WorldWideScience

Sample records for enrichment uranium fuels

  1. 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)

  2. Uranium enrichment

    International Nuclear Information System (INIS)

    1990-01-01

    This report looks at the following issues: How much Soviet uranium ore and enriched uranium are imported into the United States and what is the extent to which utilities flag swap to disguise these purchases? What are the U.S.S.R.'s enriched uranium trading practices? To what extent are utilities required to return used fuel to the Soviet Union as part of the enriched uranium sales agreement? Why have U.S. utilities ended their contracts to buy enrichment services from DOE?

  3. 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

  4. 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

  5. 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)

  6. 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)

  7. Uranium enrichment

    International Nuclear Information System (INIS)

    Rae, H.K.; Melvin, J.G.

    1988-06-01

    Canada is the world's largest producer and exporter of uranium, most of which is enriched elsewhere for use as fuel in LWRs. The feasibility of a Canadian uranium-enrichment enterprise is therefore a perennial question. Recent developments in uranium-enrichment technology, and their likely impacts on separative work supply and demand, suggest an opportunity window for Canadian entry into this international market. The Canadian opportunity results from three particular impacts of the new technologies: 1) the bulk of the world's uranium-enrichment capacity is in gaseous diffusion plants which, because of their large requirements for electricity (more than 2000 kW·h per SWU), are vulnerable to competition from the new processes; 2) the decline in enrichment costs increases the economic incentive for the use of slightly-enriched uranium (SEU) fuel in CANDU reactors, thus creating a potential Canadian market; and 3) the new processes allow economic operation on a much smaller scale, which drastically reduces the investment required for market entry and is comparable with the potential Canadian SEU requirement. The opportunity is not open-ended. By the end of the century the enrichment supply industry will have adapted to the new processes and long-term customer/supplier relationships will have been established. In order to seize the opportunity, Canada must become a credible supplier during this century

  8. 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.

  9. 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

  10. 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

  11. 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)

  12. Uranium enrichment

    International Nuclear Information System (INIS)

    Mohrhauer, H.

    1982-01-01

    The separation of uranium isotopes in order to enrich the fuel for light water reactors with the light isotope U-235 is an important part of the nuclear fuel cycle. After the basic principals of isotope separation the gaseous diffusion and the centrifuge process are explained. Both these techniques are employed on an industrial scale. In addition a short review is given on other enrichment techniques which have been demonstrated at least on a laboratory scale. After some remarks on the present situation on the enrichment market the progress in the development and the industrial exploitation of the gas centrifuge process by the trinational Urenco-Centec organisation is presented. (orig.)

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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)

  19. Uranium enrichment

    International Nuclear Information System (INIS)

    1989-01-01

    GAO was asked to address several questions concerning a number of proposed uranium enrichment bills introduced during the 100th Congress. The bill would have restructured the Department of Energy's uranium enrichment program as a government corporation to allow it to compete more effectively in the domestic and international markets. Some of GAO's findings discussed are: uranium market experts believe and existing market models show that the proposed DOE purchase of a $750 million of uranium from domestic producers may not significantly increase production because of large producer-held inventories; excess uranium enrichment production capacity exists throughout the world; therefore, foreign producers are expected to compete heavily in the United States throughout the 1990s as utilities' contracts with DOE expire; and according to a 1988 agreement between DOE's Offices of Nuclear Energy and Defense Programs, enrichment decommissioning costs, estimated to total $3.6 billion for planning purposes, will be shared by the commercial enrichment program and the government

  20. 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...

  1. 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.

  2. 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

  3. 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)

  4. 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)

  5. 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

  6. 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)

  7. 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)

  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. Uranium Enrichment, an overview

    International Nuclear Information System (INIS)

    Coates, J.H.

    1994-01-01

    This general presentation on uranium enrichment will be followed by lectures on more specific topics including descriptions of enrichment processes and assessments of the prevailing commercial and industrial situations. I shall therefore avoid as much as possible duplications with these other lectures, and rather dwell on: some theoretical aspects of enrichment in general, underlying the differences between statistical and selective processes, a review and comparison between enrichment processes, remarks of general order regarding applications, the proliferation potential of enrichment. It is noteworthy that enrichment: may occur twice in the LWR fuel cycle: first by enriching natural uranium, second by reenriching uranium recovered from reprocessing, must meet LWR requirements, and in particular higher assays required by high burn up fuel elements, bears on the structure of the entire front part of the fuel cycle, namely in the conversion/reconversion steps only involving UF 6 for the moment. (author). tabs., figs., 4 refs

  10. 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

  11. 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.

  12. 78 FR 63518 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Science.gov (United States)

    2013-10-24

    ... 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... Louisiana Energy Services (LES), LLC, National Enrichment Facility in Eunice, New Mexico, and has authorized...

  13. 77 FR 18272 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-03-27

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC, National Enrichment Facility, Eunice... Louisiana Energy Services (LES), LLC, National enrichment Facility in Eunice, New Mexico, and has verified...

  14. 77 FR 65729 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-10-30

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC, National Enrichment Facility, Eunice... Services (LES), LLC, National Enrichment Facility in Eunice, New Mexico, and has verified that cascades...

  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. 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

  17. Uranium enrichment

    International Nuclear Information System (INIS)

    1991-08-01

    This paper reports that in 1990 the Department of Energy began a two-year project to illustrate the technical and economic feasibility of a new uranium enrichment technology-the atomic vapor laser isotope separation (AVLIS) process. GAO believes that completing the AVLIS demonstration project will provide valuable information about the technical viability and cost of building an AVLIS plant and will keep future plant construction options open. However, Congress should be aware that DOE still needs to adequately demonstrate AVLIS with full-scale equipment and develop convincing cost projects. Program activities, such as the plant-licensing process, that must be completed before a plant is built, could take many years. Further, an updated and expanded uranium enrichment analysis will be needed before any decision is made about building an AVLIS plant. GAO, which has long supported legislation that would restructure DOE's uranium enrichment program as a government corporation, encourages DOE's goal of transferring AVLIS to the corporation. This could reduce the government's financial risk and help ensure that the decision to build an AVLIS plant is based on commercial concerns. DOE, however, has no alternative plans should the government corporation not be formed. Further, by curtailing a planned public access program, which would have given private firms an opportunity to learn about the technology during the demonstration project, DOE may limit its ability to transfer AVLIS to the private sector

  18. 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.

  19. 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.

  20. Uranium enrichment

    International Nuclear Information System (INIS)

    1991-11-01

    This paper analyzes under four different scenarios the adequacy of a $500 million annual deposit into a fund to pay for the cost of cleaning up the Department of Energy's (DOE) three aging uranium enrichment plants. These plants are located in Oak Ridge, Tennessee; Paducah, Kentucky; and Portsmouth, Ohio. In summary the following was found: A fixed annual $500 million deposit made into a cleanup fund would not be adequate to cover total expected cleanup costs, nor would it be adequate to cover expected decontamination and decommissioning (D and D) costs. A $500 million annual deposit indexed to an inflation rate would likely be adequate to pay for all expected cleanup costs, including D and D costs, remedial action, and depleted uranium costs

  1. Promotion of uranium enrichment business

    International Nuclear Information System (INIS)

    Kurushima, Morihiro

    1981-01-01

    The Committee on Nuclear Power has studied on the basic nuclear power policy, establishing its five subcommittees, entrusted by the Ministry of Nternational Trade and Industry. The results of examination by the subcommittee on uranium enrichment business are given along with a report in this connection by the Committee. In order to establish the nuclear fuel cycle, the aspect of uranium enrichment is essential. The uranium enrichment by centrifugal process has proceeded steadily in Power Reactor and Nuclear Fuel Development Corporation. The following matters are described: the need for domestic uranium enrichment, the outlook for overseas enrichment services and the schedule for establishing domestic enrichment business, the current state of technology development, the position of the prototype enrichment plant, the course to be taken to establish enrichment business the main organization operating the prototype and commercial plants, the system of supplying centrifuges, the domestic conversion of natural uranium the subsidies for uranium enrichment business. (J.P.N.)

  2. Uranium enrichment. Enrichment processes

    International Nuclear Information System (INIS)

    Alexandre, M.; Quaegebeur, J.P.

    2009-01-01

    Despite the remarkable progresses made in the diversity and the efficiency of the different uranium enrichment processes, only two industrial processes remain today which satisfy all of enriched uranium needs: the gaseous diffusion and the centrifugation. This article describes both processes and some others still at the demonstration or at the laboratory stage of development: 1 - general considerations; 2 - gaseous diffusion: physical principles, implementation, utilisation in the world; 3 - centrifugation: principles, elementary separation factor, flows inside a centrifuge, modeling of separation efficiencies, mechanical design, types of industrial centrifuges, realisation of cascades, main characteristics of the centrifugation process; 4 - aerodynamic processes: vortex process, nozzle process; 5 - chemical exchange separation processes: Japanese ASAHI process, French CHEMEX process; 6 - laser-based processes: SILVA process, SILMO process; 7 - electromagnetic and ionic processes: mass spectrometer and calutron, ion cyclotron resonance, rotating plasmas; 8 - thermal diffusion; 9 - conclusion. (J.S.)

  3. 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)

  4. 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

  5. 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.

  6. 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)

  7. 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.

  8. 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)

  9. 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)

  10. 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)

  11. 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.)

  12. Volatile behaviour of enrichment uranium in the total nuclear fuel price

    International Nuclear Information System (INIS)

    Arnaiz, J.; Inchausti, J. M.; Tarin, F.

    2004-01-01

    In this article the historical high volatile behaviour of the total nuclear fuel price is evaluated quantitatively and it is concluded that it has been due mainly to the fluctuations of the price of the principal components of enriched uranium (concentrates and enrichment). In order to avoid the negative effects of this volatiles behaviour as far as possible, a basic strategy in the uranium procurement activities is recommended (union of buyers, diversification of supplier, stock management, optimisation of contract portfolio and suitable currency management that guarantees a reliable uranium supply at reasonable prices. These guidelines are those that ENUSA has been following on behalf of the Spanish Utilities in the Commission of Uranium Procurement (CAU in Spanish). (Author) 11 refs

  13. Derived enriched uranium market

    International Nuclear Information System (INIS)

    Rutkowski, E.

    1996-01-01

    The potential impact on the uranium market of highly enriched uranium from nuclear weapons dismantling in the Russian Federation and the USA is analyzed. Uranium supply, conversion, and enrichment factors are outlined for each country; inventories are also listed. The enrichment component and conversion components are expected to cause little disruption to uranium markets. The uranium component of Russian derived enriched uranium hexafluoride is unresolved; US legislation places constraints on its introduction into the US market

  14. Uranium Conversion & Enrichment

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-06

    The isotopes of uranium that are found in nature, and hence in ‘fresh’ Yellowcake’, are not in relative proportions that are suitable for power or weapons applications. The goal of conversion then is to transform the U3O8 yellowcake into UF6. Conversion and enrichment of uranium is usually required to obtain material with enough 235U to be usable as fuel in a reactor or weapon. The cost, size, and complexity of practical conversion and enrichment facilities aid in nonproliferation by design.

  15. Determination of enrichment of recycle uranium fuels for different burnup values

    International Nuclear Information System (INIS)

    Zabunoglu, Okan H.

    2008-01-01

    Uranium (U) recovered from spent LWR fuels by reprocessing, which contains small amounts of U-236, is to be enriched before being re-irradiated as the recycle U. During the enrichment of recovered U in U-235, the mass fraction of U-236 also increases. Since the existence of U-236 in the recycle U has a negative effect on neutron economy, a greater enrichment of U-235 in the recycle U is required for reaching the same burnup as can be reached by the fresh U fuel. Two burnup values play the most important role in determining the enrichment of recycle U: (1) discharge burnup of spent fuel from which the recycle U is obtained and (2) desired discharge burnup of the recycle U fuel. A step-by-step procedure for calculating the enrichment of the recycle U as a function of these two burnup values is introduced. The computer codes MONTEBURNS and ORIGEN-S are made use of and a three-component (U-235, U-236, U-238) enrichment scheme is applied for calculating the amount of U-236 in producing the recycle U from the recovered U. As was aimed, the resulting expression is simple enough for quick/hand calculations of the enrichment of the recycle U for any given discharge burnup of spent fuel and for any desired discharge burnup of the recycle U fuel, most accurately within the range of 33,000-50,000 MWd/tonU

  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. 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)

  18. 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)

  19. Uranium enrichment plans

    International Nuclear Information System (INIS)

    Thomas, D.C.; Gagne, R.W.

    1978-01-01

    The following topics are covered: the status of the Government's existing uranium enrichment services contracts, natural uranium requirements based on the latest contract information, uncertainty in predicting natural uranium requirements based on uranium enrichment contracts, and domestic and foreign demand assumed in enrichment planning

  20. Status of the natural and enriched uranium market: the basic economical factor for the development of the fuel cycle

    International Nuclear Information System (INIS)

    Nochev, T.

    1999-01-01

    Status of the Natural and Enriched Uranium Market - the Basic. Economical Factor for the Development of the Fuel Cycle An overview of the status of the natural and enriched uranium market has been performed and it offers a possibility to estimate the changes and tendencies, the knowledge of which is needed in negotiations about the fresh fuel. The simplified financial analysis presented here demonstrates the economical profitability of the storage of the spent fuel making now the allocations for the future reprocessing

  1. Blueprint for domestic uranium enrichment

    International Nuclear Information System (INIS)

    1981-01-01

    The AEC advisory committee on domestic production of uranium enrichment has studied for more than a year how to achieve the domestic enrichment of uranium by the construction and operation of a commercial enriching plant using centrifugal separation method, and the report was submitted to the Atomic Energy Commission on August 18, 1980. Japan has depended wholly on overseas services for her uranium enrichment needs, but the development of domestic enrichment has been carried on in parallel. The AEC decided to construct a uranium enrichment pilot plant using centrifuges, and it has been forwarded as a national project. The plant is operated by the Power Reactor and Nuclear Fuel Development Corp. since 1979. The capacity of the plant will be raised to approximately 75 ton SWU a year. The centrifuges already operated have provided the first delivery of fuel of about 1 ton for the ATR ''Fugen''. The demand-supply balance of uranium enrichment service, the significance of the domestic enrichment of uranium, the evaluation of uranium enrichment technology, the target for domestic enrichment plan, the measures to promote domestic uranium enrichment, and the promotion of the construction of a demonstration plant are reported. (Kako, I.)

  2. 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

  3. 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.

  4. Beta activity of enriched uranium

    International Nuclear Information System (INIS)

    Nambiar, P.P.V.J.; Ramachandran, V.

    1975-01-01

    Use of enriched uranium as reactor fuel necessitates its handling in various forms. For purposes of planning and organising radiation protection measures in enriched uranium handling facilities, it is necessary to have a basic knowledge of the radiation status of enriched uranium systems. The theoretical variations in beta activity and energy with U 235 enrichment are presented. Depletion is considered separately. Beta activity build up is also studied for two specific enrichments, in respect of which experimental values for specific alpha activity are available. (author)

  5. 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)

  6. recovery of enriched uranium from waste solution obtained from fuel fabrication laboratories

    International Nuclear Information System (INIS)

    Othman, S.H.A.

