin water. In an effort to characterize the extent of corrosion of the Mark 31A target slugs, two unirradiated slug assemblies were removed from basin storage and sent to the Savannah River Technology Center for evaluation. This paper presents the results of the metallography and photographic documentation of this evaluation. The metallography confirmed that pitting depths varied, with the deepest pit found to be about 0.12 inches (3.05 nun). Less than 2% of the aluminum cladding was found to be breached resulting in less than 5% of the uranium surface area being affected by corrosion. The overall integrity of the target slug remained intact

  1. Process for separately recovering uranium, transuranium elements, and fission products of uranium from atomic reactor fuel

    International Nuclear Information System (INIS)

    Balal, A.L.; Metscher, K.; Muehlig, B.; Reichmuth, C.; Schwarz, B.; Zimen, K.E.

    1976-01-01

    Spent reactor fuel elements are dissolved in dilute nitric acid. After addition of acetic acid as a complexing agent, the nitric acid is partly decomposed and the mixture subjected to electrolysis while a carrier liquid, which may be dilute acetic acid or a dilute mixture of acetic acid and nitric acid is caused to flow in the electric field between the electrodes either against the direction of ion migration or transversely thereto. The ions of uranium, plutonium, and other transuranium elements, and of fission products accumulate in discrete portions of the electrolyte and are separately withdrawn as at least three fractions after one or more stages of electrolysis

  2. Extraction of Uranium Using Nitrogen Dioxide and Carbon Dioxide for Spent Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Kayo Sawada; Daisuke Hirabayashi; Youichi Enokida [EcoTopia Science Institute, Nagoya University, Nagoya, 464-8603 (Japan)

    2008-07-01

    For the reprocessing of spent nuclear fuels, a new method to extract actinides from spent fuel using highly compressed gases, nitrogen dioxide and carbon dioxide was proposed. Uranium extraction from broken pieces, whose average grain size was 5 mm, of uranium dioxide pellet with nitrogen dioxide and carbon dioxide was demonstrated in the present study. (authors)

  3. Progress in irradiation performance of experimental uranium - Molybdenum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.

    2002-01-01

    High-density dispersion fuel experiment, RERTR-4, was removed from the Advanced Test Reactor (ATR) after reaching a peak U-235 burnup of ∼80% and is presently undergoing postirradiation examination at the ANL alpha-gamma hot cells. This test consists of 32 mini fuel plates of which 27 were fabricated with nominally 6 and 8 g cm -3 atomized and machined uranium alloy powders containing 7 wt% and 10 wt% molybdenum. In addition, two miniplates containing solid U-10 wt% Mo foils and three containing 6 g cm -3 U 3 Si 2 are part of the test. The results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from previous tests performed to lower burnup will be presented. (author)

  4. The relationship between natural uranium and advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    Lane, A.D.; McDonnell, F.N.; Griffiths, J.

    1988-11-01

    CANDU is the most uranium-economic type of thermal power reactor, and is the only type used in Canada. CANDU reactors consume approximately 15% of Canadian uranium production and support a fuel service industry valued at ∼$250 M/a. In addition to their once-through, natural-uranium fuel cycle, CANDU reactors are capable of operating with slightly-enriched uranium (SEU), uranium-plutonium and thorium cycles, more efficiently than other reactors. Only SEU is economically attractive in Canada now, but the other cycles are of interest to countries without indigenous fuel resources. A program is underway to establish the fuel technologies necessary for the use of SEU and the other fuel cycles in CANDU reactors. 22 refs

  5. Status of fuel element technology for plate type dispersion fuels with high uranium density

    International Nuclear Information System (INIS)

    Hrovat, M.; Huschka, H.; Koch, K.H.; Nazare, S.; Ondracek, G.

    1983-01-01

    A number of about 20 Material Test and Research Reactors in Germany and abroad is supplied with fuel elements by the company NUKEM. The power of these reactors differs widely ranging from up to about 100 MW. Consequently, the uranium density of the fuel elements in the meat varies considerably depending on the reactor type and is usually within the range from 0.4 to 1.3 g U/cm 3 if HEU is used. In order to convert these reactors to lower uranium enrichment (19.75% 235-U) extensive work is carried out at NUKEM since about two years with the goal to develop fuel elements with high U-density. This work is sponsored by the German Ministry for Research and Technology in the frame of the AF-program. This paper reports on the present state of development for fuel elements with high U-density fuels at NUKEM is reported. The development works were so far concentrated on UAl x , U 3 O 8 and UO 2 fuels which will be described in more detail. In addition fuel plates with new fuels like e.g. U-Si or U-Fe compounds are developed in collaboration with KfK. The required uranium densities for some typical reactors with low, medium, and high power are listed allowing a comparison of HEU and LEU uranium density requirements. The 235-U-content in the case of LEU is raised by 18%. Two different meat thicknesses are considered: Standard thickness of 0.5 mm; and increased thickness of 0.76 mm. From this data compilation the objective follows: in the case of conversion to LEU (19.75% 235-U-enrichment), uranium densities have to be made available up to 24 gU/cm 3 meat for low power level reactors, up to 33 gU/cm 3 meat for medium power level reactors, and between 5.75 and 7.03 g/cm 3 meat for high power level reactors according to this consideration

  6. Radiation Re-solution Calculation in Uranium-Silicide Fuels

    International Nuclear Information System (INIS)

    Matthews, Christopher; Andersson, Anders David Ragnar; Unal, Cetin

    2017-01-01

    The release of fission gas from nuclear fuels is of primary concern for safe operation of nuclear power plants. Although the production of fission gas atoms can be easily calculated from the fission rate in the fuel and the average yield of fission gas, the actual diffusion, behavior, and ultimate escape of fission gas from nuclear fuel depends on many other variables. As fission gas diffuses through the fuel grain, it tends to collect into intra-granular bubbles, as portrayed in Figure 1.1. These bubbles continue to grow due to absorption of single gas atoms. Simultaneously, passing fission fragments can cause collisions in the bubble that result in gas atoms being knocked back into the grain. This so called ''re-solution'' event results in a transient equilibrium of single gas atoms within the grain. As single gas atoms progress through the grain, they will eventually collect along grain boundaries, creating inter-granular bubbles. As the inter-granular bubbles grow over time, they will interconnect with other grain-face bubbles until a pathway is created to the outside of the fuel surface, at which point the highly pressurized inter-granular bubbles will expel their contents into the fuel plenum. This last process is the primary cause of fission gas release. From the simple description above, it is clear there are several parameters that ultimately affect fission gas release, including the diffusivity of single gas atoms, the absorption and knockout rate of single gas atoms in intra-granular bubbles, and the growth and interlinkage of intergranular bubbles. Of these, the knockout, or re-solution rate has an particularly important role in determining the transient concentration of single gas atoms in the grain. The re-solution rate will be explored in the following sections with regards to uranium-silicide fuels in order to support future models of fission gas bubble behavior.

  7. High enrichment to low enrichment core's conversion. Technical securities

    International Nuclear Information System (INIS)

    Abbate, P.; Madariaga, M.R.

    1990-01-01

    This work presents the fulfillment of the technical securities subscribed by INVAP S.E. for the conversion of a high enriched uranium core. The reactor (of 5 thermal Mw), built in the 50's and 60's, is of the 'swimming pool' type, with light water and fuel elements of the curve plates MTR type, enriched at 93.15 %. These are neutronic and thermohydraulic securities. (Author) [es

  8. Method of chemical reprocessing of irradiated nuclear fuels (especially fuels containing uranium)

    International Nuclear Information System (INIS)

    Koch, G.

    1975-01-01

    The invention deals with a method for the extraction especially of fast breeder fuels of high burn-up. A quaternary ammonium nitrate of high molecular weight is put into an organic diluting medium as extraction agent, corresponding to the general formula NRR'R''R'''NO 3 where R,R' and R'' are aliphatic radicals, R''' a methyl radical and the sum of the C atoms is greater than 16. After the extraction of the aqueous nitric acid containing nuclear fuel solution with this extracting agent, uranium, plutonium (or also thorium) can be found to a very high percentage in the organic phase and can be practically quantitatively back-extracted by means of diluted nitric acid, sulphuric acid or acetic acid. By using 30 volume percent tricapryl methyl ammonium nitrate in diethyl benzene for example, a distribution coefficient of 10.3 is obtained for uranium. (RB/LH) [de

  9. FY16 Status Report for the Uranium-Molybdenum Fuel Concept

    International Nuclear Information System (INIS)

    Bennett, Wendy D.; Doherty, Ann L.; Henager, Charles H.; Lavender, Curt A.; Montgomery, Robert O.; Omberg, Ronald P.; Smith, Mark T.; Webster, Ryan A.

    2016-01-01

    The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties. With sufficient development, it may be able to provide the Light Water Reactor industry with a melt-resistant, accident-tolerant fuel with improved safety response. The Pacific Northwest National Laboratory has been tasked with extrusion development and performing ex-reactor corrosion testing to characterize the performance of Uranium-Molybdenum fuel in both these areas. This report documents the results of the fiscal year 2016 effort to develop the Uranium-Molybdenum metal fuel concept for light water reactors.

  10. The Fabrication Problem Of U3Si2-Al Fuel With Uranium High Loading

    International Nuclear Information System (INIS)

    Supardjo

    1996-01-01

    The quality of U 3 Si 2 -Al dispersion fuel product is the main aim for each fabricator. Low loading of uranium fuel element is easily fabricated, but with the increased, uranium loading, homogeneity of uranium distribution is difficult to achieve and it always formed white spots, blister, and dogboning in the fuel plates. The problem can be eliminated by the increasing treatment of the fuel/Al powder. The precise selection of fuel/Al particles diameter is needed indeed to make easier in the homogeneous process of powder and the porosities arrangement in the fuel plates. The increasing of uranium loading at constant meat thickness will increase the meat hardness, therefore to withdraw the dogboning forming, the use of harder cladding materials is necessity

  11. Research on calculation of mixing fraction for natural uranium equivalent fuel

    International Nuclear Information System (INIS)

    Huang Shien; Wang Lianjie; Wei Yanqin; Li Qing; Zheng Jiye

    2013-01-01

    Based on the first-order perturbation theory and reasonable approximations, the calculation method of recycled uranium (RU) and depleted uranium (DU) mixing fraction for natural uranium equivalent (NUE) fuel was studied, so the equivalence between NUE fuel and natural uranium (NU) fuel was assured. The adopted calculation method accurately takes the variation of micro cross sections alone with fuel depletion into account. A computer code named ALPHA was programmed to execute the calculation procedure. Then the ALPHA code and the WIMS-AECL code compose a processing system, which is applicable to the mixing fraction calculation for heavy water reactor NUE fuel. The validation shows that the processing system can accurately calculate the mixing fraction for NUE fuel. (authors)

  12. FY16 Status Report for the Uranium-Molybdenum Fuel Concept

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, Wendy D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Doherty, Ann L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Henager, Charles H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Montgomery, Robert O. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Omberg, Ronald P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Mark T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Webster, Ryan A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-22

    The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties. With sufficient development, it may be able to provide the Light Water Reactor industry with a melt-resistant, accident-tolerant fuel with improved safety response. The Pacific Northwest National Laboratory has been tasked with extrusion development and performing ex-reactor corrosion testing to characterize the performance of Uranium-Molybdenum fuel in both these areas. This report documents the results of the fiscal year 2016 effort to develop the Uranium-Molybdenum metal fuel concept for light water reactors.