    2003-01-01

    reversed-phase partition chromatography is shown to be a convenient and applicable method for the quantitative recovery of uranium (19.7% enriched with 235 U) from highly impure solution . the processing of uranium compounds for atomic energy project especially in FMPP(Egyptian fuel manufacture pilot plant) gives rise to a variety of wastes in which the uranium content is of considerable importance. the recovery of uranium from concentrated mother liquors produced from ADU (ammonium diuranate ) precipitation, as well as those due to ADU washing is studied in this work. column of poly-trifluoro-monochloro-ethilene (Kel-F) supporting tri-n-butyl-phosphate (TBP) retains uranium .impurities are eluted with 6.5 M HCl, and the uranium is eluted with water and the recovery of uranium is better than 94%. A mathematical model was suggested to stimulate the sorption process of uranium ions (or any other ion ) by column of solvent impregnated resin containing organic extractant (the same as the previous column) . An excellent agreement was founded between the experimental results and the mathematical model

  7. Development of ISA procedure for uranium fuel fabrication and enrichment facilities

    International Nuclear Information System (INIS)

    Yamate, Kazuki; Arakawa, Tomoyuki; Yamashita, Masahiro; Sasaki, Noriaki; Hirano, Mitsumasa

    2011-01-01

    The integrated safety analysis (ISA) procedure has been developed to apply risk-informed regulation to uranium fuel fabrication and enrichment facilities. The major development efforts are as follows: (a) preparing the risk level matrix as an index for items-relied-on-for-safety (IROFS) identification, (b) defining requirements of IROFS, and (c) determining methods of IROFS importance based on the results of risk- and scenario-based analyses. For the risk level matrix, the consequence and likelihood categories have been defined by taking into account the Japanese regulatory laws, rules, and safety standards. The trial analyses using the developed procedure have been performed for several representative processes of the reference uranium fuel fabrication and enrichment facilities. This paper presents the results of the ISA for the sintering process of the reference fabrication facility. The results of the trial analyses have demonstrated the applicability of the procedure to the risk-informed regulation of these facilities. (author)

  8. Optimal management of fuel in nuclear reactors with slightly enriched uranium and heavy water

    International Nuclear Information System (INIS)

    Serghiuta, D.

    1994-01-01

    This Ph.D. thesis presents the general principles guiding the optimal management of the fuel in CANDU type reactors with slightly enriched uranium. A method is devised which is based on the specific physical characteristics of this type of reactors and makes use of the multipurpose mathematical programming satisfying economical and nuclear safety requirements. The main goal of this work was the establishing of a refueling optimal methodology at equilibrium maintaining the reactor critical during operation. It also minimizes the fuel cycle cost through minimization of the utilized fissile material and at the same time by maximizing the reactor duty time through an optimal chain of refilling operations. This work can be considered as a contribution to a future project of CANDU type reactor core based on slightly enriched uranium. 74 Figs., 9 Tabs., 62 Refs

  9. 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)

  10. 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.

  11. 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

  12. 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.

  13. 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.

  14. Uranium enrichment plans

    International Nuclear Information System (INIS)

    Gagne, R.W.; Thomas, D.C.

    1977-01-01

    The status of existing uranium enrichment contracts in the US is reviewed and expected natural uranium requirements for existing domestic uranium enrichment contracts are evaluated. Uncertainty in natural uranium requirements associated with requirements-type and fixed-commitment type contracts is discussed along with implementation of variable tails assay

  15. Recovery of enriched Uranium (20% U-235) from wastes obtained in the preparation of fuel elements for argonaut type reactors

    International Nuclear Information System (INIS)

    Uriarte, A.; Ramos, L.; Estrada, J.; del Val, J. L.

    1962-01-01

    Results obtained with the two following installations for recovering enriched uranium (20% U-235) from wastes obtained in the preparation of fuel elements for Argonaut type reactors are presented. Ion exchange unit to recover uranium form mother liquors resulting from the precipitation ammonium diuranate (ADU) from UO 2 F 2 solutions. Uranium recovery unit from solid wastes from the process of manufacture of fuel elements, consisting of a) waste dissolution, and b) extraction with 10% (v/v) TBP. (Author) 9 refs

  16. How can Korea secure uranium enrichment and spent fuel reprocessing rights?

    International Nuclear Information System (INIS)

    Roh, Seungkook; Kim, Wonjoon

    2014-01-01

    South Korea is heavily dependent on energy resources from other countries and nuclear energy accounts for 31% of Korea's electric power generation as a major energy. However, Korea has many limitations in uranium enrichment and spent fuel reprocessing under the current Korea-U.S. nuclear agreement, although they are economically and politically important to Korea due to a significant problems in nuclear fuel storages. Therefore, in this paper, we first examine those example countries – Japan, Vietnam, and Iran – that have made nuclear agreements with the U.S. or have changed their agreements to allow the enrichment of uranium and the reprocessing of spent fuel. Then, we analyze those countries' nuclear energy policies and review their strategic repositioning in the relationship with the U.S. We find that a strong political stance for peaceful usage of nuclear energy including the legislation of nuclear laws as was the case of Japan. In addition, it is important for Korea to acquire advanced technological capability such as sodium-cooled fast reactor (SFR) because SFR technologies require plutonium to be used as fuel rather than uranium-235. In addition, Korea needs to leverage its position in nuclear agreement between China and the U.S. as was the case of Vietnam

  17. 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)

  18. 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

  19. 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 I; Jordanov, T; Christoskov, I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

    1994-12-31

    The idea of conversion of highly enriched uranium (HEU) from warheads without mixing it with natural uranium as well as the utilization of plutonium as fuel component is discussed. A nuclear fuel which is a mixture of 4% {sup 235}U (HEU) as a fissile isotope and 96 % {sup 232}Th (ThO{sub 2}) as a non-fissile isotope in a mixed oxide with thorium fuel is proposed. It is assumed that plutonium can also be used in the proposed fuel in a mixture with {sup 235}U. The following advantages of the use of HEU in LWRs in mixed {sup 235}U - Th fuel are pointed out: (1) No generation of long-living plutonium and americium isotopes (in case of reprocessing the high level radioactive wastes will contain only fission fragments and uranium); (2) The high conversion ratio of Th extends the expected burnup by approximately 1/3 without higher initial enrichment (the same initial enrichment simplifies the problem for compensation of the excess reactivity in the beginning with burnable poison and boric acid); (3) The high conversion ratio of Th allows the fuel utilization with less initial enrichment (by approx. 1/3) for the same burnup; thus less excess reactivity has to be compensated after reloading; in case of fuel reprocessing all fissile materials ({sup 235}U + {sup 233}U) could be chemically extracted. Irrespectively to the optimistic expectations outlined, further work including data on optimal loading and reloading schemes, theoretical calculations of thermal properties of {sup 235}U + Th fuel rods, manufacturing of several test fuel assemblies and investigations of their operational behaviour in a reactor core is still needed. 1 fig., 7 refs.

  20. 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

  1. Repository emplacement costs for Al-clad high enriched uranium spent fuel

    International Nuclear Information System (INIS)

    McDonell, W.R.; Parks, P.B.

    1994-01-01

    A range of strategies for treatment and packaging of Al-clad high-enriched uranium (HEU) spent fuels to prevent or delay the onset of criticality in a geologic repository was evaluated in terms of the number of canisters produced and associated repository costs incurred. The results indicated that strategies in which neutron poisons were added to consolidated forms of the U-Al alloy fuel generally produced the lowest number of canisters and associated repository costs. Chemical processing whereby the HEU was removed from the waste form was also a low cost option. The repository costs generally increased for isotopic dilution strategies, because of the substantial depleted uranium added. Chemical dissolution strategies without HEU removal were also penalized because of the inert constituents in the final waste glass form. Avoiding repository criticality by limiting the fissile mass content of each canister incurred the highest repository costs

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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)

  7. 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

  8. 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

  9. RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA

    Energy Technology Data Exchange (ETDEWEB)

    Kelly Cummins; Igor Bolshinsky; Ken Allen; Tihomir Apostolov; Ivaylo Dimitrov

    2009-07-01

    In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required to complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.

  10. Russian-Origin Highly Enriched Uranium Spent Nuclear Fuel Shipment From Bulgaria

    International Nuclear Information System (INIS)

    Cummins, Kelly; Bolshinsky, Igor; Allen, Ken; Apostolov, Tihomir; Dimitrov, Ivaylo

    2009-01-01

    In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required to complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.

  11. Return of 80% highly enriched uranium fresh fuel from Yugoslavia to Russia

    International Nuclear Information System (INIS)

    Pesic, M.; Sotic, O.; Subotic, K.; Hopwood, W. Jr; Moses, S.; Wander, T.; Smirnov, A.; Kanashov, B.; Eshcherkin, A.; Efarov, S.; Olivieri, C.; Loghin, N. E.

    2003-01-01

    The transport of almost 50 kg of highly enriched (80%) uranium (HEU), in the form of fresh TVR-S fuel elements, from the Vin a Institute of Nuclear Sciences, Yugoslavia, to the Russian Federation for uranium reprocessing was carried out in August 2002. This act was a contribution of the Government of the Federal Republics of Yugoslavia (now Serbia and Montenegro) to the world's joint efforts to prevent possible actions of terrorists against nuclear material that potentially would be usable for the production of nuclear weapons. Basic aspects of this complex operation, carried out mainly by transport teams of the Vinca Institute and of the Institute for Safe Transport of Nuclear Materials from Dimitrovgrad, Russian Federation, are described in this paper. A team of IAEA safety inspectors and experts from the DOE, USA, for transport and non-proliferation, supported the whole operation. (author)

  12. Return of 80% highly enriched uranium fresh fuel from Yugoslavia to Russia

    International Nuclear Information System (INIS)

    Pesic, M.; Sotic, O.; Subotic, K.; Hopwood, W. Jr; Moses, S.; Wander, T.; Smirnov, A.; Kanashov, B.; Eshcherkin, A.; Efarov, S.; Olivieri, C.; Loghin, N. E.

    2003-01-01

    The transport of almost 50 kg of highly enriched (80%) uranium (HEU), in the form of fresh TVR-S fuel elements, from the Vinca Institute of Nuclear Sciences, Yugoslavia, to the Russian Federation for uranium reprocessing was carried out in August 2002. This act was a contribution of the Government of the Federal Republics of Yugoslavia (now Serbia and Montenegro) to the world's joint efforts to prevent possible actions of terrorists against nuclear material that potentially would be usable for the production of nuclear weapons. Basic aspects of this complex operation, carried out mainly by transport teams of the Vinca Institute and of the Institute for Safe Transport of Nuclear Materials from Dimitrovgrad, Russian Federation, are described in this paper. A team of IAEA safety inspectors and experts from the DOE, USA, for transport and non-proliferation, supported the whole operation. (author)

  13. 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)

  14. 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

  15. 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

  16. Recovery of enriched Uranium (20% U-235) from wastes obtained in the preparation of fuel elements for argonaut type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Uriarte, A; Ramos, L; Estrada, J; Val, J L. del

    1962-07-01

    Results obtained with the two following installations for recovering enriched uranium (20% U-235) from wastes obtained in the preparation of fuel elements for Argonaut type reactors are presented. Ion exchange unit to recover uranium form mother liquors resulting from the precipitation ammonium diuranate (ADU) from UO{sub 2}F{sub 2} solutions. Uranium recovery unit from solid wastes from the process of manufacture of fuel elements, consisting of a) waste dissolution, and b) extraction with 10% (v/v) TBP. (Author) 9 refs.

  17. 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.

  18. 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.

  19. 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.

  20. Uranium enrichment techniques

    International Nuclear Information System (INIS)

    Hamdoun, N.A.

    2007-01-01

    This article includes an introduction about the isotopes of natural uranium, their existence and the difficulty of the separation between them. Then it goes to the details of a number of methods used to enrich uranium: Gaseous Diffusion method, Electromagnetic method, Jet method, Centrifugal method, Chemical method, Laser method and Plasma method.

  1. 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

  2. Advances in uranium enrichment processes

    International Nuclear Information System (INIS)

    Rae, H.K.; Melvin, J.G.; Slater, J.B.

    1986-05-01

    Advances in gas centrifuges and development of the atomic vapour laser isotope separation process promise substantial reductions in the cost of enriched uranium. The resulting reduction in LWR fuel costs could seriously erode the economic advantage of CANDU, and in combination with LWR design improvements, shortened construction times and increased operational reliability could allow the LWR to overtake CANDU. CANDU's traditional advantages of neutron economy and high reliability may no longer be sufficient - this is the challenge. The responses include: combining neutron economy and dollar economy by optimizing CANDU for slightly enriched uranium fuel; developing cost-reducing improvements in design, manufacture and construction; and reducing the cost of heavy water. Technology is a renewable resource which must be continually applied to a product for it to remain competitive in the decades to come. Such innovation is a prerequisite to Canada increasing her share of the international market for nuclear power stations. The higher burn-up achievable with enriched fuel in CANDU can reduce the fuel cycle costs by 20 to 40 percent for a likely range of costs for yellowcake and separative work. Alternatively, some of the benefits of a higher fissile content can take the form of a cheaper reactor core containing fewer fuel channels and less heavy water, and needing only a single fuelling machine. An opportunity that is linked to this need to introduce an enriched uranium fuel cycle into CANDU is to build an enrichment business in Canada. This could offer greater value added to our uranium exports, security of supply for enriched CANDUs, technological growth in Canada and new employment opportunities. AECL has a study in progress to define this opportunity

  3. Safety concerning the alteration in fuel material usage (new installation of the uranium enrichment pilot plant) at Ningyo Pass Mine of Power Reactor and Nuclear Fuel Development Corporation

    International Nuclear Information System (INIS)

    1978-01-01

    A report of the Committee on Examination of Nuclear Fuel Safety was presented to the Atomic Energy Commission of Japan, which is concerned with the safety in the alteration of fuel material usage (new installation of the uranium enrichment pilot plant) at the Ningyo Pass Mine. Its safety was confirmed. The alteration, i.e. installation of the uranium enrichment pilot plant, is as follows. Intended for the overall test of centrifugal uranium enrichment technology, the pilot plant includes a two-storied main building of about 9,000 m 2 floor space, containing centrifuges, UF 6 equipment, etc., a uranium storage of about 1,000 m 2 floor space, and a waste water treatment facility, two-storied with about 300 m 2 floor space. The contents of the examination are safety of the facilities, criticality control, radiation control, waste treatment, and effects of accidents on the surrounding environment. (Mori, K

  4. 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)

  5. Uranium enrichment: an overview

    International Nuclear Information System (INIS)

    Cazalet, J.

    1995-01-01

    This paper is a general presentation of uranium enrichment processes and assessments of the prevailing commercial and industrial situations. It gives first some theoretical aspects of enrichment in general and explains the differences between statistical and selective processes in particular. Then a review of the different processes is made with a comparison between them. Finally, some general remarks concerning applications are given and the risks of proliferation related to enrichment are mentioned. (J.S.). 4 refs., 5 figs., 8 tabs

  6. Uranium enrichment. Technology, economics, capacity

    International Nuclear Information System (INIS)

    Voigt, W.R. Jr.; Saire, D.E.; Gestson, D.K.; Peske, S.E.; Vanstrum, P.R.