  13. Natural uranium equivalent fuel. An innovative design for proven CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Pineiro, F.; Ho, K.; Khaial, A.; Boubcher, M.; Cottrell, C.; Kuran, S. [Candu Energy Inc., Mississauga, Ontario (Canada); Zhenhua, Z.; Zhiliang, M. [Third Qinshan Nuclear Power Co., Haiyan, Zhejiang (China)

    2015-09-15

    The high neutron economy, on-power refuelling capability and fuel bundle design simplicity in CANDU® reactors allow for the efficient utilization of alternative fuels. Candu Energy Inc. (Candu), in collaboration with the Third Qinshan Nuclear Power Company (TQNPC), the China North Nuclear Fuel Corporation (CNNFC), and the Nuclear Power Institute of China (NPIC), has successfully developed an advanced fuel called Natural Uranium Equivalent (NUE). This innovative design consists of a mixture of recycled and depleted uranium, which can be implemented in existing CANDU stations thereby bringing waste products back into the energy stream, increasing fuel resources diversity and reducing fuel costs. (author)

  14. Natural uranium equivalent fuel an innovative design for proven CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Pineiro, F.; Ho, K.; Khaial, A.; Boubcher, M.; Cottrell, C.; Kuran, S., E-mail: fabricia.pineiro@candu.com [Candu Energy Inc., Mississauga, ON (Canada); Zhenhua, Z.; Zhiliang, M. [Third Qinshan Nuclear Power Company, Haiyan, Zhejiang (China)

    2015-07-01

    The high neutron economy, on-power refuelling capability and fuel bundle design simplicity in CANDU reactors allow for the efficient utilization of alternative fuels. Candu Energy Inc. (Candu), in collaboration with the Third Qinshan Nuclear Power Company (TQNPC), the China North Nuclear Fuel Corporation (CNNFC), and the Nuclear Power Institute of China (NPIC), has successfully developed an advanced fuel called Natural Uranium Equivalent (NUE). This innovative design consists of a mixture of recycled and depleted uranium, which can be implemented in existing CANDU stations thereby bringing waste products back into the energy stream, increasing fuel resources diversity and reducing fuel costs. (author)

  15. The use of uranium isotopes and the U/Th ratio to evaluate the fingerprint of plants following uranium releases from fuel cycle settlements

    International Nuclear Information System (INIS)

    Pourcelot, L.; Boulet, B.; Cariou, N.

    2015-01-01

    This paper uses data from the environmental monitoring of fuel cycle settlements. It aims to evaluate uranium released into the terrestrial environment. Measurement of uranium isotopes in terrestrial plants allows illustrating the consequences of chronic and incidental releases of depleted uranium into the atmosphere. However, such an analytical approach reaches its limits when natural uranium is released. Indeed, distinguishing natural uranium from releases and uranium from the radiological background is difficult. For this reason, we propose normalizing uranium activity measured in plants taken in the surroundings of nuclear sites with respect to 232 Th, considering that the source of this latter is the background. (authors)

  16. Sustainable and safe energy supply with seawater uranium fueled HTGR and its economy

    International Nuclear Information System (INIS)

    Fukaya, Y.; Goto, M.

    2017-01-01

    Highlights: • We discussed uranium resources with an energy security perspective. • We concluded seawater uranium is preferable for sustainability and energy security. • We evaluated electricity generation cost of seawater uranium fueled HTGR. • We concluded electricity generation with seawater uranium is reasonable. - Abstract: Sustainable and safe energy supply with High Temperature Gas-cooled Reactor (HTGR) fueled by uranium from seawater have been investigated and discussed. From the view point of safety feature of self-regulation with thermal reactor of HTGR, the uranium resources should be inexhaustible. The seawater uranium is expected to be alternative resources to conventional resources because it exists so much in seawater as a solute. It is said that 4.5 billion tons of uranium is dissolved in the seawater, which corresponds to a consumption of approximately 72 thousand years. Moreover, a thousand times of the amount of 4.5 trillion tU of uranium, which corresponds to the consumption of 72 million years, also is included in the rock on the surface of the sea floor, and that is also recoverable as seawater uranium because uranium in seawater is in an equilibrium state with that. In other words, the uranium from seawater is almost inexhaustible natural resource. However, the recovery cost with current technology is still expensive compared with that of conventional uranium. Then, we assessed the effect of increase in uranium purchase cost on the entire electricity generation cost. In this study, the economy of electricity generation of cost of a commercial HTGR was evaluated with conventional uranium and seawater uranium. Compared with ordinary LWR using conventional uranium, HTGR can generate electricity cheaply because of small volume of simple direct gas turbine system compared with water and steam systems of LWR, rationalization by modularizing, and high thermal efficiency, even if fueled by seawater uranium. It is concluded that the HTGR

  17. Selection of fuel design for conversion and upgradation of Pakistan Research Reactor (PARR-1)

    International Nuclear Information System (INIS)

    Arshad, M.

    1991-01-01

    The Pakistan Research Reactor (PARR-1) is being converted from the use of Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel and its power is also being upgraded. In order to select new fuel for the converted and upgraded core ten different fuel element designs were analyzed and their relative performance was compared. Results of this study were later used to select appropriate design of the new fuel for PARR-1. This paper describes the computational methodology utilized for the analysis of various fuel element designs. Criteria for selecting the new fuel element are discussed and guidelines forming the selection basis of the new fuel design are given. (author)

  18. Consistent Set of Experiments from ICSBEP Handbook for Evaluation of Criticality Calculation Prediction of Apparatus of External Fuel Cycle with Different Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Golovko, Yury E. [FSUE ' SSC RF-IPPE' , 249033, Bondarenko Square 1, Obninsk (Russian Federation)

    2008-07-01

    Experiments with plutonium, low enriched uranium and uranium-233 from the ICSBEP1 Handbook are being considered in this paper. Among these experiments it was selected only those, which seem to be the most relevant to the evaluation of uncertainty of critical mass of mixtures of plutonium or low enriched uranium or uranium-233 with light water. All selected experiments were examined and covariance matrices of criticality uncertainties were developed along with some uncertainties were revised. Statistical analysis of these experiments was performed and some contradictions were discovered and eliminated. Evaluation of accuracy of prediction of criticality calculations was performed using the internally consistent set of experiments with plutonium, low enriched uranium and uranium-233 remained after the statistical analyses. The application objects for the evaluation of calculational prediction of criticality were water-reflected spherical systems of homogeneous aqueous mixtures of plutonium or low enriched uranium or uranium-233 of different concentrations which are simplified models of apparatus of external fuel cycle. It is shows that the procedure allows to considerably reduce uncertainty in k{sub eff} caused by the uncertainties in neutron cross-sections. Also it is shows that the results are practically independent of initial covariance matrices of nuclear data uncertainties. (authors)

  19. Critical review of uranium resources and production capability to 2020

    International Nuclear Information System (INIS)

    1998-08-01

    This report was prepared to assess the changing uranium supply and demand situation as well as the adequacy of uranium resources and the production capability to supply uranium concentrate to meet reactor demand through 2020. Uranium production has been meeting only 50 to 60 percent of the world requirements with the balance met from sale of excess inventory offered on the market at low prices. It is generally agreed by most specialists that the end of the excess inventory is approaching. With inventory no longer able to meet the production shortfall it is necessary to significantly expand uranium production to fill an increasing share of demand. Non-production supplies of uranium, such as the blending of highly enriched uranium (HEU) warheads to produce low enriched reactor fuel and reprocessing of spent fuel, are also expected to grow in importance as a fuel source. This analysis addresses three major concerns as follows: adequacy of resources to meet projected demand; adequacy of production capability to produce the uranium; and market prices to sustain production to fill demand. This analysis indicates uranium mine production to be the primary supply providing about 76 to 78 percent of cumulative needs through 2020. Alternative sources supplying the balance, in order of relative importance are: (1) low enriched uranium (LEU) blended from 500 tonnes of highly enriched uranium (HEU) Russian weapons, plus initial US Department of Energy (US DOE) stockpile sales (11 to 13%); (2) reprocessing of spent nuclear fuel (6%) and; (3) utility and Russian stockpiles. Further this report gives uranium production profiles by countries: CIS producers (Kazakhstan, Russian Federation, Ukraine, Uzbekistan) and other producers (Australia, Canada, China, Gabon, Mongolia, Namibia, Niger, South Africa, United States of America)

  20. Colloids from the aqueous corrosion of uranium nuclear fuel

    Science.gov (United States)

    Kaminski, M. D.; Dimitrijevic, N. M.; Mertz, C. J.; Goldberg, M. M.

    2005-12-01

    Colloids may enhance the subsurface transport of radionuclides and potentially compromise the long-term safe operation of the proposed radioactive waste repository at Yucca Mountain. Little data is available on colloid formation for the many different waste forms expected to be buried in the repository. This work expands the sparse database on colloids formed during the corrosion of metallic uranium nuclear fuel. We characterized spherical UO 2 and nickel-rich montmorilonite smectite-clay colloids formed during the corrosion of uranium metal fuel under bathtub conditions at 90 °C. Iron and chromium oxides and calcium carbonate colloids were present but were a minor population. The estimated upper concentration of the UO 2 and clays was 4 × 10 11 and 7 × 10 11-3 × 10 12 particles/L, respectively. However, oxygen eventually oxidized the UO 2 colloids, forming long filaments of weeksite K 2(UO 2) 2Si 6O 15 · 4H 2O that settled from solution, reducing the UO 2 colloid population and leaving predominantly clay colloids. The smectite colloids were not affected by oxygen. Plutonium was not directly observed within the UO 2 colloids but partitioned completely to the colloid size fraction. The plutonium concentration in the colloidal fraction was slightly higher than the value used in the viability assessment model, and does not change in concentration with exposure to oxygen. This paper provides conclusive evidence for single-phase radioactive colloids composed of UO 2. However, its impact on repository safety is probably small since oxygen and silica availability will oxidize and effectively precipitate the UO 2 colloids from concentrated solutions.

  1. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-09-20

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project.

  2. Nuclear fuel cycle head-end enriched uranium purification and conversion into metal

    International Nuclear Information System (INIS)

    Bonini, A.; Cabrejas, J.; Lio, L. de; Dell'Occhio, L.; Devida, C.; Dupetit, G.; Falcon, M.; Gauna, A.; Gil, D.; Guzman, G.; Neuringer, P.; Pascale, A.; Stankevicius, A.