    1983-01-01

    Large-scale enrichment of uranium has now been carried out for 40 years. While the gaseous diffusion process was the original choice of several countries and continues today to provide the major component of the world production of separative work, the last two decades have witnessed the development of a number of alternative processes for enrichment. These processes, which are being studied and deployed around the world, offer a wide range of technical and economic characteristics which will be useful in assuring adequate capacity to meet projected reactor fuel market needs through the rest of this century at competitive prices. With present uncertainties in future enriched uranium needs, it is apparent that flexibility in the deployment and operation of any enrichment process will be one of the prime considerations for the future. More economical production of separative work not only can have a beneficial impact on reactor fuel costs, but also tends to conserve natural uranium resources. This paper reviews the world scene in the enrichment component of the fuel cycle, including existing or planned commercial-scale facilities and announced R+D efforts on various processes. (author)

  7. Uranium enrichment: technology, economics, capacity

    Energy Technology Data Exchange (ETDEWEB)

    Voigt, Jr., W. R.; Vanstrum, P. R.; Saire, D. E.; Gestson, D. K.; Peske, S. E.

    1982-08-01

    Large-scale enrichment of uranium has now been carried out for 40 years. While the gaseous diffusion process was the original choice of several countries and continues today to provide the major component of the world production of separative work, the last two decades have witnessed the development of a number of alternative processes for enrichment. These processes, which are being studied and deployed around the world, offer a wide range of technical and economic characteristics which will be useful in assuring adequate capacity to meet projected reactor fuel market needs through the rest of this century at competitive prices. With present uncertainties in future enriched uranium needs, it is apparent that flexibility in the deployment and operation of any enrichment process will be one of the prime considerations for the future. More economical production of separative work not only can have a beneficial impact on reactor fuel costs, but also tends to conserve natural uranium resources. This paper reviews the world scene in the enrichment component of the fuel cycle, including existing or planned commercial-scale facilities and announced R and D efforts on various processes.

  8. Uranium enrichment: technology, economics, capacity

    International Nuclear Information System (INIS)

    Voigt, W.R. Jr.; Vanstrum, P.R.; Saire, D.E.; Gestson, D.K.; Peske, S.E.

    1982-01-01

    Large-scale enrichment of uranium has now been carried out for 40 years. While the gaseous diffusion process was the original choice of several countries and continues today to provide the major component of the world production of separative work, the last two decades have witnessed the development of a number of alternative processes for enrichment. These processes, which are being studied and deployed around the world, offer a wide range of technical and economic characteristics which will be useful in assuring adequate capacity to meet projected reactor fuel market needs through the rest of this century at competitive prices. With present uncertainties in future enriched uranium needs, it is apparent that flexibility in the deployment and operation of any enrichment process will be one of the prime considerations for the future. More economical production of separative work not only can have a beneficial impact on reactor fuel costs, but also tends to conserve natural uranium resources. This paper reviews the world scene in the enrichment component of the fuel cycle, including existing or planned commercial-scale facilities and announced R and D efforts on various processes

  9. Present state of development of uranium enrichment

    International Nuclear Information System (INIS)

    1979-01-01

    The pilot plant for uranium enrichment started the operation on September 12, 1979. The pilot plant has been constructed by the Power Reactor and Nuclear Fuel Development Corp. in Ningyo Pass, Okayama Prefecture. 7000 centrifugal separators will be installed by mid 1981, and yearly production of 70 t SWU is expected. The Uranium Enrichment Committee of Japan Atomic Industrial Forum has made the proposal on the method of forwarding the development of uranium enrichment in Japan to Atomic Energy Commission and related government offices in December, 1978. This survey summarized the trends of uranium enrichment in Japan and foreign countries and the problems about nuclear non-proliferation, and provides with the reference materials. The demand and supply of uranium enrichment in the world, the present states and plans in USA, Europe, USSR and others, the demand and supply of uranium enrichment and the measures for securing it in Japan, the present state and future plan of uranium enrichment project in Japan, the international regulation of uranium enrichment, the recent policy of USA and INFCE, and the trend of the regulation of utilizing enriched uranium are described. Moreover, the concept of separation works in uranium enrichment and the various technologies of separation are explained. (Kako, I.)

  10. Neutronics and thermalhydraulics characteristics of the CANDU core fueled with slightly enriched uranium 0.9% U235

    International Nuclear Information System (INIS)

    Raica, V.; Sindile, A.

    1999-01-01

    The interest concerning the slightly enriched uranium (SEU) fuel cycle is due to the possibility to adapt (to convert) the current reactor design using natural uranium fuel to this cycle. Preliminary evaluations based on discharged fuel burnup estimates versus enrichment and on Canadian experience in fuel irradiation suggest that for a 0.93% U-235 enrichment no design modifications are required, not even for the fuel bundle. The purpose of this paper is to resume the results of the studies carried on in order to clarify this problem. The calculation methodology used in reactor physics and thermal-hydraulics analyses that were performed adapted and developed the AECL suggested methodology. In order to prove the possibility to use the SEU 0.93% without any design modification, all the main elements from the CANDU Reactor Physics Design Manual were studied. Also, some thermal-hydraulics analyses were performed to ensure that the operating and safety parameters were respected. The estimations sustain the assumption that the current reactor and fuel bundle design is compatible to the using of the SEU 0.93% fuel. (author)

  11. Report of the Subcommittee on Domestic Uranium Enrichment

    International Nuclear Information System (INIS)

    1981-01-01

    A report by the Subcommittee on Domestic Uranium Enrichment to the Atomic Energy Commission is described; which covers the procedure of the domestic uranium enrichment by centrifugal process up to the commercial production, reviewing the current situation in this field. Domestic uranium enrichment is important in the aspects of securing stable enrichment service, establishing sound fuel cycle, and others. As the future target, the production around the year 2000 is set at 3,000 tons SWU per year at least. The business of uranium enrichment, which is now developed in the Power Reactor and Nuclear Fuel Development Corporation, is to be carried out by private enterprise. The contents are as follows: demand and supply balance of uranium enrichment service, significance of domestic uranium enrichment, evaluation of centrifugal uranium enrichment technology, the target of domestic uranium enrichment, the policy of domestic uranium enrichment promotion. (J.P.N.)

  12. 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)

  13. Conversion of the RB reactor neutrons by highly enriched uranium fuel and lithium deuteride

    International Nuclear Information System (INIS)

    Strugar, P.; Sotic, O.; Ninkovic, M.; Pesic, M.; Altiparmakov, D.

    1981-01-01

    A thermal-to-fast-neutron converter has been constructed at the RB reactor. The material used for the conversion of thermal neutrons is highly enriched uranium fuel of Soviet production applied in Yugoslav heavy water experimental reactors RA and RB. Calculations and preliminary measurements show that the spectrum of converted neutrons only slightly differs from that of fission neutrons. The basic characteristics of converted neutrons can be expressed by the neutron radiation dose of 800 rad (8 Gy) for 1 h of reactor operation at a power level of 1 kW. This dose is approximately 10 times higher than the neutron dose at the same place without converter. At the same time, thermal neutron and gamma radiation doses are negligible. The constructed neutron converter offers wide possibilities for applications in reactor and nuclear physics and similar disciplines, where neutron spectra of high energies are required, as well as in the domain of neutron dosimetry and biological irradiations in homogeneous fields of larger dimensions. The possibility of converting thermal reactor neutrons with energies of about 14 MeV with the aid of lithium deuteride from natural lithium has been considered too. (author)

  14. 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

  15. International collaboration to study the feasibility of implementing the use of slightly enriched uranium fuel in the Embalse CANDU reactor

    International Nuclear Information System (INIS)

    Rouben, B.; Chow, H.C.; Leung, L.K.H.; Inch, W.; Fink, J.; Moreno, C.

    2004-01-01

    In the last few years, Nucleoelectrica Argentina S.A. and Atomic Energy of Canada Limited have collaborated on a study of the technical feasibility of implementing Slightly Enriched Uranium (SEU) fuel in the Embalse CANDU reactor in Argentina. The successful conversion to SEU fuel of the other Argentine heavy-water reactor, Atucha 1, served as a good example. SEU presents an attractive incentive from the point of view of fuel utilization: if fuel enriched to 0.9% 235 U were used in Embalse instead of natural uranium, the average fuel discharge burnup would increase significantly (by a factor of about 2), with consequent reduction in fuel requirements, leading to lower fuel-cycle costs and a large reduction in spent-fuel volume per unit energy produced. Another advantage is the change in the axial power shape: with SEU fuel, the maximum bundle power in a channel decreases and shifts towards the coolant inlet end, consequently increasing the thermalhydraulics safety margin. Two SEU fuel carriers, the traditional 37-element bundle and the 43-element CANFLEX bundle, which has enhanced thermalhydraulic characteristics as well as lower peak linear element ratings, have been examined. The feasibility study gave the organizations an excellent opportunity to perform cooperatively a large number of analyses, e.g., in reactor physics, thermalhydraulics, fuel performance, and safety. A Draft Plan for a Demonstration Irradiation of SEU fuel in Embalse was prepared. Safety analyses have been performed for a number of hypothetical accidents, such as Large Loss of Coolant, Loss of Reactivity Control, and an off-normal condition corresponding to introducing 8 SEU bundles in a channel (instead of 2 or 4 bundles). There are concrete safety improvements which result from the reduced maximum bundle powers and their shift towards the inlet end of the fuel channel. Further improvements in safety margins would accrue with CANFLEX. In conclusion, the analyses identified no issues that

  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. ALARA (As Low As Reasonable Achievable) procedure applied to fuel assembly fabrication with enriched reprocessing uranium (ERU)

    International Nuclear Information System (INIS)

    Guimaraes, Leonam dos Santos; Degrange, Jean Pierre

    1998-01-01

    The study introduced by this paper compose the first step to the implementation of ALARA (As Low As Reasonable Achievable) for a nuclear fuel assembly factory which one of its two production lines will be designed to work with Enriched Reprocessing Uranium (ERU). This step includes the reference situation analysis is based on previsional dosimetric evaluations for individual and collective exposures of each factory operator (117 in total) working on 7 work stations, considering 6 annual production scenarios (10, 50 75, 100 and 150 ERU tons), which corresponds to an annual production of 600 tons (ERU plus enriched natural uranium ENU). The exposure indicators evolution, expressed in terms of collective dose, annual individual dose and radiological detrimental cost for workers, is also used in a complimentary way to guide the analysis. (author)

  18. Advanced uranium enrichment processes

    International Nuclear Information System (INIS)

    Clerc, M.; Plurien, P.

    1986-01-01

    Three advanced Uranium enrichment processes are dealt with in the report: AVLIS (Atomic Vapour LASER Isotope Separation), MLIS (Molecular LASER Isotope Separation) and PSP (Plasma Separation Process). The description of the physical and technical features of the processes constitutes a major part of the report. If further presents comparisons with existing industrially used enrichment technologies, gives information on actual development programmes and budgets and ends with a chapter on perspectives and conclusions. An extensive bibliography of the relevant open literature is added to the different subjects discussed. The report was drawn up by the nuclear research Centre (CEA) Saclay on behalf of the Commission of the European Communities

  19. 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)

  20. Nonproliferation and safeguards aspects of fuel cycle programs in reduction of excess separated plutonium and high-enriched uranium

    International Nuclear Information System (INIS)

    Persiani, P.J.

    1995-01-01

    The purpose of this preliminary investigation is to explore alternatives and strategies aimed at the gradual reduction of the excess inventories of separated plutonium and high-enriched uranium (HEU) in the civilian nuclear power industry. The study attempts to establish a technical and economic basis to assist in the formation of alternative approaches consistent with nonproliferation and safeguards concerns. Reference annual mass flows and inventories for a representative 1,400 Mwe Pressurized Water Reactor (PWR) fuel cycle have been investigated for three cases: the 100 percent uranium oxide UO 2 fuel loading once through cycle, and the 33 percent mixed oxide MOX loading configuration for a first and second plutonium recycle. The analysis addresses fuel cycle developments; plutonium and uranium inventory and flow balances; nuclear fuel processing operations; UO 2 once-through and MOX first and second recycles; and the economic incentives to draw-down the excess separated plutonium stores. The preliminary analysis explores several options in reducing the excess separated plutonium arisings and HEU, and the consequences of the interacting synergistic effects between fuel cycle processes and isotopic signatures of nuclear materials on nonproliferation and safeguards policy assessments

  1. South Australia, uranium enrichment

    International Nuclear Information System (INIS)

    1976-02-01

    The Report sets out the salient data relating to the establishment of a uranium processing centre at Redcliff in South Australia. It is conceived as a major development project for the Commonwealth, the South Australian Government and Australian Industry comprising the refining and enrichment of uranium produced from Australian mines. Using the data currently available in respect of markets, demand, technology and possible financial return from overseas sales, the project could be initiated immediately with hexafluoride production, followed rapidly in stages by enrichment production using the centrifuge process. A conceptual development plan is presented, involving a growth pattern that would be closely synchronised with the mining and production of yellowcake. The proposed development is presented in the form of an eight-and-half-year programme. Costs in this Report are based on 1975 values, unless otherwise stated. (Author)

  2. 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

  3. Developments in uranium enrichment

    International Nuclear Information System (INIS)

    Mohrhauer, H.

    1995-01-01

    The enrichment services market is still characterized by overcapacities. While consumption worldwide will rise by some 15% to 39,000 t SWU/a over the next ten years, capacities amount to nearly 50,000 t SWU/a. The price for enrichment services probably has reached its all time low. Prices below U.S. $ 100/kg SWU are not likely to cover costs even of the economically most advanced enrichment processes. Urenco has prepared for the difficult enrichment business in the years to come by streamlining and cost cutting measures. The company intends to hold and increase its share of more than 10% in the world market. The uranium enrichment plant of Gronau will be expanded further. Expansion beyond 1000 t is subject to another permit being granted under the Atomic Energy Act, an application for which was filed in December 1994. Centrifuge technology is the superior enrichment technology, i.e., there is still considerable potential for further development. Construction of enrichment plants employing the centrifuge technology in the United States and in France is being pursued in various phases, from feasibility studies to licensing procedures. Before these plants could be implemented, however, considerable problems of organization would have to be solved, and the market would have to change greatly, respectively. The laser process, at the present time, does not seem to be able to develop into a major industrial competitor. (orig.) [de

  4. 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

  5. Allocation of uranium enrichment services to fuel foreign and domestic nuclear reactors

    International Nuclear Information System (INIS)

    1975-01-01

    This interim report was made in response to a request for information concerning the sale of U.S. uranium enrichment services to foreign countries and its effect on AEC's ability to meet domestic demands. Long-term enrichment services (June 30, 1974), both domestic and foreign, totaled 364,000 MW, or 44,000 MW more than its available capability. The first-come-first-served policy was modified to give preferential treatment to Yugoslav and Mexican requests because of IAEA commitments, and to shift six standard contracts from Japan. From Aug. to Sept. 1974, standard contracts were signed for all 15 pending domestic requests and for 33 pending foreign requests, with the remaining 45 foreign requests depending on NRC's approval of Pu recycle, although private enrichment or stockpile enriched U could meet these needs. There is no firm commitment in the private sector to build and operate the needed enrichment plant. The acceleration of foreign nuclear programs coupled with ERDA's termination of further long-term contracts, may lead to the emergence of foreign supply sources, and U.S. may lose its favorable balance-of-payments and its influence on international nuclear policies

  6. 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)

  7. United States uranium enrichment policies

    International Nuclear Information System (INIS)

    Roberts, R.W.

    1977-01-01

    ERDA's uranium enrichment program policies governing the manner in which ERDA's enrichment complex is being operated and expanded to meet customer requirements for separative work, research and development activities directed at providing technology alternatives for future enrichment capacity, and establishing the framework for additional domestic uranium enrichment capacity to meet the domestic and foreign nuclear industry's growing demand for enrichment services are considered. The ERDA enrichment complex consists of three gaseous diffusion plants located in Oak Ridge, Tennessee; Paducah, Kentucky; and Portsmouth, Ohio. Today, these plants provide uranium enrichment services for commercial nuclear power generation. These enrichment services are provided under contracts between the Government and the utility customers. ERDA's program involves a major pilot plant cascade, and pursues an advanced isotope separation technique for the late 1980's. That the United States must develop additional domestic uranium enrichment capacity is discussed

  8. Uranium enrichment: a vital new industry

    International Nuclear Information System (INIS)

    1975-10-01

    The energy problem facing the nation and the need for nuclear power are pointed out. Uranium enrichment and the demand for it are discussed. Reasons for, and obstacles to, private enrichment are outlined. The President's plan (including the Nuclear Fuel Assurance Act) is summarized

  9. The future cost of uranium enrichment

    International Nuclear Information System (INIS)

    Pouris, A.

    1986-01-01

    The cost of uranium enrichment is the most important factor determining the fuel cost of nuclear energy. This paper attempts to forecast the future direction of the price of separative work by examining the forces that determine it. It is argued that the interplay among the characteristics of enrichment technologies, the structure of the international market, and the balance of supply and demand determine the enrichment price. The analysis indicates that all forces point towards a price much lower than the current one. It is predicted that, depending on the technological advances, the price of separative work unit for uranium enrichment will range between $40 and $90 by the year 2000. (author)

  10. 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

  11. 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

  12. Critical experiments simulating accidental water immersion of highly enriched uranium dioxide fuel elements

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoi, N.N.; Glushkov, L.S.

    2003-01-01

    The paper focuses on experimental analysis of nuclear criticality safety at accidental water immersion of fuel elements of the Russian TOPAZ-2 space nuclear power system reactor. The structure of water-moderated heterogeneous critical assemblies at the NARCISS facility is described in detail, including sizes, compositions, densities of materials of the main assembly components for various core configurations. Critical parameters of the assemblies measured for varying number of fuel elements, height of fuel material in fuel elements and their arrangement in the water moderator with a uniform or variable spacing are presented. It has been found from the experiments that at accidental water immersion of fuel elements involved, the minimum critical mass equal to approximately 20 kg of uranium dioxide is achieved at 31-37 fuel elements. The paper gives an example of a physical model of the water-moderated heterogeneous critical assembly with a detailed characterization of its main components that can be used for calculations using different neutronic codes, including Monte Carlo ones. (author)

  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. Development of ISA procedure for uranium fuel fabrication and enrichment facilities: overview of ISA procedure and its application

    International Nuclear Information System (INIS)

    Yamate, Kazuki; Yamada, Takashi; Takanashi, Mitsuhiro; Sasaki, Noriaki

    2013-01-01

    Integrated Safety Analysis (ISA) procedure for uranium fuel fabrication and enrichment facilities has been developed for aiming at applying risk-informed regulation to these uranium facilities. The development has carried out referring to the ISA (NUREG-1520) by the Nuclear Regulatory Commission (NRC). The paper presents purpose, principles and activities for the development of the ISA procedure, including Risk Level (RL) matrix and grading evaluation method of IROFS (Items Relied on for Safety), as well as general description and features of the procedure. Also described in the paper is current status in application of risk information from the ISA. Japanese four licensees of the uranium facilities have been conducting ISA for their representative processes using the developed procedure as their voluntary safety activities. They have been accumulating experiences and knowledge on the ISA procedure and risk information through the field activities. NISA (Nuclear and Industrial Safety Agency) and JNES (Japan Nuclear Energy Safety Organization) are studying how to use such risk information for the safety regulation of the uranium facilities, taking into account the licensees' experiences and knowledge. (authors)

  15. 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)

  16. 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)

  17. Uranium enrichment services in the United States

    International Nuclear Information System (INIS)

    Jelinek, P.; Lenders, M.

    1994-01-01

    The United States of America is the world's largest market for uranium enrichment services. After the disintegration of the Soviet Union, Russian uranium is entering the world market on an increasing scale. The U.S. tries to protect its market and, in this connection, also the European market from excessive price drops by taking anti-dumping measures. In order to become more competitive, American companies have adapted modern enrichment techniques from Europe. European - U.S. joint ventures are to help, also technically and economically, to integrate military uranium, accumulating as a consequence of worldwide disarmament, into the commercial fuel cycle for the peaceful use of nuclear power. (orig.) [de

  18. Uranium enriched granites in Sweden

    International Nuclear Information System (INIS)

    Wilson, M.R.; Aakerblom, G.

    1980-01-01

    Granites with uranium contents higher than normal occur in a variety of geological settings in the Swedish Precambrian, and represent a variety of granite types and ages. They may have been generated by (1) the anatexis of continental crust (2) processes occurring at a much greater depth. They commonly show enrichement in F, Sn, W and/or Mo. Only in one case is an important uranium mineralization thought to be directly related to a uranium-enriched granite, while the majority of epigenetic uranium mineralizations with economic potential are related to hydrothermal processes in areas where the bedrock is regionally uranium-enhanced. (Authors)

  19. NARCISS critical stand experiments for studying the nuclear safety in accident water immersion of highly enriched uranium dioxide fuel elements

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoj, N.N.; Glushkov, E.S.; Bubelev, V.G.

    2005-01-01

    A brief description of the Topaz-2 SNPS designed under scientific supervision of RRC KI in Russia, and of the NARCISS critical facility, is given. At the NARCISS critical facility, neutronic peculiarities and nuclear safety issues of the Topaz-2 system reactor were studied experimentally. This work is devoted to a detailed description of experiments on investigation of criticality safety in accident water immersion og highly enriched uranium dioxide fuel elements, performed at the NARCISS facility. The experiments were carried out at water-moderated critical assemblies with varying height, number, and spacing of fuel elements. The results obtained in the critical experiments, computational models of the investigated critical configurations, and comparison of the computational and experimental results are given [ru

  20. Uranium enrichment capacity: public versus private ownership

    International Nuclear Information System (INIS)

    Fraser, J.T.

    1977-01-01

    Continual growth of conventional nuclear capacity requires an assured supply of enriched uranium and, hence, potential expansion of domestic uranium enrichment capacity. The question of ownership of new enrichment capacity, i.e., public or private, entails not only the social-opportunity costs of alternative investments but also technical parameters of uranium utilization and advanced reactor development. Inclusion of risk preferences in both the public and private sectors produces interesting results in terms of optimal investment strategies with respect to choice of technology and scale of investment. Utilization of a nuclear fuel cycle requirements process model allows explicit specification of production technology. Integration of process model output with a least-cost investment model permits flexibility in parametric analysis. Results indicate minimum incentive for Government subsidy of a private enrichment sector through 2000 given moderate to low nuclear growth assumptions. The long-run scenario, to 2020, exhibits potentially greater incentives for private enrichment investment

  1. World nuclear-fuel procurement: relationships between uranium and enrichment markets. Final report. International energies studies program

    International Nuclear Information System (INIS)

    Neff, T.L.

    1982-03-01

    This article explores the relationships between international uranium and enrichment markets under current contracting and equity arrangements and in comparison with actual feed requirements for existing and committed reactors. We begin with an overview of the world situation, examining current and prospective conditions. We then consider enrichment and uranium supply and demand situations of the three consumer nations outside the United States with the largest nuclear programs: France, Japan, and the Federal Republic of Germany. We conclude with an evaluation of likely directions of change in the coupled markets for uranium and enrichment services

  2. AEC determines uranium enrichment policy

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The Advisory Committee on Uranium Enrichment of the Atomic Energy Commission (AEC) has submitted a report to AEC chairman concerning the promotion of the introduction of advanced material, high performance centrifuges to replace conventional metallic drum centrifuges, and the development of next generation advanced centrifuges. The report also called for the postponement until around 1997 of the decision whether the development should be continued or not on atomic vapor laser isotope separation (AVLIS) and molecular laser isotope separation (MLIS) processes, as well as the virtual freezing of the construction of a chemical process demonstration plant. The report was approved by the AEC chairman in August. The uranium enrichment service market in the world will continue to be characterized by oversupply. The domestic situation of uranium enrichment supply-demand trend, progress of the expansion of Rokkasho enrichment plant, the trend in the development of gas centrifuge process and the basic philosophy of commercializing domestic uranium enrichment are reported. (K.I.)

  3. 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.

  4. 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

  5. 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

  6. Update on international uranium and enrichment supply

    International Nuclear Information System (INIS)

    Cleveland, J.M.

    1987-01-01

    Commercial nuclear power generation came upon us in the late 1950s and should have been relatively uneventful due to its similarities to fossil-powered electrical generation. Procurement of nuclear fuel appears to have been treated totally different from the procurement of fossil fuel, however, and only recently have these practices started to change. The degree of utility reliance on US-mined uranium and US Dept. of Energy (DOE)-produced enrichment services has changed since the 1970s as federal government uncertainty, international fuel market opportunity, and public service commission scrutiny has increased. Accordingly, the uranium and enrichment market has recognized that it is international just like the fossil fuel market. There is now oversupply-driven competition in the international nuclear fuel market. Competition is increasing daily, as third-world countries develop their own nuclear resources. American utilities are now diversifying their fuel supply arrangements, as they do with their oil, coal, and gas supply. The degree of foreign fuel arrangements depends on each utility's risk posture and commitment to long-term contracts. In an era of rising capital, retrofit, operating, and maintenance costs, economical nuclear fuel supply is even more important. This economic advantage, however, may be nullified by congressional and judicial actions limiting uranium importation and access to foreign enrichment. Such artificial trade barriers will only defeat US nuclear generation and the US nuclear fuel industry in the long term

  7. Enriched uranium cycles in pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Mazzola, A.

    1994-01-01

    A study was made on the substitution of natural uranium with enriched and on plutonium recycle in unmodified PHWRs (pressure vessel reactor). Results clearly show the usefulness of enriched fuel utilisation for both uranium ore consumption (savings of 30% around 1.3% enrichment) and decreasing fuel cycle coasts. This is also due to a better plutonium exploitation during the cycle. On the other hand plutonium recycle in these reactors via MOX-type fuel appears economically unfavourable under any condition

  8. 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

  9. U.S. forms uranium enrichment corporation

    International Nuclear Information System (INIS)

    Seltzer, R.

    1993-01-01

    After almost 40 years of operation, the federal government is withdrawing from the uranium enrichment business. On July 1, the Department of Energy turned over to a new government-owned entity--the US Enrichment Corp. (USEC)--both the DOE enrichment plants at Paducah, Ky., and Portsmouth, Ohio, and domestic and international marketing of enriched uranium from them. Pushed by the inability of DOE's enrichment operations to meet foreign competition, Congress established USEC under the National Energy Policy Act of 1992, envisioning the new corporation as the first step to full privatization. With gross revenues of $1.5 billion in fiscal 1992, USEC would rank 275th on the Fortune 500 list of top US companies. USEC will lease from DOE the Paducah and Portsmouth facilities, built in the early 1950s, which use the gaseous diffusion process for uranium enrichment. USEC's stock is held by the US Treasury, to which it will pay annual dividends. Martin Marietta Energy Systems, which has operated Paducah since 1984 and Portsmouth since 1986 for DOE, will continue to operate both plants for USEC. Closing one of the two facilities will be studied, especially in light of a 40% world surplus of capacity over demand. USEC also will consider other nuclear-fuel-related ventures. USEC will produce only low-enriched uranium, not weapons-grade material. Indeed, USEC will implement a contract now being completed under which the US will purchase weapons-grade uranium from dismantled Russian nuclear weapons and convert it into low-enriched uranium for power reactor fuel

  10. 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)

  11. Ningyo Toge uranium enrichment pilot plant comes into full

    International Nuclear Information System (INIS)

    1982-01-01

    The uranium enrichment pilot plant of the Power Reactor and Nuclear Fuel Development Corporation at Ningyo Toge went into full operation on March 26, 1982. This signifies that the front end of the nuclear fuel cycle in Japan, from uranium ore to enrichment, is only a step away from commercialization. On the same day, the pilot plant of uranium processing and conversion to UF 6 , the direct purification of uranium ore into uranium hexafluoride, began batch operation at the same works. The construction of the uranium enrichment pilot plant has been advanced in three stages: i.e. OP-1A with 1000 centrifuges, OP-1B with 3000 centrifuges and OP-2 with 3000 centrifuges. With a total of 7000 centrifuges, the pilot plant, the first enrichment plant in Japan, has now a capacity of supplying enriched uranium for six months operation of a 1,000 MW nuclear power plant. (J.P.N.)

  12. Standard specification for uranium metal enriched to more than 15 % and less Than 20 % 235U

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2000-01-01

    1.1 This specification covers nuclear grade uranium metal that has either been processed through an enrichment plant, or has been produced by the blending of highly enriched uranium with other uranium, to obtain uranium of any 235U concentration below 20 % (and greater than 15 %) and that is intended for research reactor fuel fabrication. The scope of this specification includes specifications for enriched uranium metal derived from commercial natural uranium, recovered uranium, or highly enriched uranium. Commercial natural uranium, recovered uranium and highly enriched uranium are defined in Section 3. The objectives of this specification are to define the impurity and uranium isotope limits for commercial grade enriched uranium metal. 1.2 This specification is intended to provide the nuclear industry with a standard for enriched uranium metal which is to be used in the production of research reactor fuel. In addition to this specification, the parties concerned may agree to other appropriate conditions. ...

  13. 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.

  14. 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

  15. Feeding the nuclear fuel cycle with a long term view; AREVA's front-end business units, uranium mining, UF6 conversion and isotopic enrichment

    International Nuclear Information System (INIS)

    Capus, G.A.P.; Autegert, R.