    1998-01-01

    The CNEA (Comision Nacional de Energia Atomica - Argentina) operated two facilities at the Ezeiza Atomic Center which supply purified enriched uranium employed in the production of nuclear fuels. At one of those facilities, the Triple Height Laboratory scraps from the production of MTR type fuel elements (mainly out of specification U 3 O 8 plates or powder) are purified to nuclear grade. The purification is accomplished by a solvent extraction process. The other facility, the Enriched Uranium Laboratory produces 90% enriched uranium metal to be used in Mo 99 production (originally the uranium was used for the manufacture of MTR fuel elements made of aluminium-uranium alloy). This laboratory also provided metallic uranium with a lower enrichment (20%) for a first uranium-silicon testing fuel element, and in the near future it is going to recommence 20% enriched uranium related activities in order to provide the metal for the silicon-based fuel elements production (according to the policy of enrichment reduction for MTR reactors). (author)

  3. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    2000-01-01

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project

  4. Evaluation of the characteristics of uranium and plutonium Mixed Oxide (MOX) fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    MOX fuel irradiation test up to high burnup has been performed for five years. Irradiation test of MOX fuel having high plutonium content has also been performed from JFY 2007 and it still continues. A lot of irradiation data have been obtained through these tests. The activities done in JFY 2012 are mainly focused on Post Irradiation Examination (PIE) data analysis concerning thermal property change and fission gas release. In the former work thermal conductivity degradation due to burnup is examined and in the latter work the dependence of fission gas release mechanism on fuel pellet microstructure is examined. This report mainly covers the result of analysis. It is found that thermal conductivity degradation of MOX fuel due to burnup is less than that of UO{sub 2} fuel and that fission gas release mechanism of high enriched fissile zone (so called Pu spot) is much different from that of low enriched fissile zone (so called Matrix). (author)

  5. Uranium recovery from waste of the nuclear fuel cycle plants at IPEN-CNEN/SP, Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Freitas, Antonio A.; Ferreira, Joao C.; Zini, Josiane; Scapin, Marcos A.; Carvalho, Fatima Maria Sequeira de, E-mail: afreitas@ipen.b, E-mail: jcferrei@ipen.b, E-mail: jzini@ipen.b, E-mail: mascapin@ipen.b, E-mail: fatimamc@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    Sodium diuranate (DUS) is a uranium concentrate produced in monazite industry with 80% typical average grade of U{sup 3}O{sup 8}, containing sodium, silicon, phosphorus, thorium and rare earths as main impurities. Purification of such concentrate was achieved at the nuclear fuel cycle pilot plants of uranium at IPEN by nitric dissolution and uranium extraction into an organic phase using TBP/Varsol, while the aqueous phase retains impurities and a small quantity of non extracted uranium; both can be recovered later by precipitation with sodium hydroxide. Then the residual sodium diuranate goes to a long term storage at a safeguards deposit currently reaching 20 tonnes. This work shows how uranium separation and purification from such bulk waste can be achieved by ion exchange chromatography, aiming at decreased volume and cost of storage, minimization of environmental impacts and reduction of occupational doses. Additionally, the resulting purified uranium can be reused in nuclear fuel cycle.(author)

  6. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    Recent decisions by the Australian Government will ensure a significant expansion of the uranium industry. Development at Roxby Downs may proceed and Ranger may fulfil two new contracts but the decision specifies that apart from Roxby Downs, no new mines should be approved. The ACTU maintains an anti-uranium policy but reaction to the decision from the trade union movement has been muted. The Australian Science and Technology Council (ASTEC) has been asked by the Government to conduct an inquiry into a number of issues relating to Australia's role in the nuclear fuel cycle. The inquiry will examine in particular Australia's nuclear safeguards arrangements and the adequacy of existing waste management technology. In two additional decisions the Government has dissociated itself from a study into the feasibility of establishing an enrichment operation and has abolished the Uranium Advisory Council. Although Australian reserves account for 20% of the total in the Western World, Australia accounts for a relatively minor proportion of the world's uranium production

  7. Analysis of the Reuse of Uranium Recovered from the Reprocessing of Commercial LWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    DelCul, Guillermo Daniel [ORNL; Trowbridge, Lee D [ORNL; Renier, John-Paul [ORNL; Ellis, Ronald James [ORNL; Williams, Kent Alan [ORNL; Spencer, Barry B [ORNL; Collins, Emory D [ORNL

    2009-02-01

    This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reactor). Reuse is an attractive alternative to the current Advanced Fuel Cycle Initiative (AFCI) Global Nuclear Energy Partnership (GNEP) baseline plan, which stores the reprocessed uranium (RU) for an uncertain future or attempts to dispose of it as 'greater-than-Class C' waste. Considering that the open fuel cycle currently deployed in the United States already creates a huge excess quantity of depleted uranium, the closed fuel cycle should enable the recycle of the major components of spent fuel, such as the uranium and the hazardous, long-lived transuranic (TRU) actinides, as well as the managed disposal of fission product wastes. Compared with the GNEP baseline scenario, the reuse of RU in the uranium fuel cycle has a number of potential advantages: (1) avoidance of purchase costs of 11-20% of the natural uranium feed; (2) avoidance of disposal costs for a large majority of the volume of spent fuel that is reprocessed; (3) avoidance of disposal costs for a portion of the depleted uranium from the enrichment step; (4) depending on the {sup 235}U assay of the RU, possible avoidance of separative work costs; and (5) a significant increase in the production of {sup 238}Pu due to the presence of {sup 236}U, which benefits somewhat the transmutation value of the plutonium and also provides some proliferation resistance.

  8. Analysis of the Reuse of Uranium Recovered from the Reprocessing of Commercial LWR Spent Fuel

    International Nuclear Information System (INIS)

    DelCul, Guillermo D.; Trowbridge, Lee D.; Renier, John-Paul; Ellis, Ronald James; Williams, Kent Alan; Spencer, Barry B.; Collins, Emory D.

    2009-01-01

    This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reactor). Reuse is an attractive alternative to the current Advanced Fuel Cycle Initiative (AFCI) Global Nuclear Energy Partnership (GNEP) baseline plan, which stores the reprocessed uranium (RU) for an uncertain future or attempts to dispose of it as 'greater-than-Class C' waste. Considering that the open fuel cycle currently deployed in the United States already creates a huge excess quantity of depleted uranium, the closed fuel cycle should enable the recycle of the major components of spent fuel, such as the uranium and the hazardous, long-lived transuranic (TRU) actinides, as well as the managed disposal of fission product wastes. Compared with the GNEP baseline scenario, the reuse of RU in the uranium fuel cycle has a number of potential advantages: (1) avoidance of purchase costs of 11-20% of the natural uranium feed; (2) avoidance of disposal costs for a large majority of the volume of spent fuel that is reprocessed; (3) avoidance of disposal costs for a portion of the depleted uranium from the enrichment step; (4) depending on the 235 U assay of the RU, possible avoidance of separative work costs; and (5) a significant increase in the production of 238 Pu due to the presence of 236 U, which benefits somewhat the transmutation value of the plutonium and also provides some proliferation resistance

  9. Comparison of the radiological impacts of thorium and uranium nuclear fuel cycles

    International Nuclear Information System (INIS)

    Meyer, H.R.; Witherspoon, J.P.; McBride, J.P.; Frederick, E.J.

    1982-03-01

    This report compares the radiological impacts of a fuel cycle in which only uranium is recycled, as presented in the Final Generic Environmental Statement on the Use of Recycle Plutonium in Mixed Oxide Fuel in Light Water Cooled Reactors (GESMO), with those of the light-water breeder reactor (LWBR) thorium/uranium fuel cycle in the Final Environmental Statement, Light Water Breeder Reactor Program. The significant offsite radiological impacts from routine operation of the fuel cycles result from the mining and milling of thorium and uranium ores, reprocessing spent fuel, and reactor operations. The major difference between the impacts from the two fuel cycles is the larger dose commitments associated with current uranium mining and milling operations as compared to thorium mining and milling. Estimated dose commitments from the reprocessing of either fuel type are small and show only moderate variations for specific doses. No significant differences in environmental radiological impact are anticipated for reactors using either of the fuel cycles. Radiological impacts associated with routine releases from the operation of either the thorium or uranium fuel cycles can be held to acceptably low levels by existing regulations

  10. 78 FR 23312 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Science.gov (United States)

    2013-04-18

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National Enrichment Facility, Eunice, New Mexico..., National Enrichment Facility in Eunice, New Mexico, and has authorized the introduction of uranium...

  11. High enrichment to low enrichment core's conversion. Accidents analysis

    International Nuclear Information System (INIS)

    Abbate, P.; Rubio, R.; Doval, A.; Lovotti, O.

    1990-01-01

    This work analyzes the different accidents that may occur in the reactor's facility after the 20% high-enriched uranium core's conversion. The reactor (of 5 thermal Mw), built in the 50's and 60's, is of the 'swimming pool' type, with light water and fuel elements of the curve plates MTR type, enriched at 93.15 %. This analysis includes: a) accidents by reactivity insertion; b) accidents by coolant loss; c) analysis by flow loss and d) fission products release. (Author) [es

  12. Analysis of alternative light water reactor (LWR) fuel cycles

    International Nuclear Information System (INIS)

    Heeb, C.M.; Aaberg, R.L.; Boegel, A.J.; Jenquin, U.P.; Kottwitz, D.A.; Lewallen, M.A.; Merrill, E.T.; Nolan, A.M.

    1979-12-01

    Nine alternative LWR fuel cycles are analyzed in terms of the isotopic content of the fuel material, the relative amounts of primary and recycled material, the uranium and thorium requirements, the fuel cycle costs and the fraction of energy which must be generated at secured sites. The fuel materials include low-enriched uranium (LEU), plutonium-uranium (MOX), highly-enriched uranium-thorium (HEU-Th), denatured uranium-thorium (DU-Th) and plutonium-thorium (Pu-Th). The analysis is based on tracing the material requirements of a generic pressurized water reactor (PWR) for a 30-year period at constant annual energy output. During this time period all the created fissile material is recycled unless its reactivity worth is less than 0.2% uranium enrichment plant tails

  13. Determination of uranium metal concentration in irradiated fuel storage basin sludge using selective dissolution

    International Nuclear Information System (INIS)

    Delegard, C.H.; Sinkov, S.I.; Chenault, J.W.; Schmidt, A.J.; Pool, K.N.; Welsh, T.L.

    2014-01-01

    Irradiated uranium metal fuel was stored underwater in the K East and K West storage basins at the US Department of Energy Hanford Site. The uranium metal under damaged cladding reacted with water to generate hydrogen gas, uranium oxides, and spalled uranium metal particles which intermingled with other particulates to form sludge. While the fuel has been removed, uranium metal in the sludge remains hazardous. An expeditious routine method to analyze 0.03 wt% uranium metal in the presence of >30 wt% total uranium was needed to support safe sludge management and processing. A selective dissolution method was designed based on the rapid uranium oxide dissolution but very low uranium metal corrosion rates in hot concentrated phosphoric acid. The uranium metal-bearing heel from the phosphoric acid step then is rinsed before the uranium metal is dissolved in hot concentrated nitric acid for analysis. Technical underpinnings of the selective dissolution method, including the influence of sludge components, were investigated to design the steps and define the reagents, quantities, concentrations, temperatures, and times within the selective dissolution analysis. Tests with simulant sludge proved the technique feasible. Tests with genuine sludge showed a 0.0028 ± 0.0037 wt% (at one standard deviation) uranium metal analytical background, a 0.011 wt% detection limit, and a 0.030 wt% quantitation limit in settled (wet) sludge. In tests using genuine K Basin sludge spiked with uranium metal at concentrations above the 0.030 wt% ± 25 % (relative) quantitation limit, uranium metal recoveries averaged 99.5 % with a relative standard deviation of 3.5 %. (author)

  14. Neutronic performance of a 14 MW TRIGA reactor: LEU vs HEU fuel

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Snelgrove, J.L.; Cornella, R.J.