    2005-01-01

    As a leading provider of technological solutions for nuclear power generation and electricity transmission, the AREVA group has the unique capability of offering a fully integrated fuel supply, when requested by its customers. At the core of the AREVA group, COGEMA Front End Division is an essential part of the overall fuel supply chain. Composed of three Business Units and gathering several subsidiaries and joint 'ventures, this division enjoys several leading positions as shown by its market shares and historical production records. Current Uranium market evolutions put the natural uranium supply under focus. The uranium conversion segment also recently revealed some concerning evolutions. And no doubt, the market pressure will soon be directed also at the enrichment segment. Looking towards the long term, AREVA strongly believes that a nuclear power renewal is needed, especially to help limiting green house effect gas release. Therefore, to address future supplies needed to fuel the existing fleet of nuclear power plants, but also new ones, the AREVA group is planning very significant investments to build new facilities in all the three front-end market segments. As far as uranium mining is concerned, these new mines will be based upon uranium reserves of outstanding quality. As for uranium conversion and enrichment, two large projects will be based on the most advanced technologies. This paper is aimed at recalling COGEMA Front End Division experience, the current status of its plants and operating entities and will provide a detailed overview of its major projects. (authors)

  16. Optimization of enrichment distributions in nuclear fuel assemblies loaded with Uranium and Plutonium via a modified linear programming technique

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas Vivas, Gabriel Francisco

    1999-12-01

    A methodology to optimize enrichment distributions in Light Water Reactor (LWR) fuel assemblies is developed and tested. The optimization technique employed is the linear programming revised simplex method, and the fuel assembly's performance is evaluated with a neutron transport code that is also utilized in the calculation of sensitivity coefficients. The enrichment distribution optimization procedure begins from a single-value (flat) enrichment distribution until a target, maximum local power peaking factor, is achieved. The optimum rod enrichment distribution, with 1.00 for the maximum local power peaking factor and with each rod having its own enrichment, is calculated at an intermediate stage of the analysis. Later, the best locations and values for a reduced number of rod enrichments is obtained as a function of a target maximum local power peaking factor by applying sensitivity to change techniques. Finally, a shuffling process that assigns individual rod enrichments among the enrichment groups is performed. The relative rod power distribution is then slightly modified and the rod grouping redefined until the optimum configuration is attained. To verify the accuracy of the relative rod power distribution, a full computation with the neutron transport code using the optimum enrichment distribution is carried out. The results are compared and tested for assembly designs loaded with fresh Low Enriched Uranium (LEU) and plutonium Mixed Oxide (MOX) isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide range of applicability of the optimization technique. The feature of the assembly designs used for evaluation purposes included burnable absorbers and internal water regions, and were prepared to resemble the configurations of modern assemblies utilized in commercial Boiling Water Reactor (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or in the

  17. Enriched uranium recovery at Los Alamos

    International Nuclear Information System (INIS)

    Herrick, C.C.

    1984-01-01

    Graphite casting scrap, fuel elements and nongraphite combustibles are calcined to impure oxides. These materials along with zircaloy fuel elements and refractory solids are leach-dissolved separately in HF-HNO 3 acid to solubilize the contained enriched uranium. The resulting slurry is filtered and the clear filtrate (to which mineral acid solutions bearing enriched uranium may be added) are passed through solvent extraction. The solvent extraction product is filtered, precipitated with H 2 O 2 and the precipitate calcined to U 3 O 8 . Metal is made from U 3 O 8 by conversion to UO 2 , hydrofluorination and reduction to metal. Throughput is 150 to 900 kg uranium per year depending on the type of scrap

  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. Supply of enriched uranium for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H. [NUKEM GmbH, Alzenau (Germany)

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

  20. Supply of enriched uranium for research reactors

    International Nuclear Information System (INIS)

    Mueller, H.

    1997-01-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA's ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future

  1. 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

  2. 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.

  3. Experiments on light water lattices with enriched uranium fuel; Analyse des donnees experimentales sur les reseaux a eau legere et uranium enrichi

    Energy Technology Data Exchange (ETDEWEB)

    Audinet, M [Societe des Forges et Ateliers du Creusot, 75 - Paris (France); Lamare, J de [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Panossian, J [Societe Alsacienne de Constructions Mecaniques (France)

    1958-07-01

    Experiments a light water lattices with slightly enriched uranium fuel, have been performed at Brookhaven and Bettis Plant Laboratories. The results are studied and compared with simple theories on reactor calculations. By taking into account shadow effects and non Maxwellian neutron spectrum, which are important in this kind of reactors, we have been able to explain the observed results fairly well. We can thus give a constituent set of formulas with which to calculate lattices similar to there we studied. (author) [French] Les resultats d'experiences effectuees aux Laboratoires de Brookbaven et de Bettis Plant, sur des reseaux heterogenes a eau legere et uranium metallique legerement enrichi, sont analyses et confrontes avec les theories simples du calcul de pile. En tenant compte des effets d'interaction et d'echauffement du spectre de neutrons qui sont importants dans ce type de reacteurs, on parvient a rendre compte convenablement des resultats observes. On a ainsi mis au point un formulaire permettant le calcul des reseaux quivpeuvent etre consideres comme assez semblables aux reseaux etudies. (auteur)

  4. 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.

  5. Uranium enrichment by gas centrifuge

    International Nuclear Information System (INIS)

    Heriot, I.D.

    1988-01-01

    After recalling the physical principles and the techniques of centrifuge enrichment the report describes the centrifuge enrichment programmes of the various countries concerned and compares this technology with other enrichment technologies like gaseous diffusion, laser, aerodynamic devices and chemical processes. The centrifuge enrichment process is said to be able to replace with advantage the existing enrichment facilities in the short and medium term. Future prospects of the process are also described, like recycled uranium enrichment and economic improvements; research and development needs to achieve the economic prospects are also indicated. Finally the report takes note of the positive aspect of centrifuge enrichment as far as safeguards and nuclear safety are concerned. 27 figs, 113 refs

  6. Study of Reduced-Enrichment Uranium Fuel Possibility for Research Reactors

    Directory of Open Access Journals (Sweden)

    Ruppel V.A.

    2015-01-01

    Full Text Available Having analyzed the results obtained in the work, it is possible to conclude that the flux density of fast and thermal neutrons in the shell of fuel elements in EFA in REU-zone decreased on average by 5% for UO2 fuel and by 7% for U9%Mo fuel. Change of neutrons flux density during the cycle does not exceed 4% for both fuel types. On average the fuel burnup in reactor core during the cycle for UO2 and U9%Mo increased by 2.8%. It is 1% less that in HEU-zone, which is conditioned by higher initial loading of 235U in fuel assembly with REU fuel.

  7. Profile of World Uranium Enrichment Programs - 2007

    International Nuclear Information System (INIS)

    Laughter, Mark D.

    2007-01-01

    It is generally agreed that the most difficult step in building a nuclear weapon is acquiring weapons grade fissile material, either plutonium or highly enriched uranium (HEU). Plutonium is produced in a nuclear reactor, while HEU is produced using a uranium enrichment process. Enrichment is also an important step in the civil nuclear fuel cycle, in producing low enriched uranium (LEU) for use in fuel for nuclear reactors. However, the same equipment used to produce LEU for nuclear fuel can also be used to produce HEU for weapons. Safeguards at an enrichment plant are the array of assurances and verification techniques that ensure uranium is only enriched to LEU, no undeclared LEU is produced, and no uranium is enriched to HEU or secretly diverted. There are several techniques for enriching uranium. The two most prevalent are gaseous diffusion, which uses older technology and requires a lot of energy, and gas centrifuge separation, which uses more advanced technology and is more energy efficient. Gaseous diffusion plants (GDPs) provide about 40% of current world enrichment capacity, but are being phased out as newer gas centrifuge enrichment plants (GCEPs) are constructed. Estimates of current and future enrichment capacity are always approximate, due to the constant upgrades, expansions, and shutdowns occurring at enrichment plants, largely determined by economic interests. Currently, the world enrichment capacity is approximately 53 million kg-separative work units (SWU) per year, with 22 million in gaseous diffusion and 31 million in gas centrifuge plants. Another 23 million SWU/year of capacity are under construction or planned for the near future, almost entirely using gas centrifuge separation. Other less-efficient techniques have also been used in the past, including electromagnetic and aerodynamic separations, but these are considered obsolete, at least from a commercial perspective. Laser isotope separation shows promise as a possible enrichment technique

  8. 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

  9. Uranium enrichment: an evolving market

    International Nuclear Information System (INIS)

    Longenecker, J.; Witzel, R.

    1997-01-01

    With over half of the world uranium enrichment market uncommitted to any supplier early in the next century, competition is certain to be fierce. In the meantime the outlood remains unclear, with the market dominated by a number of developments -privatisation of the United States Enrichment Corp (USEC), increasing availability of Russian and US military inventories, the deployment of advanced technology and the closure of nuclear power plants due to deregulation. (author)

  10. Optimal set of selected uranium enrichments that minimizes blending consequences

    International Nuclear Information System (INIS)

    Nachlas, J.A.; Kurstedt, H.A. Jr.; Lobber, J.S. Jr.

    1977-01-01

    Identities, quantities, and costs associated with producing a set of selected enrichments and blending them to provide fuel for existing reactors are investigated using an optimization model constructed with appropriate constraints. Selected enrichments are required for either nuclear reactor fuel standardization or potential uranium enrichment alternatives such as the gas centrifuge. Using a mixed-integer linear program, the model minimizes present worth costs for a 39-product-enrichment reference case. For four ingredients, the marginal blending cost is only 0.18% of the total direct production cost. Natural uranium is not an optimal blending ingredient. Optimal values reappear in most sets of ingredient enrichments

  11. The isotopic enrichment of uranium in 1979

    International Nuclear Information System (INIS)

    Baron, M.

    1979-01-01

    The Eurodif uranium enrichment plant built on the Tricastin site is described. The uranium isotope separation plants in service abroad are presented. The main characteristics of the international enrichment market are defined [fr

  12. 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

  13. 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

  14. Enriched uranium recovery flowsheet improvements

    International Nuclear Information System (INIS)

    Holt, D.L.

    1986-01-01

    Savannah River uses 7.5% TBP to recover and purify enriched uranium. Adequate decontamination from fission products is necessary to reduce personnel exposure and to ensure that the enriched uranium product meets specifications. Initial decontamination of the enriched uranium from the fission products is carried out in the 1A bank, 16 stages of mixer-settlers. Separation of the enriched uranium from the fission product, 95 Zr, has been adequate, but excessive solvent degradation caused by the long phase contact times in the mixer-settlers has limited the 95 Zr decontamination factor (DF). An experimental program is investigating the replacement of the current 1A bank with either centrifugal contactors or a combination of centrifugal contactors and mixer-settlers. Experimental work completed has compared laboratory-scale centrifugal contactors and mixer-settlers for 95 Zr removal efficiencies. Feed solutions spiked with actual plant solutions were used. The 95 Zr DF was significantly better in the mixer-settlers than in the centrifugal contactors. As a result of this experimental study, a hybrid equipment flowsheet has been proposed for plant use. The hybrid equipment flowsheet combines the advantages of both types of solvent extraction equipment. Centrifugal contactors would be utilized in the extraction and initial scrub sections, followed by additional scrub stages of mixer-settlers

  15. Environmental Development Plan: uranium enrichment

    International Nuclear Information System (INIS)

    1979-09-01

    This Environmental Development Plan identifies and examines the environmental, health, safety, and socioeconomic concerns and corresponding requirements associated with the DOE research, development, demonstration, and operation of the Uranium Enrichment program, including the gaseous diffusion process, the centrifuge process, centrifuge rotor fabrication, and related research and development activities

  16. 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

  17. Safety of uranium enrichment plant

    International Nuclear Information System (INIS)

    Yonekawa, Shigeru; Morikami, Yoshio; Morita, Minoru; Takahashi, Tsukasa; Tokuyasu, Takashi.

    1991-01-01

    With respect to safety evaluation of the gas centrifuge enrichment facility, several characteristic problems are described as follows. Criticality safety in the cascade equipments can be obtained to maintain the enrichment of UF 6 below 5 %. External radiation dose equivalent rate of the 30B cylinder is low enough, the shield is not necessary. Penetration ratio of the two-stage HEPA filters for UF 6 aerosol is estimated at 10 -9 . From the experimental investigation, vacuum tightness is not damaged by destruction of gas centrifuge rotor. Carbon steel can be used for uranium enrichment equipments under the condition below 100degC. (author)

  18. 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

  19. Nuclear fuel cycle, nuclear fuel makes the rounds: choosing a closed fuel cycle, nuclear fuel cycle processes, front-end of the fuel cycle: from crude ore to enriched uranium, back-end of the fuel cycle: the second life of nuclear fuel, and tomorrow: multiple recycling while generating increasingly less waste

    International Nuclear Information System (INIS)

    Philippon, Patrick

    2016-01-01

    France has opted for a policy of processing and recycling spent fuel. This option has already been deployed commercially since the 1990's, but will reach its full potential with the fourth generation. The CEA developed the processes in use today, and is pursuing research to improve, extend, and adapt these technologies to tomorrow's challenges. France has opted for a 'closed cycle' to recycle the reusable materials in spent fuel (uranium and plutonium) and optimise ultimate waste management. France has opted for a 'closed' nuclear fuel cycle. Spent fuel is processed to recover the reusable materials: uranium and plutonium. The remaining components (fission products and minor actinides) are the ultimate waste. This info-graphic shows the main steps in the fuel cycle currently implemented commercially in France. From the mine to the reactor, a vast industrial system ensures the conversion of uranium contained in the ore to obtain uranium oxide (UOX) fuel pellets. Selective extraction, purification, enrichment - key scientific and technical challenges for the teams in the Nuclear Energy Division (DEN). The back-end stages of the fuel cycle for recycling the reusable materials in spent fuel and conditioning the final waste-forms have reached maturity. CEA teams are pursuing their research in support of industry to optimise these processes. Multi-recycle plutonium, make even better use of uranium resources and, over the longer term, explore the possibility of transmuting the most highly radioactive waste: these are the challenges facing future nuclear systems. (authors)

  20. Uranium enrichment plans and policies

    International Nuclear Information System (INIS)

    Schwennesen, J.L.

    1981-01-01

    Significant progress has been made in US efforts to expand its enrichment capacity. The Cascade Improvement Program (CIP) and Cascade Upgrading Program (CUP) are now complete at Oak Ridge and Paducah and almost complete at Portsmouth. Considerable progress has also been made in constructing the Gas Centrifuge Enrichment Plant (GCEP), and physical construction of the first process building is well under way. Current plans are to have two process buildings on-line by 1989 with the remaining six buildings to be added sequentially as needed to meet demand. The status of DOE enrichment services contracts is essentially unchanged from that reported at last year's seminar. The OUEA latest forecast of nuclear power growth, however, is considerably lower than reported last year, although a leveling trend is becoming apparent. The Variable Tails Assay Option (VTAO) of the AFC contract was made available for the third time for FY 1983. The DOE inventories of natural uranium still remain high. The Department of Energy will dispose of this material by using it for Government programs and for enrichment plant operations. It appears that Government inventories of uranium are adequate through at least the mid-1990s. It remains DOE policy not to dispose of its natural uranium stocks through direct sales in the marketplace, except for very small quantities or if an emergency situation would exist and all reasonable attempts had been made, without success, to obtain natural uranium from commercial sources. Finally, with regard to DOE plans on future transaction tails assays, it still appears likely that the current 0.20 percent uranium-235 reference tails assay will be maintained until well into the 1990s, at which time it might be increased up to 0.25 percent uranium-235

  1. 31 CFR 540.316 - Uranium enrichment.

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Uranium enrichment. 540.316 Section 540.316 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE OF... REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

  2. A PHWR with slightly enriched uranium about the first core

    International Nuclear Information System (INIS)

    Notari, C.

    1997-01-01

    Many different studies have been performed in Argentina regarding the use of slightly enriched uranium in the PHWR nuclear plants. These referred mainly to operating plants so that a transition had to be considered from the present natural uranium fuel cycle to the slightly enriched one. In this analysis, technical and economical arguments are presented which favor the use of a natural uranium initial core. The levelized fuel costs are shown to be practically insensitive to the first core and a fast transition is more influential than an initially enriched core. (author)

  3. How is uranium supply affecting enrichment?

    International Nuclear Information System (INIS)

    Steve Kidd

    2007-01-01

    As a result of the enlivened uranium market, momentum has in turn picked up in the enrichment sector. What are the consequences of higher uranium prices? There is, of course, a link between uranium and enrichment supply to the extent that they are at least partial substitutes. On the enrichment supply side, the most obvious feature is the gradual replacement of the old gas diffusion facilities of Usec in the USA and EURODIF in France with more modern and economical centrifuge plants. Assuming Usec can overcome the financing and technical issues surrounding its plans, the last gas diffusion capacity should disappear around 2015 and the entire enrichment market should then be using centrifuges. On the commercial side, the key anticipated developments are mostly in Russia. Although there should still continue to be substantial quantities of surplus Russian HEU available for down blending in the period beyond 2013, it is now reasonable to expect that it will be mostly consumed by internal needs, to fuel Russian-origin reactors both at home and in export markets such as China and India. Finally, as a key sensitive area for the non-proliferation of nuclear weapons, the enrichment sector is likely to be a central point of the new international arrangements which must be developed to support a buoyant nuclear sector throughout this century.

  4. Prospects and problems of uranium enrichment

    International Nuclear Information System (INIS)

    Imai, Ryukichi

    1974-01-01

    The problem of uranium enrichment now concerns principally peaceful nuclear power generation. With the current oil crisis, energy resources assume unprecedented importance. However, the requirements for enriched uranium vary with the vicissitude of the world situation in nuclear power generation; the enterprise of uranium enrichment is related to economic aspect. The following matters are described: dimension of enrichment problem, political factors, changes in requirements, projects in each country, and strategy of enrichment in Japan. (Mori, K.)

  5. Uranium enrichment. Industrial and commercial aspect

    International Nuclear Information System (INIS)

    Lamorlette, G.

    1983-01-01

    The uranium enrichment, a key stage in the fuel cycle of light-water nuclear power stations, applies sophisticated and protected techniques in installations on a very large scale. This article shows how there was a sudden change from a monopoly position in production to a severe competition in a market which is depressed today but offers good prospects for the future. It indicates how the enrichment industrialist have adapted themselves to the fluctuations of the demand, while safeguarding the reliability of the rendered service and the necessary security of supplies for the proper development of the nuclear electric power [fr

  6. 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

  7. Criticality analysis in uranium enrichment plant

    International Nuclear Information System (INIS)

    Okamoto, Tsuyoshi; Kiyose, Ryohei

    1977-01-01

    In a large scale uranium enrichment plant, uranium inventory in cascade rooms is not very large in quantity, but the facilities dealing with the largest quantity of uranium in that process are the UF 6 gas supply system and the blending system for controlling the product concentration. When UF 6 spills out of these systems, the enriched uranium is accumulated, and the danger of criticality accident is feared. If a NaF trap is placed at the forestage of waste gas treatment system, plenty of UF 6 and HF are adsorbed together in the NaF trap. Thus, here is the necessity of checking the safety against criticality. Various assumptions were made to perform the computation surveying the criticality of the system composed of UF 6 and HF adsorbed on NaF traps with WIMS code (transport analysis). The minimum critical radius resulted in about 53 cm in case of 3.5% enriched fuel for light water reactors. The optimum volume ratio of fissile material in the double salt UF 6 .2NaF and NaF.HF is about 40 vol. %. While, criticality survey computation was also made for the annular NaF trap having the central cooling tube, and it was found that the effect of cooling tube radius did not decrease the multiplication factor up to the cooling tube radius of about 5 cm. (Wakatsuki, Y.)

  8. Supply of enriched uranium for research reactors

    International Nuclear Information System (INIS)

    Mueller, H.

    2004-01-01

    Since the RERTR-Meeting in Newport/USA in 1990 recommended in several papers to the research reactor community to agree upon a worldwide unified technical specification for low enriched uranium (LEU) and high enriched uranium (HEU) in order to facilitate supplies of LEU and HEU to fabricators for acceptance and for fabrication of fresh fuel elements. This target for unified and simplified specification has only been partially reached due to different interests of the fabricators because they want to receive the uranium as pure as possible. As a result of various investigations, however, it became clear that both LEU and HEU received from the United States since the late fifties had different qualities which we have to deal with today due to the availability of stocks. We are now one step forward to know more precisely the properties of LEU and HEU we have received in the past. This uranium was never virgin and we have to cope with this situation. Therefore in my present paper I have concentrated on the documentation of analytical work performed on samples of LEU and HEU received in the past. I propose furthermore a frame of unified specifications for so-called virgin LEU and HEU including uranium from a Zero-experiment. In addition I am giving a recommendation for specifications of LEU obtained by blending of reprocessed HEU. Finally I am touching the question of secure supplies of fresh LEU. (author)

  9. Evaluation of the uranium enrichment demonstration plant project

    International Nuclear Information System (INIS)

    Sugitsue, Noritake

    2001-01-01

    In this report, the organization system of the uranium enrichment business is evaluated, based on the operation of the uranium enrichment demonstration plant. As a result, in uranium enrichment technology development or business, it was acknowledged that maintenance of the organization which has the Trinity of a research/engineering/operation was necessary in an industrialization stage by exceptional R and D cycle. Japan Nuclear Fuel Ltd. (JNFL) set up the Rokkashomura Aomori Uranium Enrichment Research and Development Center in November 2000. As a result, the system that company directly engaged in engineering development was prepared. And results obtained in this place is expected toward certain establishment of the uranium enrichment business of Japan. (author)

  10. Study on the use of slightly enriched uranium fuel cycle in an existing CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Choong Sub; Kim, Hyun Dae [Institute for Advanced Engineering, Seoul (Korea, Republic of)

    1997-12-31

    To test the viability of CANFLEX-SEU bundles in an existing CANDU 6 reactor, core follow-up simulation has been carried out using the reactor fueling simulation program of the CANDU 6, RFSP computer code, and a lattice physics code, WIMS-AECL. During the core follow-up, bundle and channel powers and zone levels have been checked against their operating limits at each simulation. It is observed from the simulation results that an equilibrium core loaded with 0.9 w/o CANFLEX-SEU bundles could be refueled and maintained for 550 FPD without any significant violations in the channel and bundle power limits and the permissible operating range of the liquid zone controllers. 8 refs., 2 figs., 1 tab. (Author)

  11. Study on the use of slightly enriched uranium fuel cycle in an existing CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Choong Sub; Kim, Hyun Dae [Institute for Advanced Engineering, Seoul (Korea, Republic of)

    1998-12-31

    To test the viability of CANFLEX-SEU bundles in an existing CANDU 6 reactor, core follow-up simulation has been carried out using the reactor fueling simulation program of the CANDU 6, RFSP computer code, and a lattice physics code, WIMS-AECL. During the core follow-up, bundle and channel powers and zone levels have been checked against their operating limits at each simulation. It is observed from the simulation results that an equilibrium core loaded with 0.9 w/o CANFLEX-SEU bundles could be refueled and maintained for 550 FPD without any significant violations in the channel and bundle power limits and the permissible operating range of the liquid zone controllers. 8 refs., 2 figs., 1 tab. (Author)

  12. Uranium enrichment. 1980 annual report

    International Nuclear Information System (INIS)

    1981-05-01

    This report contains data and related information on the production of enriched uranium at the gaseous diffusion plants and an update on the construction and project control center for the gas centrifuge plant. Power usage at the gaseous diffusion plants is illustrated. The report contains several glossy color pictures of the plants and processes described. In addition to gaseous diffusion and the centrifuge process, three advanced isotope separation process are now being developed. The business operation of the enrichment plants is described; charts on revenue, balance sheets, and income statements are included

  13. 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.

  14. 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

  15. 76 FR 72984 - Revised Application for a License To Export High-Enriched Uranium

    Science.gov (United States)

    2011-11-28

    ... NUCLEAR REGULATORY COMMISSION Revised Application for a License To Export High-Enriched Uranium The application for a license to export high-enriched Uranium has been revised as noted below. Notice... fabricate fuel France. Security Complex; October 18, Uranium (93.35%). uranium (174.0 elements in France...

  16. PC based uranium enrichment analyser

    International Nuclear Information System (INIS)

    Madan, V.K.; Gopalakrishana, K.R.; Bairi, B.R.

    1991-01-01

    It is important to measure enrichment of unirradiated nuclear fuel elements during production as a quality control measure. An IBM PC based system has recently been tested for enrichment measurements for Nuclear Fuel Complex (NFC), Hyderabad. As required by NFC, the system has ease of calibration. It is easy to switch the system from measuring enrichment of fuel elements to pellets and also automatically store the data and the results. The system uses an IBM PC plug in card to acquire data. The card incorporates programmable interval timers (8253-5). The counter/timer devices are executed by I/O mapped I/O's. A novel algorithm has been incorporated to make the system more reliable. The application software has been written in BASIC. (author). 9 refs., 1 fig

  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. Comments on applications of reduced enrichment fuels

    International Nuclear Information System (INIS)

    Winkler, M.H.

    1983-01-01

    Full text: I will briefly describe the experience gained using different fuels in the SAPHIR reactor in Switzerland. The SAPHIR has been operating since 1957 and was the first swimming pool reactor built outside of the United States, which was originally known as the Geneva Conference Reactor. The first core was loaded with 20 percent enriched high density UO 2 fuel with a density of about 2.5 grams per cc, fabricated in 1955 by Oak Ridge National Laboratory. After a few years of operation at a power level of one MW, more than one batch of the elements released small amounts of fission products mainly Xe and Kr. When these releases were discovered, high enriched fuel was becoming available so that the fuel fabricators began to produce the lower density high enriched fuels. During this transition from fabrication of low to high enriched fuels no one could foresee that the stone age of nuclear fuel fabrication would come back again. Therefore, we did not investigate the reasons for the fission product release from the high density low enriched UO 2 fuel. The second fuel type used in the SAPHIR was the 90 percent enriched low density U 3 O 8 fuel fabricated by NUKEM. This high enriched fuel has performed satisfactorily over the years. Since 1968, the core has been using improved 23 plate fuel elements with a loading of 280 grams of uranium. The reactor power has been recently increased to five MW. An additional increase in the power level to 10 MW is planned at the end of next year so that heavier loaded elements will be needed. In order to follow the recommendations of the INFCE working group 8C and in cooperation with the reduced enrichment program, we intend to initially reduce the fuel enrichment to 45 percent. Last year we ordered five fuel elements with a loading of 320 grams 235 U/element and 45 percent enrichment for full power tests. Unfortunately, the delivery of the necessary enriched fuel uranium has been delayed and it is not available at this time. If

  19. Uranium enrichment in the United States

    International Nuclear Information System (INIS)

    Hill, J.H.; Parks, J.W.

    1975-01-01

    History, improvement programs, status of electrical power availability, demands for uranium enrichment, operating plan for the U. S. enriching facilities, working inventory of enriched uranium, possible factors affecting deviations in the operating plan, status of gaseous diffusion technology, status of U. S. gas centrifuge advances, transfer of enrichment technology, gaseous diffusion--gas centrifuge comparison, new enrichment capacity, U. S. separative work pricing, and investment in nuclear energy are discussed. (LK)

  20. URENCO. Uranium enrichment with advanced technology

    International Nuclear Information System (INIS)

    2011-01-01

    URENCO Deutschland is a subsidiary of URENCO Enrichment Company Limited, an international enterprise founded in 1970 in the State Treaty of Almelo, which offers uranium enrichment for nuclear power plants all over the world with the use of advanced technology. URENCO facilities at present are operated in the United Kingdom, the Netherlands, USA, and in Germany. The German URENCO location is Gronau, Westphalia, where cascades have been in operation since 1985 using centrifuge technology to enrich nuclear fuel to up to 5% uranium-235. The URENCO Group supplies nuclear power plants in Europe and overseas countries with a world market share, at present, of more than 25% with a rising tendency. The first uranium separation plant in Gronau (UTA-1) attained its full separation performance of 1,800 t USW/a in late 2005. In February 2005, construction and operation of another plant had been licensed, which can raise the aggregate capacity on site to 4,500 t USW per annum. Construction of the new plant (UTA-2) was begun in summer 2005. UTA-2 will use the latest, most powerful URENCO centrifuge. URENCO has more than 3,500 visitors a year at its German location alone, thus demonstrating its pro-active information policy and offering to the public a maximum of opportunities to acquire information by attending presentations and tours of the plant. (orig.)

  1. R and D on laser uranium enrichment

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    An AEC Advisory Committee on Uranium Enrichment has completed investigations into the actual condition of laser isotope separation. The working group set up for the purpose has issued a report on the series of investigations made on its development and measures for promoting it. The report says that the development of the process in Japan is at a fundamental stage. Noting that further efforts are needed before its future can be predicted, the report proposes a cource of research and development for the immediate future. For the atomic vapor laser isotope separation (AVLIS), government organizations are engaged in data base buildup and conducting basis engineering tests, and Japan Atomic Energy Research Institute will consider the re-enrichment of uranium recovered from reprocessing. Non-governmental unions of researchers will promote the combination of copper-vapor laser and dye laser. For the molecular laser isotope separation (MLIS), the Institute of Physical and Chemical Research will take up studies with the cooperation of the Power Reactor and Nuclear Fuel Development Corporation. In chapters covering the philosophy of laser uranium enrichment technology development, the report deals with its significance, actual conditions and tasks, and goals and measures for its promotion. (Nogami, K.)

  2. Profile of World Uranium Enrichment Programs-2009

    International Nuclear Information System (INIS)

    Laughter, Mark D.

    2009-01-01

    It is generally agreed that the most difficult step in building a nuclear weapon is acquiring fissile material, either plutonium or highly enriched uranium (HEU). Plutonium is produced in a nuclear reactor, whereas HEU is produced using a uranium enrichment process. Enrichment is also an important step in the civil nuclear fuel cycle, in producing low enriched uranium (LEU) for use as fuel for nuclear reactors to generate electricity. However, the same equipment used to produce LEU for nuclear reactor fuel can also be used to produce HEU for weapons. Safeguards at an enrichment plant are the array of assurances and verification techniques that ensure uranium is not diverted or enriched to HEU. There are several techniques for enriching uranium. The two most prevalent are gaseous diffusion, which uses older technology and requires a lot of energy, and gas centrifuge separation, which uses more advanced technology and is more energy efficient. Gaseous diffusion plants (GDPs) provide about 40% of current world enrichment capacity but are being phased out as newer gas centrifuge enrichment plants (GCEPs) are constructed. Estimates of current and future enrichment capacity are always approximate, due to the constant upgrades, expansions, and shutdowns occurring at enrichment plants, largely determined by economic interests. Currently, the world enrichment capacity is approximately 56 million kilogram separative work units (SWU) per year, with 22.5 million in gaseous diffusion and more than 33 million in gas centrifuge plants. Another 34 million SWU/year of capacity is under construction or planned for the near future, almost entirely using gas centrifuge separation. Other less-efficient techniques have also been used in the past, including electromagnetic and aerodynamic separations, but these are considered obsolete, at least from a commercial perspective. Laser isotope separation shows promise as a possible enrichment technique of the future but has yet to be

  3. Safety criteria of uranium enrichment plants

    International Nuclear Information System (INIS)

    Nardocci, A.C.; Oliveira Neto, J.M. de

    1994-01-01

    The applicability of nuclear reactor safety criteria applied to uranium enrichment plants is discussed, and a new criterion based on the soluble uranium compounds and hexafluoride chemical toxicities is presented. (L.C.J.A.). 21 refs, 4 tabs

  4. Gasket for uranium enrichment plant

    Energy Technology Data Exchange (ETDEWEB)

    Kishi, S; Aiyoshi, H

    1977-02-08

    A gasket to be inserted between flange joints in the equipments and pipe lines of an uranium enrichment plant having neither permeability nor adsorptivity to water while maintaining mechanical, physical and chemical properties of an elastomer gasket is described. A gasket made of an elastomeric material such as a polymer is integratedly formed at its surface with anti-slip projections. The gasket is further surrounded at its upper and lower peripheral sides, as well as outer circumferential portion with a U-sectioned cover (enclosure) made of fluoro-plastics. In this arrangement, the gasket main body shows a gas-tightness for uranium hexafluoride gas and the cover exhibits a gas-tightness for other component gases such as moisture to thereby prevent degradation of the gasket due to absorption and permeation of the moisture.