    1983-01-01

    A primary objective of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is to develop means for replacing, wherever possible, currently used highly-enriched uranium (HEU) fuel ( 235 U enrichment > 90%) with low-enriched uranium (LEU) fuel ( 235 U enrichment < 20%) without significantly degrading the performance of research and test reactors. The General Atomic Company has developed a low-enriched but high uranium content Er-U-ZrH/sub 1.6/ fuel to enable the conversion of TRIGA reactors (and others) from HEU to LEU. One possible application is to the water-moderated 14 MW TRIGA Steady State Reactor (SSR) at the Romanian Institute for Nuclear Power Reactors. The work reported here was undertaken for the purpose of comparing the neutronic performance of the SSR for HEU fuel with that for LEU fuel. In order to make these relative comparisons as valid as possible, identical methods and models were used for the neutronic calculations

  15. Uranium/fuel cycle 74, New Orleans, Louisiana, 17--20 March 1974. Program report

    International Nuclear Information System (INIS)

    1974-01-01

    The highlight of papers presented at the conference are summarized. The sessions covered uranium raw material, transportation of spent fuel and radioactive waste, plutonium recycle, waste management, and safeguards. (U.S.)

  16. Fuel assemblies for use in nuclear reactors

    International Nuclear Information System (INIS)

    Mochida, Takaaki.

    1987-01-01

    Purpose: To increase the plutonium utilization amount and improve the uranium-saving effect in the fuel assemblies of PWR type reactor using mixed uranium-plutonium oxides. Constitution: MOX fuel rods comprising mixed plutonium-uranium oxides are disposed to the outer circumference of a fuel assembly and uranium fuel rods only composed of uranium oxides are disposed to the central portion thereof. In such a fuel assembly, since the uranium fuel rods are present at the periphery of the control rod, the control rod worth is the same as that of the uranium fuel assembly in the prior art. Further, since about 25 % of the entire fuel rods is composed of the MOX fuel rods, the plutonium utilization amount is increased. Further, since the MOX fuel rods at low enrichment degree are present at the outer circumferential portion, mismatching at the boundary to the adjacent MOX fuel assembly is reduced and the problem of local power peaking increase in the MOX fuel assembly is neither present. (Kamimura, M.)

  17. Uranium Enrichment Determination of the InSTEC Sub Critical Ensemble Fuel by Gamma Spectrometry

    International Nuclear Information System (INIS)

    Borrell Munnoz, Jose L.; LopezPino, Neivy; Diaz Rizo, Oscar; D'Alessandro Rodriguez, Katia; Padilla Cabal, Fatima; Arbelo Penna, Yunieski; Garcia Rios, Aczel R.; Quintas Munn, Ernesto L.; Casanova Diaz, Amaya O.

    2009-01-01

    Low background gamma spectrometry was applied to analyze the uranium enrichment of the nuclear fuel used in the InSTEC Sub Critical ensemble. The enrichment was calculated by two variants: an absolute method using the Monte Carlo method to simulated detector volumetric efficiency, and an iterative procedure without using standard sources. The results confirm that the nuclear fuel of the ensemble is natural uranium without any additional degree of enrichment. (author)

  18. Disposition of surplus highly enriched uranium: Draft environmental impact statement

    International Nuclear Information System (INIS)

    1995-10-01

    This document assesses the environmental impacts at four potential sites that may result from alternatives for the disposition of United States-origin weapons-usable highly enriched uranium (HEU) that has been or may be declared surplus to national defense or defense-related program needs. In addition to the no action alternative, it assesses four alternatives that would eliminate the weapons-usability of HEU by blending it with depleted uranium, natural uranium, or low-enriched uranium (LEU) to create low-enriched uranium, either as commercial reactor fuel feedstock or as low-level radioactive waste. The potential blending sites are DOE's Y-12 Plant at Oak Ridge Reservation in Oak Ridge, Tennessee; DOE's Savannah River Site in Aiken, South Carolina; the Babcock ampersand Wilcox Naval Nuclear Fuel Division Facility in Lynchburg, Virginia; and the Nuclear Fuel Services Fuel Fabrication Plant in Erwin, Tennessee. Evaluations of impacts on site infrastructure, water resources, air quality and noise, socioeconomic resources, waste management, public and occupational health, and environmental justice for the potential blending sites are included in the assessment. The intersite transportation of nuclear and hazardous materials is also assessed. The preferred alternative is to blend down surplus HEU to LEU for maximum commercial use as reactor fuel feed which would likely be done at a combination of DOE and commercial sites

  19. Plutonium recovery from spent reactor fuel by uranium displacement

    Science.gov (United States)

    Ackerman, J.P.

    1992-03-17

    A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

  20. Plutonium recovery from spent reactor fuel by uranium displacement

    International Nuclear Information System (INIS)

    Ackerman, J.P.

    1992-01-01

    A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished

  1. Uranium separation from phosphates and the fuel cycle process

    International Nuclear Information System (INIS)

    Lavi, J.

    1978-01-01

    A short introduction on the recycle of uranium and plutonium is presented. The uranium world market at present, the prices during the last few years, the actual requirements and those for the years 1978-1983 are given. In a special paragraph the present resources of uranium in Israel as well as the extraction possibilities are discussed. (B.G.)

  2. Status of Uranium Atomic Vapor Laser Isotope Separation Program

    International Nuclear Information System (INIS)

    Chen, Hao-Lin; Feinberg, R.M.

    1993-06-01

    This report discusses demonstrations of plant-scale hardware embodying AVLIS technology which were completed in 1992. These demonstrations, designed to provide key economic and technical bases for plant deployment, produced significant quantities of low enriched uranium which could be used for civilian power reactor fuel. We are working with industry to address the integration of AVLIS into the fuel cycle. To prepare for deployment, a conceptual design and cost estimate for a uranium enrichment plant were also completed. The U-AVLIS technology is ready for commercialization

  3. Development of high uranium-density fuels for use in research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ugajin, Mitsuhiro; Akabori, Mitsuo; Itoh, Akinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-02-01

    The uranium silicide U{sub 3}Si{sub 2} possesses uranium density 11.3 gU/cm{sup 3} with a congruent melting point of 1665degC, and is now successfully in use as a research reactor fuel. Another uranium silicide U{sub 3}Si and U{sub 6}Me-type uranium alloys (Me=Fe,Mn,Ni) have been chosen as new fuel materials because of the higher uranium densities 14.9 and 17.0 gU/cm{sup 3}, respectively. Experiments were carried out to fabricate miniature aluminum-dispersion plate-type and aluminum-clad disk-type fuels by using the conventional picture-frame method and a hot-pressing technique, respectively. These included the above-mentioned new fuel materials as well as U{sub 3}Si{sub 2}. Totally 14 miniplates with uranium densities from 4.0 to 6.3 gU/cm{sup 3} of fuel meat were prepared together with 28 disk-type fuel containing structurally-modified U{sub 3}Si, and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Some results of postirradiation examinations are presented. (author)

  4. Development of high uranium-density fuels for use in research reactors

    International Nuclear Information System (INIS)

    Ugajin, Mitsuhiro; Akabori, Mitsuo; Itoh, Akinori

    1996-01-01

    The uranium silicide U 3 Si 2 possesses uranium density 11.3 gU/cm 3 with a congruent melting point of 1665degC, and is now successfully in use as a research reactor fuel. Another uranium silicide U 3 Si and U 6 Me-type uranium alloys (Me=Fe,Mn,Ni) have been chosen as new fuel materials because of the higher uranium densities 14.9 and 17.0 gU/cm 3 , respectively. Experiments were carried out to fabricate miniature aluminum-dispersion plate-type and aluminum-clad disk-type fuels by using the conventional picture-frame method and a hot-pressing technique, respectively. These included the above-mentioned new fuel materials as well as U 3 Si 2 . Totally 14 miniplates with uranium densities from 4.0 to 6.3 gU/cm 3 of fuel meat were prepared together with 28 disk-type fuel containing structurally-modified U 3 Si, and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Some results of postirradiation examinations are presented. (author)

  5. The nuclear fuel cycle, From the uranium mine to waste disposal

    International Nuclear Information System (INIS)

    2002-09-01

    Fuel is a material that can be burnt to provide heat. The most familiar fuels are wood, coal, natural gas and oil. By analogy, the uranium used in nuclear power plants is called 'nuclear fuel', because it gives off heat too, although, in this case, the heat is obtained through fission and not combustion. After being used in the reactor, spent nuclear fuel can be reprocessed to extract recyclable energy material, which is why we speak of the nuclear fuel cycle. This cycle includes all the following industrial operations: - uranium mining, - fuel fabrication, - use in the reactor, - reprocessing the fuel unloaded from the reactor, - waste treatment and disposal. 'The nuclear fuel cycle includes an array of industrial operations, from uranium mining to the disposal of radioactive waste'. Per unit or mass (e.g. per kilo), nuclear fuel supplies far more energy than a fossil fuel (coal or oil). When used in a pressurised water reactor, a kilo of uranium generates 10,000 times more energy than a kilo of coal or oil in a conventional power station. Also, the fuel will remain in the reactor for a long time (several years), unlike conventional fuels, which are burnt up quickly. Nuclear fuel also differs from others in that uranium has to undergo many processes between the time it is mined and the time it goes into the reactor. For the sake of simplicity, the following pages will only look at nuclear fuel used in pressurised water reactors (or PWRs), because nuclear power plants consisting of one or more PWRs are the most widely used around the world. (authors)

  6. Reoxidation of uranium in electrolytically reduced simulated oxide fuel during residual salt distillation

    International Nuclear Information System (INIS)

    Eun-Young Choi; Jin-Mok Hur; Min Ku Jeon; University of Science and Technology, Yuseong-gu, Daejeon

    2017-01-01

    We report that residual salt removal by high-temperature distillation causes partial reoxidation of uranium metal to uranium oxide in electrolytically reduced simulated oxide fuel. Specifically, the content of uranium metal in the above product decreases with increasing distillation temperatures, which can be attributed to reoxidation by Li 2 O contained in residual salt (LiCl). Additionally, we estimate the fractions of Li 2 O reacted with uranium metal under these conditions, showing that they decrease with decreasing temperature, and calculate some thermodynamic parameters of the above reoxidation. (author)

  7. Isotopic composition and radiological properties of uranium in selected fuel cycles

    International Nuclear Information System (INIS)

    Fleischman, R.M.; Liikala, R.C.

    1975-04-01

    Three major topic areas are discussed: First, the properties of the uranium isotopes are defined relative to their respective roles in the nuclear fuel cycle. Secondly, the most predominant fuel cycles expected in the U. S. are described. These are the Light Water Reactor (LWR), High Temperature Gas Cooled Reactor (HTGR), and Liquid Metal Fast Breeder Reactor (LMFBR) fuel cycles. The isotopic compositions of uranium and plutonium fuels expected for these fuel cycles are given in some detail. Finally the various waste streams from these fuel cycles are discussed in terms of their relative toxicity. Emphasis is given to the high level waste streams from reprocessing of spent fuel. Wastes from the various fuel cycles are compared based on projected growth patterns for nuclear power and its various components. (U.S.)

  8. Analysis of fuel cycles with natural uranium; Analiza gorivnih ciklusa sa prirodnim uranom

    Energy Technology Data Exchange (ETDEWEB)

    Stojanovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-05-15

    A method was developed and a computer code was written for analysis of fuel cycles and it was applied for heavy water and graphite moderated power reactors. Among a variety of possibilities, three methods which enable best utilization of natural uranium and plutonium production were analyzed. Analysis has shown that reprocessing of irradiated uranium and plutonium utilization in the same or similar type of reactor could increase significantly utilization of natural uranium. Increase of burnup is limited exclusively by costs of reprocessing, plutonium extraction and fabrication of new fuel elements.

  9. Development of molecular dynamics potential for uranium silicide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Jianguo; Zhang, Yongfeng; Hales, Jason D.

    2016-09-01

    Use of uranium–silicide (U-Si) in place of uranium dioxide (UO2) is one of the promising concepts being proposed to increase the accident tolerance of nuclear fuels. This is due to a higher thermal conductivity than UO2 that results in lower centerline temperatures. U-Si also has a higher fissile density, which may enable some new cladding concepts that would otherwise require increased enrichment limits to compensate for their neutronic penalty. However, many critical material properties for U-Si have not been determined experimentally. For example, silicide compounds (U3Si2 and U3Si) are known to become amorphous under irradiation. There was clear independent experimental evidence to support a crystalline to amorphous transformation in those compounds. However, it is still not well understood how the amorphous transformation will affect on fuel behavior. It is anticipated that modeling and simulation may deliver guidance on the importance of various properties and help prioritize experimental work. In order to develop knowledge-based models for use at the engineering scale with a minimum of empirical parameters and increase the predictive capabilities of the developed model, inputs from atomistic simulations are essential. First-principles based density functional theory (DFT) calculations will provide the most reliable information. However, it is probably not possible to obtain kinetic information such as amorphization under irradiation directly from DFT simulations due to size and time limitations. Thus, a more feasible way may be to employ molecular dynamics (MD) simulation. Unfortunately, so far no MD potential is available for U-Si to discover the underlying mechanisms. Here, we will present our recent progress in developing a U-Si potential from ab initio data. This work is supported by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program funded by the U.S. Department of Energy, Office of Nuclear Energy.

  10. Uranium-236 as an indicator of fuel-cycle uranium in ground water

    International Nuclear Information System (INIS)

    Jaquish, R.E.

    1989-08-01

    Environmental monitoring on and around the Hanford Site includes regular sampling of onsite monitoring wells and offsite farm wells. Uranium has been identified in the ground water onsite and also in water from farm wells located on the east side of the Columbia River, across from the Hanford Site. Information on the hydrology of the area indicates that the source of the offsite uranium is not the Hanford Site. This study evaluated the isotopic composition of the uranium in water from the various wells to differentiate the onsite uranium contamination from natural uranium offsite. 5 refs., 2 figs., 2 tabs

  11. Micromechanical approach of behavior of uranium dioxide nuclear fuel

    International Nuclear Information System (INIS)

    Soulacroix, Julian

    2014-01-01

    Uranium dioxide (UO 2 ) is the reference fuel for pressurized water nuclear reactors. Our study deals with understanding and modeling of mechanical behavior at the microstructure scale at low temperatures (brittle fracture) and high temperature (viscoplastic strain). We have first studied the geometrical properties of polycrystals at large and of UO 2 polycrystal more specifically. As of now, knowledge of this behavior in the brittle fracture range is limited. Consequently, we developed an experimental method which allows better understanding of brittle fracture phenomenon at grain scale. We show that fracture is fully intra-granular and {100} planes seem to be the most preferential cleavage planes. Experimental results are directly used to deduce constitutive equations of intra-granular brittle fracture at crystal scale. This behavior is then used in 3D polycrystal simulation of brittle fracture. The full field calculation gives access to the initiation of fracture and propagation of the crack through the grains. Finally, we developed a mechanical behavior model of UO 2 in the viscoplastic range. We first present constitutive equations at macroscopic scale which accounts for an ageing process caused by migration of defects towards dislocations. Secondly, we have developed a crystal plasticity model which was fitted to UO 2 . This model includes the rotation of the crystal lattice. We present examples of polycrystalline simulations. (author) [fr

  12. Quantitative determination of uranium distribution homogeneity in MTR fuel type plates

    International Nuclear Information System (INIS)

    Ferrufino, Felipe Bonito Jaldin

    2011-01-01

    IPEN/CNEN-SP produces the fuel to supply its nuclear research reactor IEA-R1. The fuel is assembled with fuel plates containing an U 3 Si 2 -Al composite meat. A good homogeneity in the uranium distribution inside the fuel plate meat is important from the standpoint of irradiation performance. Considering the lower power of reactor IEA-R1, the uranium distribution in the fuel plate has been evaluated only by visual inspection of radiographs. However, with the possibility of IPEN to manufacture the fuel for the new Brazilian Multipurpose Reactor (RMB), with higher power, it urges to develop a methodology to determine quantitatively the uranium distribution into the fuel. This paper presents a methodology based on X-ray attenuation, in order to quantify the uranium concentration distribution in the meat of the fuel plate by using optical densities in radiographs and comparison with standards. The results demonstrated the inapplicability of the method, considering the current specification for the fuel plates due to the high intrinsic error to the method. However, the study of the errors involved in the methodology, seeking to increase their accuracy and precision, can enable the application of the method to qualify the final product. (author)

  13. Future of uranium enrichment

    International Nuclear Information System (INIS)

    Hosmer, C.

    1981-01-01

    The increasing amount of separative work being done in government facilities to produce low-enriched uranium fuel for nuclear utilities again raises the question: should this business-type, industrial function be burned over the private industry. The idea is being looked at by the Reagan administration, but faces problems of national security as well as from the unique nature of the business. This article suggests that a joint government-private venture combining enriching, reprocessing, and waste disposal could be the answer. Further, a separate entity using advanced laser technology to deplete existing uranium tails and lease them for fertile blankets in breeder reactors might earn substantial revenues to help reduce the national debt

  14. Improved capability of U-ZrH fuels

    Energy Technology Data Exchange (ETDEWEB)

    Gietzen, A J [General Atomic Company, San Diego, CA (United States)

    1983-08-01

    This paper provides a brief background on TRIGA fuels and a summary of the development of U-ZrH fuels utilizing higher uranium loadings in the alloy. Most of the development work was reported two years ago; however, in this paper some emphasis will be placed upon applications that General Atomic Company is now making with these higher alloy fuels. They are referred to as higher uranium content alloys because their development was not really LEU development; the TRIGA fuel has used low enrichment since its inception.

  15. Control rod studies for alternative fuel cycles in the GA 1160 MW(e) high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Neef, H. J.

    1975-06-15

    The control system, which is investigated in this paper for both the low enriched uranium high enriched uranium/thorium fuel cycles, has been developed to control the General Atomics (GA) thorium fuel cycle 1160 MW(e) reactor. It has been shown in this investigation that its effectiveness in the low enriched and subsequent thorium cycle switch-over reactor is equivalent to the effectiveness in the thorium cycle. The shutdown margin in the low enriched core is even higher compared to the thorium core, mainly due to the presence of Pa-233 in the thorium cycle. As long as the fuel cycle for the thorium cycle is not closed with the recycling of U-233, the low enriched cycle will offer an attractive alternative. It was found that the GA 1160 MW(e) control system has enough built-in control rod capacity to accommodate the low enriched uranium cycle and to perform a later switch-over to a thorium-based fuel cycle.

  16. Method to manufacture a nuclear fuel from uranium-plutonium monocarbide or uranium-plutonium mononitride

    International Nuclear Information System (INIS)

    Krauth, A.; Mueller, N.

    1977-01-01

    Pure uranium carbide or nitride is converted with plutonium oxide and carbon (all in powder form) to uranium-plutonium monocarbide or mononitride by cold pressing and sintering at about 1600 0 C. Pure uranium carbide or uranium nitride powder is firstly prepared without extensive safety measures. The pure uranium carbide or nitride powder can also be inactivated by using chemical substances (e.g. stearic acid) and be handled in air. The sinterable uranium carbide or nitride powder (or also granulate) is then introduced into the plutonium line and mixed with a nonstoichiometrically adjusted, prereacted mixture of plutonium oxide and carbon, pressed to pellets and reaction sintered. The surface of the uranium-plutonium carbide (higher metal content) can be nitrated towards the end of the sinter process in a stream of nitrogen. The protective layer stabilizes the carbide against the water and oxygen content in air. (IHOE) [de

  17. Thermal-hydraulic calculations for KUHFR with reduced enrichment uranium fuel

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Shibata, Toshikazu.

    1982-01-01

    This report provides the preliminary results of the thermal-hydraulic calculations to study the safety aspects in fueling the KUHFR with reduced enrichment uranium. The calculations were based on what was outlined in the Safety Analysis Report for the KUHFR and the guidebook for research reactor core conversion, IAEA-TECDOC-233, published by the International Atomic Energy Agency. No significant differences in the thermal-hydraulic operating conditions have been found between HEU and MEU fuels. However, in LEU cases, the combination of three factors - larger power peaking with LEU fuel, smaller thermal conductivity of U 3 O 8 -Al fuel with high uranium densities, and thicker fuel meat - resulted in higher maximum fuel and surface temperatures with the LEU oxide fuel. (author)

  18. Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

    1986-09-01

    Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed

  19. Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures

    International Nuclear Information System (INIS)

    Oliveira, Fabio Branco Vaz de

    2008-01-01

    This work has as its objective the development of a high density and low enriched nuclear fuel based on the gamma-UMo alloys, for utilization where it is necessary satisfactory behavior in high temperatures, considering its utilization as dispersion. For its accomplishment, it was started from the analysis of the RERTR ('Reduced Enrichment for Research and Test Reactors') results and some theoretical works involving the fabrication of gamma-uranium metastable alloys. A ternary addition is proposed, supported by the properties of binary and ternary uranium alloys studied, having the objectives of the gamma stability enhancement and an ease to its powder fabrication. Alloys of uranium-molybdenum were prepared with 5 to 10% Mo addition, and 1 and 3% of ternary, over a gamma U7Mo binary base alloy. In all the steps of its preparation, the alloys were characterized with the traditional techniques, to the determination of its mechanical and structural properties. To provide a process for the alloys powder obtention, its behavior under hydrogen atmosphere were studied, in thermo analyser-thermo gravimeter equipment. Temperatures varied from the ambient up to 1000 deg C, and times from 15 minutes to 16 hours. The results validation were made in a semi-pilot scale, where 10 to 50 g of powders of some of the alloys studied were prepared, under static hydrogen atmosphere. Compatibility studies were conducted by the exposure of the alloys under oxygen and aluminum, to the verification of possible reactions by means of differential thermal analysis. The alloys were exposed to a constant heat up to 1000 deg C, and their performances were evaluated in terms of their reaction resistance. On the basis of the results, it was observed that ternary additions increases the temperatures of the reaction with aluminum and oxidation, in comparison with the gamma UMo binaries. A set of conditions to the hydration of the alloys were defined, more restrictive in terms of temperature, time and

  20. PROCESSING OF URANIUM-METAL-CONTAINING FUEL ELEMENTS

    Science.gov (United States)

    Moore, R.H.

    1962-10-01

    A process is given for recovering uranium from neutronbombarded uranium- aluminum alloys. The alloy is dissolved in an aluminum halide--alkali metal halide mixture in which the halide is a mixture of chloride and bromide, the aluminum halide is present in about stoichiometric quantity as to uranium and fission products and the alkali metal halide in a predominant quantity; the uranium- and electropositive fission-products-containing salt phase is separated from the electronegative-containing metal phase; more aluminum halide is added to the salt phase to obtain equimolarity as to the alkali metal halide; adding an excess of aluminum metal whereby uranium metal is formed and alloyed with the excess aluminum; and separating the uranium-aluminum alloy from the fission- productscontaining salt phase. (AEC)

  1. High-Uranium-Loaded U3O8-Al fuel element development program. Part 1

    International Nuclear Information System (INIS)

    Martin, M.M.