  5. Uranium enrichment management review. Final report

    International Nuclear Information System (INIS)

    Ellett, J.D.; Rieke, W.B.; Simpson, J.W.; Sullivan, P.E.

    1980-01-01

    The uranium enrichment enterprise of the US Department of Energy (DOE) provides enriched nuclear fuel for private and government utilities domestically and abroad. The enterprise, in effect, provides a commercial service and represents a signficant business operation within the US government: more than $1 billion in revenues annually and future capital expenditures estimated at several billions of dollars. As a result, in May 1980, the Assistant Secretary for Resource Applications within DOE requested that a group of experienced business executives be assembled to review the operation, financing, and management of the uranium enrichment enterprise as a basis for advising the Secretary of Energy. The review group was specifically asked to focus on the management activities to which sound business practices could be applied. The group developed findings and recommendations in six areas: management of operations and construction; long-range planning; marketing of enrichment services; financial management; research and development; and general management. The chapters of this report present first the management review group's recommendations in the six areas evaluated and then the findings and issues in each area. An appendix provides the group's calendar of meetings. A list of major reference sources used in the course of the study is also included. 12 references

  6. Surplus Highly Enriched Uranium Disposition Program plan

    International Nuclear Information System (INIS)

    1996-10-01

    The purpose of this document is to provide upper level guidance for the program that will downblend surplus highly enriched uranium for use as commercial nuclear reactor fuel or low-level radioactive waste. The intent of this document is to outline the overall mission and program objectives. The document is also intended to provide a general basis for integration of disposition efforts among all applicable sites. This plan provides background information, establishes the scope of disposition activities, provides an approach to the mission and objectives, identifies programmatic assumptions, defines major roles, provides summary level schedules and milestones, and addresses budget requirements

  7. Development of empirical relation for isotope of uranium in enriched uranium matrix

    International Nuclear Information System (INIS)

    Srivastava, S.K.; Vidyasagar, D.; Jha, S.K.; Tripathi, R.M.

    2018-01-01

    Uranium enriched in 235 U is required in commercial light water reactors to produce a controlled nuclear reaction. Enrichment allows the 235 U isotopes to be increased from 0.71% to a range between 2% to 5% depending upon requirement. The enriched uranium in the form of sintered UO 2 pellet is used for any commercially operating boiling light water reactors. The enriched uranium fuel bundle surface swipes sample is being analysed to assess the tramp uranium as a quality control parameter. It is known that the 234 U isotope also enriched along with 235 U isotope in conventional gaseous diffusion enrichment process. The information about enrichment percentage of 234 U helps to characterize isotopic properties of enriched uranium. A few reports provide the empirical equation and graphs for finding out the specific activity, activity percentage, activity ratio of 234 U isotopes for enriched uranium. Most of them have not provided the reference for the data used and their source. An attempt has been made to model the relationship between 234 U and 235 U as a function of uranium enrichment at low level

  8. Natural uranium utilization without enrichment and reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Sekimoto, H.; Toshinsky, V.; Ryu, K. [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

    2001-07-01

    Two types of fast reactor are investigated to utilize the natural uranium without enrichment and reprocessing in an equilibrium state. The first trial is SFPR. Its fuel-shuffling pattern is optimized. An obtained result gives its peak fuel burnup of 22,5%, power peaking factor of 1.5 and peak excess reactivity of 2,15%. The second trial is CANDLE burnup scheme, where distribution shapes of neutron flux and nuclide densities are constant but move in axial direction with a constant velocity. A feasible solution gives the speed of burning region of 4,1 cm/year, k{sub eff} of 1,02 and average spent fuel burnup of 41%. (author)

  9. 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.)

  10. 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

  11. Criticality of mixtures of plutonium and high enriched uranium

    International Nuclear Information System (INIS)

    Grolleau, E.; Lein, M.; Leka, G.; Maidou, B.; Klenov, P.

    2003-01-01

    This paper presents a criticality evaluation of moderated homogeneous plutonium-uranium mixtures. The fissile media studied are homogeneous mixtures of plutonium and high enriched uranium in two chemical forms: aqueous mixtures of metal and mixtures of nitrate solutions. The enrichment of uranium considered are 93.2wt.% 235 U and 100wt.% 235 U. The 240 Pu content in plutonium varies from 0wt.% 240 Pu to 12wt.% 240 Pu. The critical parameters (radii and masses of a 20 cm water reflected sphere) are calculated with the French criticality safety package CRISTAL V0. The comparison of the calculated critical parameters as a function of the moderator-to-fuel atomic ratio shows significant ranges in which high enriched uranium systems, as well as plutonium-uranium mixtures, are more reactive than plutonium systems. (author)

  12. 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

  13. 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

  14. Study on the radiotoxicology of enriched uranium

    International Nuclear Information System (INIS)

    Zhu Shoupeng; Zheng Siying; Wang Guolin; Wang Chongdao; Cao Genfa

    1987-12-01

    A study on the retentive peculiarity of soluble enriched uranium UO 2 F 2 were observed after iv once or consecutive ip qd x 3d to Wistar male rats. The dynamic retention of radioactivity in the body showed that the enriched uranium UO 2 F 2 was chiefly localized in kidney, and then in skeleton and liver. The radioactivity of the enriched uranium UO 2 F 2 in skeleton rose steadily while the concentratoin in kidney and liver droped. When enriched uranium UO 2 F 2 was accumulated in organism, it caused chromosome aberrations on bone marrow cells. Results indicated that the chromosome aberration rates were elevated when the dose of the enriched uranium UO 2 F 2 was increased, at the same time, the cell division was depressed. Accumulation of insoluble enriched uranium U 3 O 8 in gastrointestinal tract was well described by a two exponential expression. Values of retention estimate for fast component, T 1 = 0.34 d, and for relatively long term component, T 2 = 4.05 d. The deposition of UO 2 F 2 in the intact skin was only 0.16 to 0.18% of the total contaminated UO 2 F 2 . Penetration of the enriched uranium UO 2 F 2 was dominantly increased in abraded skin. This value is about 25 to 32 times as compaired with that in intact skin. Retention of the enriched uranium UO 2 F 2 through abraded skins was dominantly localized in kidney and skeleton

  15. Perspectives for the uranium enrichment in Brazil

    International Nuclear Information System (INIS)

    Senna, J.G.S.M.

    1991-01-01

    Through an analysis of the electrical energy future in Brazil, the needs for enriched uranium are discussed, and therefore the importance of developing local capability for self-production. A description of the production processes that are well established is given first, then the analysis itself is performed and finally a visualization of the International Market for enriched uranium is shown. (author)

  16. Report of Sectional Committee on Industrialization of Uranium Enrichment

    International Nuclear Information System (INIS)

    1981-01-01

    In order to accelerate the development and utilization of atomic energy which is the core of the substitute energies for petroleum, it is indispensable requirement to establish independent fuel cycle as the base. In particular, the domestic production of enriched uranium is necessary to eliminate the obstacles to secure the energy supply in Japan. The construction and operation of the pilot plant for uranium enrichment by centrifugal separation method have progressed smoothly, and the technical base for the domestic production of enriched uranium is being consolidated. For the time being, the service of uranium enrichment is given by USA and France, but it is expected that the short supply will arise around 1990. The start of operation of the uranium enrichment plant in Japan is scheduled around 1990, and the scale of the plant will be expanded stepwise thereafter. The scale of production is assumed as 3000 t SWU/year in 2000. Prior to this commercial plant, the prototype plant of up to 250 t SWU/year capacity will be operated in 1986, starting the production of centrifugal separators in 1983. The production line for centrifugal separators will have the capacity of up to 125 t SWU/year. The organization for operating these plants, the home production of natural uranium conversion, the uranium enrichment by chemical method and others are described. (Kako, I.)

  17. Use of highly enriched uranium at the FRM-II

    Energy Technology Data Exchange (ETDEWEB)

    Boening, K. [Forschungs-Neutronenquelle FRM-II, Technische Universitaet Muenchen, D-85747 Garching bei Muenchen (Germany)

    2002-07-01

    The new FRM-II research reactor in Munich, Germany, provides a high flux of thermal neutrons outside of the core at only 20 MW power. This is achieved by using a single compact, cylindrical fuel element with highly enriched uranium (HEU) which is cooled by light water and placed in the center of a large heavy water tank. The paper outlines the arguments which have led to this core concept and summarizes its performance. It also reports on alternative studies which have been performed for the case of low enriched uranium (LEU) and compares the data of the two concepts, with the conclusion that the FRM-II cannot be converted to LEU. A concept using medium enriched uranium (MEU) is described as well as plans to develop such a fuel element in the future. Finally, it is argued that the use of HEU fuel elements at the FRM-II does not - realistically -involve any risk of proliferation. (author)

  18. Uranium-enriched granites in Sweden

    International Nuclear Information System (INIS)

    Wilson, M.R.; Aakerblom, G.

    1980-01-01

    Granites with uranium contents higher than normal occur in a variety of geological settings in the Swedish Precambrian, and represent a variety of granite types and ages. They may have been generated by the anatexis of continental crust or processes occurring at a much greater depth. They commonly show enrichment in F, Sn, W and/or Mo. Only in one case is an important uranium mineralization thought to be directly related to a uranium-enriched granite, while the majority of epigenetic uranium mineralizations with economic potential are related to hydrothermal processes in areas where the bedrock is regionally uranium-enhanced. (author)

  19. Uranium enrichment in South Africa

    International Nuclear Information System (INIS)

    Roux, A.J.A.; Grant, W.L.

    1976-01-01

    It is stated that the South African process is of an aerodynamic type, the separating element being in effect a high performance stationary-walled centrifuge using UF 6 in hydrogen as process fluid. Some details of the very low uranium inventory and high separation factor achievable are given. A new cascade technique is described, based on the principle that an axial flow compressor can simultaneously transmit several streams of different isotopic composition without there being significant mixing between them. The research and development programme is discussed. It is expected that an enrichment plant of 5000 t/a SW capacity, with provision for expansion up to 10,000 t/a SW capacity, will come into operation by 1984. (U.K.)

  20. 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.

  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. 75 FR 15743 - Application for a License To Export High-Enriched Uranium

    Science.gov (United States)

    2010-03-30

    ... NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant to 10 CFR 110.70(c) ``Public notice of receipt of an application,'' please take notice that the...-Enriched 160.0 kilograms To fabricate fuel France. Complex, March 3, 2010. Uranium (93.35%). uranium (149...

  3. Thermal breeder fuel enrichment zoning

    International Nuclear Information System (INIS)

    Capossela, H.J.; Dwyer, J.R.; Luce, R.G.; McCoy, D.F.; Merriman, F.C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect. 1 figure

  4. EURODIF: the uranium enrichment by gaseous diffusion

    International Nuclear Information System (INIS)

    Rougeau, J.P.

    1981-01-01

    During the seventies the nuclear power programme had an extremely rapid growth rate which entailed to increase the world uranium enrichment capacity. EURODIF is the largest undertaking in this field. This multinational joint venture built and now operates and enrichment plant using the gaseous diffusion process at Tricastin (France). This plant is delivering low enriched uranium since two years and has contracted about 110 million SWU's till 1990. Description, current activity and prospects are given in the paper. (Author) [pt

  5. Supply of enriched uranium for research reactors

    International Nuclear Information System (INIS)

    Mueller, Hans; Laucht, Juergen

    1996-01-01

    Since the RERTR meeting in 1990 at Newport/USA, NUKEM recommended that the research reactor community agree upon a worldwide unified technical specification for low enriched uranium (LEU) and high enriched uranium (HEU) since there existed numerous specifications both from suppliers/fabricators and research reactors. The target recommended by NUKEM is to arrive at a worldwide unified standard specification in order to facilitate supplies of LEU and HEU to fabricators for fabrication of research reactor fuel elements. In our paper presented at the RERTR meeting at Paris in September 1995, we pointed out that LEU and HEU supplied by the U.S. Department of Energy (DOE) in the past was never 'virgin' material, i.e., it was mixed with reprocessed uranium. Our recommendation was to include this fact in the proposed unified specification. Since the RERTR meeting in 1995 progress on a unified standard specification has been made and we would like to provide more specific information about that in this paper. Furthermore, we will deal with the question whether there is a secure supply of LEU for converted research reactors. We list current and potential suppliers of LEU, noting however, that the DOE has for a number of years been unable to supply any LEU due to production problems. The future availability of LEU of U.S. origin is, of course, essential for those research reactor operators which have converted their reactors from HEU to LEU and which are intending to return spent fuel of U.S. origin to the U.S.A. (author)

  6. Reducing enrichment of fuel for research reactors

    International Nuclear Information System (INIS)

    Kanda, Keiji; Matsuura, Shojiro.

    1980-01-01

    In research reactors, highly enriched uranium (HEU) is used as fuel for their purposes of operation. However, the United States strongly required in 1977 that these HEU should be replaced by low enrichment uranium (LEU) of 20% or less, or even in unavoidable cases, it should be replaced by medium enrichment uranium (MEU). INFCE (International Nuclear Fuel Cycle Evaluation) which started its activity just at that time decided to discuss this problem in the research reactor group of No. 8 sectional committee. Japan has been able to forward the work, taking a leading part in the international opinion because she has taken the countermeasures quickly. INFCE investigated the problem along the lines of policy that the possibility of reducing the degree of enrichment should be limited to the degree in which the core structures and equipments of research reactors will be modified as little as possible, and the change of fuel element geometry will be done within the permissible thermohydrodynamic capacity, and concluded that it might be possible in near future to reduce the degree of enrichment to about 45% MEU, while the reduction to 20% LEU might require considerable research, development and verification. On the other hand, the joint researches by Kyoto University and ANL (Argonne National Laboratory) and by Japan Atomic Energy Research Institute and ANL are being continued. IAEA has edited the guidebook (IAEA-TECDOC-233) for reducing the degree of enrichment for developing countries. (Wakatsuki, Y.)

  7. 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.

  8. 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

  9. 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.

  10. Contemporary and prospective fuel cycles for WWER-440 based on new assemblies with higher uranium capacity and higher average fuel enrichment

    International Nuclear Information System (INIS)

    Gagarinskiy, A.A.; Saprykin, V.V.