    1993-01-01

    The High-Uranium-Loaded U 3 O 8 -Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U 3 O 8 -Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U 3 O 8 -Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U 3 O 8 ). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % U involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U 3 O 8 and aluminum

  2. The improvement of technology for high-uranium-density Al-base dispersion fuel plates

    International Nuclear Information System (INIS)

    Shouhui, Dai; Rongxian, Sun; Hejian, Mao; Baosheng, Zhao; Changgen, Yin

    1987-01-01

    An improved rolling process was developed for manufacturing Al-base dispersion fuel plates. When the fuel content in the meat increased up to 50 vol%, the non-uniformity of uranium is not more than ± 7.2%, and the minimum cladding thickness is not less than 0.32 mm. (Author)

  3. Present situation of unused uranium fuel in Tokyo Institute of Technology

    International Nuclear Information System (INIS)

    Obara, T.; Ogawa, M.

    2008-01-01

    Present situation of unused enriched uranium fuel in Tokyo Institute of Technology is described. The fuels were for sub-critical experiments. There is no special facility for transportation in the site. But there is no technical problem for it. One of the important issues to be done is a duty by national regulation against nuclear disaster. (author)

  4. Mixed Uranium/Refractory Metal Carbide Fuels for High Performance Nuclear Reactors

    International Nuclear Information System (INIS)

    Knight, Travis; Anghaie, Samim

    2002-01-01

    Single phase, solid-solution mixed uranium/refractory metal carbides have been proposed as an advanced nuclear fuel for advanced, high-performance reactors. Earlier studies of mixed carbides focused on uranium and either thorium or plutonium as a fuel for fast breeder reactors enabling shorter doubling owing to the greater fissile atom density. However, the mixed uranium/refractory carbides such as (U, Zr, Nb)C have a lower uranium densities but hold significant promise because of their ultra-high melting points (typically greater than 3700 K), improved material compatibility, and high thermal conductivity approaching that of the metal. Various compositions of (U, Zr, Nb)C were processed with 5% and 10% metal mole fraction of uranium. Stoichiometric samples were processed from the constituent carbide powders, while hypo-stoichiometric samples with carbon-to-metal (C/M) ratios of 0.92 were processed from uranium hydride, graphite, and constituent refractory carbide powders. Processing techniques of cold uniaxial pressing, dynamic magnetic compaction, sintering, and hot pressing were investigated to optimize the processing parameters necessary to produce high density (low porosity), single phase, solid-solution mixed carbide nuclear fuels for testing. This investigation was undertaken to evaluate and characterize the performance of these mixed uranium/refractory metal carbides for high performance, ultra-safe nuclear reactor applications. (authors)

  5. DUSCOBS - a depleted-uranium silicate backfill for transport, storage, and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Pope, R.B.; Ashline, R.C.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

    1995-01-01

    A Depleted Uranium Silicate COntainer Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside storage, transport, and repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill all void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (1) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (2) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. In addition, the DUSCOBS improves the integrity of the package by acting as a packing material and ensures criticality control for the package during SNF storage and transport. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments

  6. Management and Handling of Rejected Fuel of MTR Type and Process Effluents Contained Uranium at FEPI

    International Nuclear Information System (INIS)

    Ghaib Widodo; Bambang Herutomo

    2007-01-01

    Research Reactor Fuel Element Production Installation (FEPI) - Serpong has performed management and handling of all kinds of rejected fuel material during production (solids, liquids, and gases) and process effluents contained uranium. The methods that has been implemented are precipitation, absorption, evaporation, electrolysis, and electrodialysis. By these methods will finally be obtained forms of product which can be used directly as fuel material feed and solid/liquid radioactive waste that fulfil the requirements (uranium contents < 50 ppm) to be send to Radioactive Waste Management Installation. (author)

  7. Experience in the development of metal uranium-base nuclear fuel for heavy-water gas-cooled reactors

    International Nuclear Information System (INIS)

    Ashikhmin, V.P.; Vorob'ev, M.A.; Gusarov, M.S.; Davidenko, A.S.; Zelenskij, V.F.; Ivanov, V.E.; Krasnorutskij, V.S.; Petel'guzov, I.A.; Stukalov, A.I.

    1978-01-01

    Investigations were carried out to solve the problem of making the development of radiation-resistant uranium fuel for power reactors including the heavy-water gas-cooled KS-150 reactor. Factors are considered that limit the lifetime of uranium fuel elements, and the ways of suppressing them are discussed. Possible reasons of the insufficient radiation resistance of uranium rod fuel element and the progress attained are analyzed. Some general problems on the fuel manufacture processes are discussed. The main results are presented on the operation of the developed fuel in research reactor loops and the commercial heavy-water KS-150 reactor. The results confirm an exceptionally high radiation resistance of fuel to burn-ups of 1.5-2%. The successful solution of a large number of problems associated with the development of metal uranium fuel provides for new possibilities of using metal uranium in power reactors

  8. Uranium-236 in light water reactor spent fuel recycled to an enriching plant

    International Nuclear Information System (INIS)

    de la Garza, A.

    1977-01-01

    The introduction of 236 U to an enriching plant by recycling spent fuel uranium results in enriched products containing 236 U, a parasitic neutron absorber in reactor fuel. Convenient approximate methodology determines 235 236 U, and total uranium flowsheets with associated separative work requirements in enriching plant operations for use by investigators of the light water reactor fuel cycle not having recourse to specialized multicomponent cascade technology. Application of the methodology has been made to compensation of an enriching plant product for 236 U content and to the value at an enriching plant of spent fuel uranium. The approximate methodology was also confirmed with more exact calculations and with some experience with 236 U in an enriching plant

  9. Back-end fuel cycle efficiencies with respect to improved uranium utilization

    International Nuclear Information System (INIS)

    Kuczera, B.; Hennies, H.H.

    1983-01-01

    The world-wide nuclear power plant (NPP) capacity is at present 160 GW(e). If one adds the power stations under construction and ordered, a plant capacity of approximately 480 GW(e) is obtained for 1990, with the share of LWRs making up more than 80%. A modern LWR consumes in the open fuel cycle about 4400 metric tonnes of natural uranium per GW(e), assuming a lifetime of 30 years and a load factor of 70%. Considering the natural uranium reserves known at present and exploitable under economic conditions, it can be conveniently estimated that, with the present NPP capacity extension perspective, the natural uranium resources may be exhausted in a few decades. This trend can be counteracted in a flexible manner by various approaches in fuel cycle technology and strategy: (i) by steady further development of the established LWR technology the uranium consumption can be reduced by about 15%; (ii) closing the nuclear fuel cycle on the basis of LWRs (i.e. thermal uranium and plutonium recycling) implies up to 40% savings in natural uranium consumption; (iii) more recent considerations include the advanced pressurized water reactor (APWR). The APWR combines the proven PWR technology with a newly developed tight lattice core with greatly improved conversion characteristics (conversion ratio = 0.90 to 0.95). In terms of uranium utilization, the APWR has an efficiency three to five times higher than a PWR; (iv) Commercial introduction of FBR systems results in an optimal utilization of uranium which, at the same time, guarantees the supply of nuclear fuel well beyond the present century. For a corresponding transition period an energy supply system can be conceived which relies essentially on extended back-end fuel cycle capacities. These would facilitate a symbiosis of PWR, APWR and FBR, characterized by high flexibility with respect to long-term developments on the energy market. (author)

  10. Profileration-proof uranium/plutonium and thorium/uranium fuel cycles. Safeguards and non-profileration. 2. rev. ed.

    Energy Technology Data Exchange (ETDEWEB)

    Kessler, G.

    2017-07-01

    A brief outline of the historical development of the proliferation problem is followed by a description of the uranium-plutonium nuclear fuel cycle with uranium enrichment, fuel fabrication, the light-water reactors mainly in operation, and the breeder reactors still under development. The next item discussed is reprocessing of spent fuel with plutonium recycling and the future possibility to incinerate plutonium and the minor actinides: neptunium, americium, and curium. Much attention is devoted to the technical and scientific treatment of the IAEA surveillance concept of the uranium-plutonium fuel cycle. In this context, especially the physically possible accuracy of measuring U/Pu flow in the fuel cycle, and the criticism expressed of the accuracy in measuring the plutonium balance in large reprocessing plants of non-nuclear weapon states are analyzed. The second part of the book initially examines the assertion that reactor-grade plutonium could be used to build nuclear weapons whose explosive yield cannot be predicted accurately, but whose minimum explosive yield is still far above that of chemical explosive charges. Methods employed in reactor physics are used to show that such hypothetical nuclear explosive devices (HNEDs) would attain too high temperatures in the required implosion lenses as a result of the heat generated by the Pu-238 isotope always present in reactor plutonium of current light-water reactors. These lenses would either melt or tend to undergo chemical auto-explosion. Limits to the content of the Pu-238 isotope are determined above which such hypothetical nuclear weapons are not feasible on technical grounds. This situation is analyzed for various possibilities of the technical state of the art of making implosion lenses and various ways of cooling up to the use of liquid helium. The outcome is that, depending on the existing state of the art, reactor-grade plutonium from spent fuel elements of light-water reactors with a burnup of 35 to 58

  11. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

  12. Conversion of highly enriched uranium in thorium-232 based oxide fuel for light water reactors: MOX-T fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vapirev, E; Jordanov, T; Khristoskov, I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

    1996-12-31

    The possibility of using highly enriched uranium available from military inventories for production of mixed oxide fuel (MOX) has been proposed. The fuel is based on U-235 dioxide as fissile isotope and Th-232 dioxide as a non-fissile isotope. It is shown that although the fuel conversion coefficient to U-233 is expected to be less than 1, the proposed fuel has several important advantages resulting in cost reduction of the nuclear fuel cycle. The expected properties of MOX fuel (cross-sections, generated chains, delayed neutrons) are estimated. Due to fuel generation the initial enrichment is expected to be 1% less for production of the same energy. In contrast to traditional fuel no long living actinides are generated which reduces the disposal and reprocessing cost. 7 refs.

  13. The uranium and thorium separation in the chemical reprocessing of the irradiated fuel of thorium and uranium mixed oxides

    International Nuclear Information System (INIS)

    Oliveira, E.F. de.

    1984-09-01

    A bibliographic research has been carried out for reprocessing techniques of irradiated thorium fuel from nuclear reactors. The Thorex/Hoechst process has been specially considered to establish a method for reprocessing thorium-uranium fuel from PWR. After a series of cold tests performed in laboratory it was possible to set the behavior of several parameters affecting the Thorex/Hoechst process. Some comments and suggestions are presented for modifications in the process flosheet conditions. A discussion is carried out for operational conditions such as the aqueous to organic flow ratio the acidity of strip and scrub solutions in the process steps for thorium and uranium recovery. The operation diagrams have been constructed using equilibrium experimental data which correspond to conditions observed in laboratory. (Author) [pt

  14. Design of high density gamma-phase uranium alloys for LEU dispersion fuel applications

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.; Ray, Allison E.

    1998-01-01

    Uranium alloys are candidates for the fuel phase in aluminium matrix dispersion fuels requiring high uranium loading. Certain uranium alloys have been shown to have good irradiation performance at intermediate burnup. previous studies have shown that acceptable fission gas swelling behavior and fuel-aluminium interaction is possible only if the fuel alloy can be maintained in the high temperature body-centered-cubic γ-phase during fabrication and irradiation, at temperatures at which αU is the equilibrium phase. transition metals in Groups V through VIII are known to allow metastable retention of the gamma phase below the equilibrium isotherm. These metals have varying degrees of effectiveness in stabilizing the gamma phase. Certain alloys are metastable for very long times at the relatively low fuel temperatures seen in research operation. In this paper, the existing data on the gamma stability of binary and ternary uranium alloys is analysed. The mechanism and kinetics of decomposition of the gamma phase are assessed with the help of metal alloy theory. Alloys with the highest possible uranium content, good gamma-phase stability, and good neutronic performance are identified for further metallurgical studies and irradiation tests. Results from theory will be compared with experimentally generated data. (author)

  15. Determination of uranium in coated fuel particle compact by potassium fluoride fusion-gravimetric method

    International Nuclear Information System (INIS)

    Ito, Mitsuo; Iso, Shuichi; Hoshino, Akira; Suzuki, Shuichi.

    1992-03-01

    Potassium fluoride-gravimetric method has been developed for the determination of uranium in TRISO type-coated fuel particle compact. Graphite matrix in the fuel compact is burned off by heating it in a platinum crucible at 850degC. The coated fuel particles thus obtained are decomposed by fusion with potassium fluoride at 900degC. The melt was dissolved with sulfuric acid. Uranium is precipitated as ammonium diuranate, by passing ammonia gas through the solution. The resulting precipitate is heated in a muffle furnace at 850degC, to convert uranium into triuranium octoxide. Uranium in the triuranium octoxide was determined gravimetrically. Ten grams of caoted fuel particles were completely decomposed by fusion with 50 g of potassium fluoride at 900degC for 3 hrs. Analytical result for uranium in the fuel compact by the proposed method was 21.04 ± 0.05 g (n = 3), and was in good agreement with that obtained by non-destructive γ-ray measurement method : 21.01 ± 0.07 g (n = 3). (author)

  16. Wastes and waste management in the uranium fuel cycle for light water reactors

    International Nuclear Information System (INIS)

    Costello, J.M.

    1975-08-01

    The manufacturing processes in the uranium fuel cycle for light water reactors have been described with particular reference to the chemical and radiological wastes produced and the waste management procedures employed. The problems and possible solutions of ultimate disposal of high activity fission products and transuranium elements from reprocessing of irradiated fuel have been reviewed. Quantities of wastes arising in each stage of the fuel cycle have been summarised. Wastes arising from reactor operation have been described briefly. (author)

  17. Criticality safety of storage barrels for enriched uranium fresh fuel at the RB research reactor

    International Nuclear Information System (INIS)

    Pesic, M. P.

    1997-01-01

    Study on criticality safety of fresh low and high enriched uranium (LEU and HEU) fuel elements in the storage/transport barrels at the RB research reactor is carried out by using the well-known MCNP computer code. It is shown that studied arrays of tightly closed fuel barrels, each entirely loaded with 100 fresh (HEU or LEU) fuel slugs, are far away from criticality, even in cases of an unexpected flooding by light water.(author)

  18. Uranium and plutonium distribution in unirradiated mixed oxide fuel from industrial fabrication

    International Nuclear Information System (INIS)

    Hanus, D.; Kleykamp, H.

    1982-01-01

    Different process variants developed in the last few years by the firm ALKEM to manufacture FBR and LWR mixed oxide fuel are given. The uranium and plutonium distribution is determined on the pellets manufactured with the help of the electron beam microprobe. The stepwise improvement of the uranium-plutonium homogeneity in the short-term developed granulate variants and in the long-term developed new processes are illustrated starting with early standard processes for FBR fuel. An almost uniform uranium-plutonium distribution could be achieved for the long-term developed new processes (OKOM, AuPuC). The uranium-plutonium homogeneity are quantified in the pellets manufactured according to the considered process variants with a newly defined quality number. (orig.)

  19. Contribution to the study of nuclear fuel materials with a metallic uranium base

    International Nuclear Information System (INIS)

    Englander, M.

    1957-11-01

    In a power reactor destined to supply industrially recoverable thermal energy, the most economical source of heat still consists of natural metallic uranium. However, the nuclear fuel material, most often employed in the form of rods of 20 to 40 mm diameter, is subjected to a series of stresses which lead to irreversible distortions usually incompatible with the substructure of the reactor. As a result the fuel material must possess at the outset a certain number of qualities which must be determined. Investigations have therefore been carried out, first on the technological characters peculiar to each of the three allotropic phases of pure uranium metal, and on their interactions on the stabilisation of the material which consists of either cast uranium or uranium pile-treated in the γ phase. (author) [fr

  20. An overview of the regulation of uranium mining, milling, refining and fuel fabrication

    International Nuclear Information System (INIS)

    Smythe, W.D.

    1980-07-01

    The mining, milling, refining and fabrication of uranium into nuclear fuel are activities that have in common the handling of natural uranium. The occupational and environmental hazards resulting from these activities vary widely. Uranium presents a radiological hazard throughout, but the principal culprit is radium which creates an occupational hazard in the mine and mill and an environmental hazard in the waste products produced in both the mill and the refinery. The chemicals used in both these latter processes also present hazards. Fuel fabrication presents the least potential for occupational and environmental hazards. The Canadian Atomic Energy Control Board licenses eight plants, and one plant for the extraction of uranium from phosphoric acid. The licensing process is characterised by approval in stages, the placing of the burden of proof on the applicant, inspection at all stages, and joint review by all regulatory agencies involved

  1. Uranium loss from BISO-coated weak-acid-resin HTGR fuel

    International Nuclear Information System (INIS)

    Pearson, R.L.; Lindemer, T.B.

    1977-02-01

    Recycle fuel for the High-Temperature Gas-Cooled Reactor (HTGR) contains a weak-acid-resin (WAR) kernel, which consists of a mixture of UC 2 , UO 2 , and free carbon. At 1900 0 C, BISO-coated WAR UC 2 or UC 2 -UO 2 kernels lose a significant portion of their uranium in several hundred hours. The UC 2 decomposes and uranium diffuses through the pyrolytic coating. The rate of escape of the uranium is dependent on the temperature and the surface area of the UC 2 , but not on a temperature gradient. The apparent activation energy for uranium loss, ΔH, is approximately 90 kcal/mole. Calculations indicate that uranium loss from the kernel would be insignificant under conditions to be expected in an HTGR

  2. As nuclear fuel bank project moves ahead, support for facility cannot falter

    Energy Technology Data Exchange (ETDEWEB)

    Shepherd, John [nuclear 24, Redditch (United Kingdom)

    2016-10-15

    During the summer 2016, the historic next steps were taken to establish an international nuclear fuel bank under the auspices of the International Atomic Energy Agency (IAEA). The 'bank', officially known as the IAEA Low Enriched Uranium (LEU) Storage Facility is scheduled to be ready for operations by this time next year. The key role of the fuel bank will be to hold a reserve of LEU, the basic ingredient of nuclear fuel.

  3. Behavior of metallic uranium-fissium fuel in TREAT transient overpower tests

    International Nuclear Information System (INIS)

    Bauer, T.H.; Klickman, A.E.; Lo, R.K.; Rhodes, E.A.; Robinson, W.R.; Stanford, G.S.; Wright, A.E.

    1986-01-01

    TREAT tests M2, M3, and M4 were performed to obtain information on two key behavior characteristics of fuel under transient overpower accident conditions in metal-fueled fast reactors: the prefailure axial self-extrusion (elongation beyond thermal expansion) of fuel within intact cladding and the margin to cladding breach. Uranium-5 wt% fissium Experimental Breeder Reactor-II driver fuel pins were used for the tests since they were available as suitable stand-ins for the uranium-plutonium-zirconium ternary fuel, which is the reference fuel of the integral fast reactor (IFR) concept. The ternary fuel will be used in subsequent TREAT tests. Preliminary results from tests M2 and M3 were presented earlier. The present report includes significant advances in analysis as well as additional data from test M4. Test results and analysis have led to the development and validation of pin cladding failure and fuel extrusion models for metallic fuel, within reasonable uncertainties for the uranium-fissium alloy. Concepts involved are straightforward and readily extendable to ternary alloys and behavior in full-size reactors

  4. Transport of high enriched uranium fresh fuel from Yugoslavia to the Russian federation

    OpenAIRE

    Pešić Milan P.; Šotić Obrad; Hopwood William H.Jr

    2002-01-01

    This paper presents the relevant data related to the recent shipment (August 2002) of fresh highly enriched uranium fuel elements from Yugoslavia back to the Russian Federation for uranium down blending. In this way, Yugoslavia gave its contribution to the Reduced Enrichment for Research and Test Reactors (RERTR) Program and to the world's joint efforts to prevent possible terrorist actions against nuclear material potentially usable for the production of nuclear weapons.

  5. Transport of high enriched uranium fresh fuel from Yugoslavia to the Russian federation

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2002-01-01

    Full Text Available This paper presents the relevant data related to the recent shipment (August 2002 of fresh highly enriched uranium fuel elements from Yugoslavia back to the Russian Federation for uranium down blending. In this way, Yugoslavia gave its contribution to the Reduced Enrichment for Research and Test Reactors (RERTR Program and to the world's joint efforts to prevent possible terrorist actions against nuclear material potentially usable for the production of nuclear weapons.

  6. The measurements of critical mass with uranium fuel elements and thorium rods

    International Nuclear Information System (INIS)

    Yao Zhiquan; Chen Zhicheng; Yao Zewu; Ji Huaxiang; Bao Borong; Zhang Jiahua

    1991-01-01

    The critical experiments with uranium elements and Thorium rods have been performed in zero power reactor at Shanghai Institute of Nuclear Research. The critical masses have been measured in various U/Th ratios. The fuels are 3% 235 U-enriched uranium. The Thorium rods are made from power of ThF 4 . Ratios of calculated values to experimental values are nearly constant at 0.995

  7. System for uranium superficial density measurement in U3Si2 MTR fuel plates using radiography

    International Nuclear Information System (INIS)

    Hey, Martin A.; Gomez Marlasca, Fernando

    2003-01-01

    The paper describes a method for measuring uranium superficial density in high density uranium silicide (U 3 Si 2 ) MTR fuel plates, through the use of industrial radiography, a set of patterns built for this purpose, a transmission optical densitometer, and a quantitative model of analysis and measurement. Our choice for this particular method responds to its high accuracy, low cost and easy implementation according to the standing quality control systems. (author)

  8. Discharge Burnup Evaluation of Natural Uranium Loaded CANFLEX-43 Fuel Bundle

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Kim, Yong Hee; Kim, Won Young; Park, Joo Hwan

    2009-11-01

    Using WIMS-AECL code, which is 2-dimensional lattice core used in CANDU physics calculation, the discharge burnup of the natural uranium loaded CANFLEX-43 fuel bundle was evaluated by comparing the discharge burnup of standard 37 element fuel bundle. When the discharge burnup of the standard 37 element fuel is 7,200 MWd/MTU, that of the CANFLEX 43 fuel bundle was evaluated as 7,077 MWd/MTU, by applying the same lattice conditions for both fuel bundles

  9. Development of Advanced High Uranium Density Fuels for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, James [Univ. of Wisconsin, Madison, WI (United States); Butt, Darryl [Boise State Univ., ID (United States); Meyer, Mitchell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

    2016-02-15

    This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

  10. Argentina-LLNL-LANL Comparative Sample Analysis on UO2 fuel pellet CRM-125A for Nuclear Forensics

    Energy Technology Data Exchange (ETDEWEB)

    Kips, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-12-01

    The recent workshop on analytical plan development provided context and background for the next step in this engagement, i.e. a comparative sample analysis on CRM 125-A. This is a commercially available certified low-enriched uranium oxide fuel pellet material from New Brunswick National Laboratory (NBL) (see certificate in Annex 1).

  11. Do we soon run out of uranium? Long-term concepts of nuclear fuel supply

    International Nuclear Information System (INIS)

    Prasser, Horst-Michael

    2008-01-01

    The extension of the worldwide light water reactor fleet will cause the demand for uranium to grow. The static reach of identified resources might soon fall below the life time of new nuclear power plants which are usually designed for 60 years of operation, if the exploration of new uranium deposits will stop resulting in exploitable resources. The article discusses, if, as frequently claimed, the energy consumption in the uranium mines renders impossible to secure the nuclear fuel supply in the long term. (orig.)

  12. The uranium fuel cycle at IPEN - Energy and Nuclear Research Institute, SP, Brazil

    International Nuclear Information System (INIS)

    Abrao, Alcidio

    1994-09-01

    This paper summarizes the progress of research concerning the uranium fuel cycle set up at the IPEN, Sao Paulo, from the raw yellow-cake to the uranium hexafluoride. It covers the reconversion of the hexafluoride to ammonium uranyl tricarbonate and the manufacturing of the fuel elements for the swimming pool IEA-R1 reactor. This review extends the coverage of two pilot plants for uranium purification based upon ion exchange, one demonstration unity for the purification of uranyl nitrate by solvent extraction in pulsed columns, the unity of uranium tetrafluoride into moving bed reactors and a second one based upon the wet chemistry via uranium dioxide and aqueous hydrogen fluoride. The paper mentions the pilot plant for the preparation of uranium trioxide by the thermal decomposition of ammonium diuranate and a second unity by the thermal denitration of uranyl nitrate. The paper outlines the fluorine plant and the unity for the hexafluoride preparation, the unity for the conversion of the hexa to the ammonium uranyl tricarbonate and the fabrication of fuel elements for the IEA-R1 reactor. (author)

  13. Minimization of the fission product waste by using thorium based fuel instead of uranium dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, A. Abdelghafar, E-mail: Agalahom@yahoo.com

    2017-04-01

    This research discusses the neutronic characteristics of VVER-1200 assembly fueled with five different fuel types based on thorium. These types of fuel based on mixing thorium as a fertile material with different fissile materials. The neutronic characteristics of these fuels are investigated by comparing their neutronic characteristics with the conventional uranium dioxide fuel using the MCNPX code. The objective of this study is to reduce the production of long-lived actinides, get rid of plutonium component and to improve the fuel cycle economy while maintaining acceptable values of the neutronic safety parameters such as moderator temperature coefficient, Doppler coefficient and effective delayed neutrons (β). The thorium based fuel has a more negative Doppler coefficient than uranium dioxide fuel. The moderator temperature coefficient (MTC) has been calculated for the different proposed fuels. Also, the fissile inventory ratio has been calculated at different burnup step. The use of Th-232 as a fertile material instead of U-238 in a nuclear fuel is the most promising fuel in VVER-1200 as it is the ideal solution to avoid the production of more plutonium components and long-lived minor actinides. The reactor grade plutonium accumulated in light water reactor with burnup can be recycled by mixing it with Th-232 to fuel the VVER-1200 assembly. The concentrations of Xe-135 and Sm-151 have been investigated, due to their high thermal neutron absorption cross section.

  14. Development of metal uranium fuel and testing of construction materials (I-VI); Part I; Razvoj metalnog goriva i ispitivanje konstrukcionih materijala (I-VI deo); I deo

    Energy Technology Data Exchange (ETDEWEB)

    Mihajlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors.

  15. Uranium-thorium fuel cycle in a very high temperature hybrid system

    International Nuclear Information System (INIS)

    Hernandez, C.R.G.; Oliva, A.M.; Fajardo, L.G.; Garcia, J.A.R.; Curbelo, J.P.; Abadanes, A.

    2011-01-01

    Thorium is a potentially valuable energy source since it is about three to four times as abundant as Uranium. It is also a widely distributed natural resource readily accessible in many countries. Therefore, Thorium fuels can complement Uranium fuels and ensure long term sustainability of nuclear power. The main advantages of the use of a hybrid system formed by a Pebble Bed critical nuclear reactor and two Pebble Bed Accelerator Driven Systems (ADSs) using a Uranium-Thorium (U + Th) fuel cycle are shown in this paper. Once-through and two step U + Th fuel cycle was evaluated. With this goal, a preliminary conceptual design of a hybrid system formed by a Graphite Moderated Gas-Cooled Very High Temperature Reactor and two ADSs is proposed. The main parameters related to the neutronic behavior of the system in a deep burn scheme are optimized. The parameters that describe the nuclear fuel breeding and Minor Actinide stockpile are compared with those of a simple Uranium fuel cycle. (author)

  16. Occupational safety data and casualty rates for the uranium fuel cycle

    International Nuclear Information System (INIS)

    O'Donnell, F.R.; Hoy, H.C.

    1981-10-01

    Occupational casualty (injuries, illnesses, fatalities, and lost workdays) and production data are presented and used to calculate occupational casualty incidence rates for technologies that make up the uranium fuel cycle, including: mining, milling, conversion, and enrichment of uranium; fabrication of reactor fuel; transportation of uranium and fuel elements; generation of electric power; and transmission of electric power. Each technology is treated in a separate chapter. All data sources are referenced. All steps used to calculate normalized occupational casualty incidence rates from the data are presented. Rates given include fatalities, serious cases, and lost workdays per 100 man-years worked, per 10 12 Btu of energy output, and per other appropriate units of output

  17. Development of the uranium recovery process from rejected fuel plates in the fabrication of MTR type nuclear fuel

    International Nuclear Information System (INIS)

    Fleming Rubio, Peter Alex

    2010-01-01

    The current work was made in Conversion laboratory belonging to Chilean Nuclear Energy Commission, CCHEN. This is constituted by the development of three hydrometallurgical processes, belonging to the recovery of uranium from fuel plates based on uranium silicide (U_3Si_2) process, for nuclear research reactors MTR (Material Testing Reactor) type, those that come from the Fuel Elements Manufacture Plant, PEC. In the manufacturing process some of these plates are subjected to destructive tests by quality requirement or others are rejected for non-compliance with technical specifications, such as: lack of homogenization of the dispersion of uraniferous compound in the meat, as well as the appearance of the defects, such as blisters, so-called "dog bone", "fish tail", "remote islands", among others. Because the uranium used is enriched in 19.75% U_2_3_5 isotope, which explains the high value in the market, it must be recovered for reuse, returning to the production line of fuel elements. The uranium silicide, contained in the plates, is dispersed in an aluminum matrix and covered with plates and frames of ASTM 6061 Aluminum, as a sandwich coating, commonly referred to as 'meat' (sandwich meat). As aluminum is the main impurity, the process begins with this metal dissolution, present in meat and plates, by NaOH reaction, followed by a vacuum filtration, washing and drying, obtaining a powder of uranium silicide, with a small impurities percentage. Then, the crude uranium silicide reacts with a solution of hydrofluoric acid, dissolving the silicon and simultaneously precipitating UF_4 by reaction with HNO_3, obtaining an impure UO_2(NO_3)_2 solution. The experimental work was developed and implemented at laboratory scale for the three stages pertaining to the uranium recovery process, determining for each one the optimum operation conditions: temperature, molarity or concentration, reagent excess, among others (author)

  18. Uranium

    International Nuclear Information System (INIS)

    Cuney, M.; Pagel, M.; Leroy, J.

    1992-01-01

    First, this book presents the physico-chemical properties of Uranium and the consequences which can be deduced from the study of numerous geological process. The authors describe natural distribution of Uranium at different scales and on different supports, and main Uranium minerals. A great place in the book is assigned to description and classification of uranium deposits. The book gives also notions on prospection and exploitation of uranium deposits. Historical aspects of Uranium economical development (Uranium resources, production, supply and demand, operating costs) are given in the last chapter. 7 refs., 17 figs

  19. Babcock and Wilcox plate fabrication experience with uranium silicide spherical fuel

    International Nuclear Information System (INIS)

    Todd, Lawrence E.; Pace, Brett W.

    1996-01-01

    This report is written to present the fuel fabrication experience of Babcock and Wilcox using atomized spherical uranium silicide powder. The intent is to demonstrate the ability to fabricate fuel plates using spherical powder and to provide useful information proceeding into the next phase of work using this type of fuel. The limited quantity of resources- spherical powder and time, did not allow for much process optimizing in this work scope. However, the information contained within provides optimism for the future of spherical uranium silicide fuel plate fabrication at Babcock and Wilcox.The success of assembling fuel elements with spherical powder will enable Babcock and Wilcox to reduce overall costs to its customers while still maintaining our reputation for providing high quality research and test reactor products. (author)

  20. Composition and Distribution of Tramp Uranium Contamination on BWR and PWR Fuel Rods

    International Nuclear Information System (INIS)

    Schienbein, Marcel; Zeh, Peter; Hurtado, Antonio; Rosskamp, Matthias; Mailand, Irene; Bolz, Michael

    2012-09-01

    In a joint research project of VGB and AREVA NP GmbH the behaviour of alpha nuclides in nuclear power plants with light water reactors has been investigated. Understanding the source and the behaviour of alpha nuclides is of big importance for planning radiation protection measures for outages and upcoming dismantling projects. Previous publications have shown the correlation between plant specific alpha contamination of the core and the so called 'tramp fuel' or 'tramp uranium' level which is linked to the defect history of fuel assemblies and accordingly the amount of previously washed out fuel from defective fuel rods. The methodology of tramp fuel estimation is based on fission product concentrations in reactor coolant but also needs a good knowledge of tramp fuel composition and in-core distribution on the outer surface of fuel rods itself. Sampling campaigns of CRUD deposits of irradiated fuel assemblies in different NPPs were performed. CRUD analyses including nuclide specific alpha analysis have shown systematic differences between BWR and PWR plants. Those data combined with literature results of fuel pellet investigations led to model improvements showing that a main part of fission products is caused by fission of Pu-239 an activation product of U-238. CRUD investigations also gave a better picture of the in-core composition and distribution of the tramp uranium contamination. It was shown that the tramp uranium distribution in PWR plants is time dependent. Even new fuel assemblies will be notably contaminated after only one cycle of operation. For PWR applies the following logic: the higher the local power the higher the contamination. With increasing burnup the local rod power usually decreases leading to decreasing tramp uranium contamination on the fuel rod surface. This is not applicable for tramp uranium contamination in BWR. CRUD contamination (including the tramp fuel deposits) is much more fixed and is constantly increasing