    2009-01-01

    RRC 'Kurchatov Institute' has performed an extensive cycle of calculations intended to validate the opportunities of improving different fuel cycles for WWER-440 reactors. Works were performed to upgrade and improve WWER-440 fuel cycles on the basis of second-generation fuel assemblies allowing core thermal power to be uprated to 107 108 % of its nominal value (1375 MW), while maintaining the same fuel operation lifetime. Currently intensive work is underway to develop fuel cycles based on second-generation assemblies with higher fuel capacity and average fuel enrichment per assembly increased up to 4.87 % of U-235. Fuel capacity of second-generation assemblies was increased by means of eliminated central apertures of fuel pellets, and pellet diameter extended due to reduced fuel cladding thickness. This paper intends to summarize the results of works performed in the field of WWER-440 fuel cycle modernization, and to present yet unemployed opportunities and prospects of further improvement of WWER-440 neutronic and operating parameters by means of additional optimization of fuel assembly designs and fuel element arrangements applied. (Authors)

  11. 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

  12. Current perspective of the uranium enrichment market

    International Nuclear Information System (INIS)

    Laughon, K.O.

    1986-01-01

    Over the past several years, developments in the uranium enrichment market have required the Department of Energy (DOE) to make a number of changes in the U.S. enrichment enterprise. These changes have been made to allow DOE to conduct our enrichment business so as to be more responsive to changing market forces. Needless to say, some of these changes have been difficult, but they have been necessary if they are to conduct a healthy and competitive uranium enrichment business in the United States. This paper discusses several topics, including: The Uranium Enrichment Market, Utility Services (US) Contracts, Reduced Prices, Incentive Pricing, Better Customer Services, and Advanced Technology. In addition to these topics, information is provided on the recent court action regarding the US Contracts and the viability finding on the uranium mining industry

  13. Calculated NWIS signatures for enriched uranium metal

    International Nuclear Information System (INIS)

    Valentine, T.E.; Mihalczo, J.T.; Koehler, P.E.

    1995-01-01

    Nuclear Weapons Identification System (NWIS) signatures have been calculated using a Monte Carlo transport code for measurement configurations of a 252 Cf source, detectors, and a uranium metal casting. NWIS signatures consist of a wide variety of time-and frequency-analysis signatures such as the time distribution of neutrons after californium fission, the time distribution of counts in a detector after a previous count, the number of times n pulses occur in a time interval, and various frequency-analysis signatures, such as auto-power and cross-power spectral densities, coherences, and a ratio of spectral densities. This ratio is independent of detection efficiency. The analysis presented here, using the MCNP-DSP code, evaluates the applicability of this method for measurement of the 235 U content of 19-kg castings of depleted uranium and uranium with enrichments of 20, 40, 60, 80, 90, and 93.2 wt % 235 U. The dependence of the wide variety of NWIS signatures on 235 U content and possible configurations of a measurement system are presented. These preliminary calculations indicate short measurement times. Additional calculations are being performed to optimize the source-detector-moderator-casting configuration for the shortest measurement time. Although the NWIS method was developed for nuclear weapons identification, the development of a small processor now allows it to be also applied in a practical way to subcriticality measurements, nuclear fuel process monitoring and qualitative nondestructive assay of special nuclear material

  14. Distribution of equilibrium burnup for an homogeneous core with fuel elements of slightly enriched uranium (0.85% U-235) at Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Sidelnik, J.I.; Perez, R.A.; Salom, G.F.

    1987-01-01

    At Atucha I, the present fuel management with natural uranium comprises three burnup areas and one irradiation path, sometimes performing four steps in the reactor core, according to the requirements. The discharge burnup is 6.0 Mw d/kg U for a waste reactivity of 6.5 m k and a heavy water purity of 99.75%. This is a preliminary study to obtain the distribution of equilibrium burnup of an homogeneous core with slightly enriched uranium (0.85% by weight U-235), using the time-averaged method implemented in the code PUMA and a representative model of one third of core and fixed rod position. It was found a strategy of three areas and two paths that agrees with the present limits of channel power and specific power in fuel rod. The discharge burnup obtained is 11.6 Mw d/kg U. This strategy is calculated with the same method and a full core representation model is used to verify the obtained results. (Author)

  15. The uranium enrichment industry and the SILEX process

    International Nuclear Information System (INIS)

    Goldsworthy, M.

    1999-01-01

    Silex Systems Limited has been developing a new laser isotope separation process since 1992. The principle application of the SILEX Technology is Uranium Enrichment, the key step in the production of fuel for nuclear power plants. The Uranium Enrichment industry, today worth ∼ US$3.5 Billion p.a., is dominated by four major players, the largest being USEC with almost 40% of the market. In 1996, an agreement was signed between Silex and USEC to develop SILEX Technology for potential application to Uranium Enrichment. The SILEX process is a low cost, energy efficient scheme which may provide significant commercial advantage over current technology and competing laser processes. Silex is also investigating possible application to the enrichment of Silicon, Carbon and other materials. Significant markets may develop for such materials, particularly in the semiconductor industry

  16. PWR fuel of high enrichment with erbia and enriched gadolinia

    International Nuclear Information System (INIS)

    Bejmer, Klaes-Håkan; Malm, Christian

    2011-01-01

    Today standard PWR fuel is licensed for operation up to 65-70 MWd/kgU, which in most cases corresponds to an enrichment of more than 5 w/o "2"3"5U. Due to criticality safety reason of storage and transportation, only fuel up to 5 w/o "2"3"5U enrichment is so far used. New fuel storage installations and transportation casks are necessary investments before the reactivity level of the fresh fuel can be significantly increased. These investments and corresponding licensing work takes time, and in the meantime a solution that requires burnable poisons in all pellets of the fresh high-enriched fuel might be used. By using very small amounts of burnable absorber in every pellet the initial reactivity can be reduced to today's levels. This study presents core calculations with fuel assemblies enriched to almost 6 w/o "2"3"5U mixed with a small amount of erbia. Some of the assemblies also contain gadolinia. The results are compared to a reference case containing assemblies with 4.95 w/o "2"3"5U without erbia, utilizing only gadolinia as burnable poison. The comparison shows that the number of fresh fuel assemblies can be reduced by 21% (which increases the batch burnup by 24%) by utilizing the erbia fuel concept. However, increased cost of uranium due to higher enrichment is not fully compensated for by the cost gain due to the reduction of the number assemblies. Hence, the fuel cycle cost becomes slightly higher for the high enrichment erbia case than for the reference case. (author)

  17. Hydrogen-enriched fuels

    Energy Technology Data Exchange (ETDEWEB)

    Roser, R. [NRG Technologies, Inc., Reno, NV (United States)

    1998-08-01

    NRG Technologies, Inc. is attempting to develop hardware and infrastructure that will allow mixtures of hydrogen and conventional fuels to become viable alternatives to conventional fuels alone. This commercialization can be successful if the authors are able to achieve exhaust emission levels of less than 0.03 g/kw-hr NOx and CO; and 0.15 g/kw-hr NMHC at full engine power without the use of exhaust catalysts. The major barriers to achieving these goals are that the lean burn regimes required to meet exhaust emissions goals reduce engine output substantially and tend to exhibit higher-than-normal total hydrocarbon emissions. Also, hydrogen addition to conventional fuels increases fuel cost, and reduces both vehicle range and engine output power. Maintaining low emissions during transient driving cycles has not been demonstrated. A three year test plan has been developed to perform the investigations into the issues described above. During this initial year of funding research has progressed in the following areas: (a) a cost effective single-cylinder research platform was constructed; (b) exhaust gas speciation was performed to characterize the nature of hydrocarbon emissions from hydrogen-enriched natural gas fuels; (c) three H{sub 2}/CH{sub 4} fuel compositions were analyzed using spark timing and equivalence ratio sweeping procedures and finally; (d) a full size pick-up truck platform was converted to run on HCNG fuels. The testing performed in year one of the three year plan represents a baseline from which to assess options for overcoming the stated barriers to success.

  18. 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

  19. Enriched uranium sales: effect on supply industry

    International Nuclear Information System (INIS)

    Andersen, R.K.

    1985-01-01

    The subject is covered in sections: introduction (combined effect of low-enriched uranium (LEU) inventory sales and utility services enrichment contract terms); enrichment market overview; enrichment market dynamics; the reaction of the US Department of Energy; elimination of artificial demand; draw down of inventories; purchase and sale of LEU inventories; tails assay option; unfulfilled requirements for U 3 O 8 ; conclusions. (U.K.)

  20. Uranium enrichment with lasers - will South Africa lead or lag?

    International Nuclear Information System (INIS)

    Du Toit, G.

    1992-01-01

    Over 30 percent of the cost of locally made nuclear fuel in South Africa is associated with increasing the concentration of uranium 235. Cheaper enrichment technologies and, in particular, the decision by the Atomic Energy Corporation of South Africa to concentrate its research efforts on laser techniques are therefore of considerable significance. The laser isotope separation programme in South Africa is reviewed. 1 ill

  1. Nickel container of highly-enriched uranium bodies and sodium

    Science.gov (United States)

    Zinn, Walter H.

    1976-01-01

    A fuel element comprises highly a enriched uranium bodies coated with a nonfissionable, corrosion resistant material. A plurality of these bodies are disposed in layers, with sodium filling the interstices therebetween. The entire assembly is enclosed in a fluid-tight container of nickel.

  2. Nickel container of highly-enriched uranium bodies and sodium

    International Nuclear Information System (INIS)

    Zinn, W.H.

    1976-01-01

    A fuel element comprises highly enriched uranium bodies coated with a nonfissionable, corrosion resistant material. A plurality of these bodies are disposed in layers, with sodium filling the interstices therebetween. The entire assembly is enclosed in a fluid-tight container of nickel

  3. 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

  4. Monte Carlo calculational design of an NDA instrument for the assay of waste products from high enriched uranium spent fuels

    International Nuclear Information System (INIS)

    Eccleston, G.W.; Schrandt, R.G.; MacDonald, J.L.; Cverna, F.H.

    1979-01-01

    The Monte Carlo design of the waste assay region of a dual assay system, to be installed at the Fluorinal and Storage Facility, is described. The instrument will be used by the facility operator to assay high-enriched spent fuel packages and waste solids produced from dissolution of the fuels. The fissile content discharged in the waste is expected to vary between 0 and 400 g of 235 U. Material accountability measurements of the waste must be obtained in the presence of large neutron (0.5 x 10 6 n/s) and gamma (50,000 R/hr) backgrounds. The assay system employs fast-neutron irradiation of the sample, using a 5 mg 252 Cf source, followed by delayed neutron counting after the source is transferred to storage. Calculations indicate a +-4-g (2 sigma) assay for a waste canister containing 300 g of 235 U is achievable with an end-of-life (1 mg) 252 Cf source and a background rate of 0.5 x 10 6 n/s

  5. 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)

  6. Development of on-line uranium enrichment monitor of gaseous UF6 for uranium enrichment plant

    International Nuclear Information System (INIS)

    Lu Xuesheng; Liu Guorong; Jin Huimin; Zhao Yonggang; Li Jinghuai; Hao Xueyuan; Ying Bin; Yu Zhaofei

    2013-01-01

    An on-line enrichment monitor was developed to measure the enrichment of UF 6 , flowing through the processing pipes in uranium enrichment plant. A Nal (Tl) detector was used to measure the count rates of the 185.7 keV γ-ray emitted from 235 U, and the total quantity of uranium was determined from thermodynamic characteristics of gaseous uranium hexafluoride. The results show that the maximum relative standard deviation is less than 1% when the measurement time is 120 s or more and the pressure is more than 2 kPa in the measurement chamber. Uranium enrichment of gaseous uranium hexafluoride in the output end of cascade can be monitored continuously by using the device. It should be effective for nuclear materials accountability verifications and materials balance verification at uranium enrichment plant. (authors)

  7. 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

  8. Uranium enrichment (a strategy analysis overview)

    International Nuclear Information System (INIS)

    Blahnik, C.

    1979-08-01

    An analysis of available information on enrichment technology, separative work supply and demand, and SWU cost is presented. Estimates of present and future enrichment costs are provided for use in strategy analyses of alternate nuclear fuel cycles and systems. (auth)

  9. 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

  10. 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

  11. A feasibility study concerning the conversion of the TR-2 reactor from using highly enriched uranium to light enriched uranium

    International Nuclear Information System (INIS)

    Aldemir, T.; Turgut, H.M.; Bretscher, M.M.; Snelgrove, L.J.

    1983-01-01

    A study has been made of the feasibility of converting the 5-MW TR-2 reactor at CNAEM to use fuel with uranium enrichment of 3 O 8 -Al fuel meat with a uranium density in the range 2.3 to 3.0 g/cm 3 in the fuel meat with meat thickness varying between 0.9 and 1.00 mm, the number of plates in the LEU element being reduced from 23 in the HEU element to 19 to 20 to maintain adequate cooling. Fuels within this density range are expected to be commercially available within the next two years. From the results of the study it appears to be feasible to safely operate the TR-2 reactor using LEU fuel without increased fuel cycle costs or decreased performance using U 2 O 8 fuels with densities in the 2.3 to 3.0 gU/cm 3 range. (author)

  12. The outline of clearance plan for Rokkasho uranium enrichment plant

    International Nuclear Information System (INIS)

    Kojima, Takuo; Sasaki, Hitoshi; Shouno, Shuuzou; Nozawa, Kenji

    2011-01-01

    Japan Nuclear Fuel Limited (JNFL) started operation of uranium enrichment by metal cylinder centrifuge at Rokkasho Uranium Enrichment Plant in 1992. Since operation start, JNFL has extended the plant capacity sequentially, but metal cylinder centrifuges ceased operation gradually with time. Replacement to advanced centrifuge is under construction now. Generally, Uranium Enrichment Plant continues operation by replacing centrifuges after a certain period of operation. So, many used centrifuges (metal waste) are generated through the operation period. JNFL is now considering the disposal plan. We can reduce the radioactivity level that is not necessary to treat as the radioactive waste by decontaminating the radioactive material sticking to the surface of metal materials of used centrifuge. And JNFL plants to recycle (reuse) metal material by making much of the clearance system. (author)

  13. Enrichment measurement in TRIGA type fuels

    International Nuclear Information System (INIS)

    Aguilar H, F.; Mazon R, R.

    2001-05-01

    The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

  14. New Prototype Safeguards Technology Offers Improved Confidence and Automation for Uranium Enrichment Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Brim, Cornelia P.

    2013-04-01

    An important requirement for the international safeguards community is the ability to determine the enrichment level of uranium in gas centrifuge enrichment plants and nuclear fuel fabrication facilities. This is essential to ensure that countries with nuclear nonproliferation commitments, such as States Party to the Nuclear Nonproliferation Treaty, are adhering to their obligations. However, current technologies to verify the uranium enrichment level in gas centrifuge enrichment plants or nuclear fuel fabrication facilities are technically challenging and resource-intensive. NNSA’s Office of Nonproliferation and International Security (NIS) supports the development, testing, and evaluation of future systems that will strengthen and sustain U.S. safeguards and security capabilities—in this case, by automating the monitoring of uranium enrichment in the entire inventory of a fuel fabrication facility. One such system is HEVA—hybrid enrichment verification array. This prototype was developed to provide an automated, nondestructive assay verification technology for uranium hexafluoride (UF6) cylinders at enrichment plants.

  15. 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

  16. The Passive Neutron Enrichment Meter for Uranium Cylinder Assay

    Energy Technology Data Exchange (ETDEWEB)

    Miller, Karen A.; Menlove, Howard O.; Swinhoe, Martyn T.; Marlow, Johanna B. [Safeguards Science and Technology Group (N-1), Los Alamos National Laboratory, Los Alamos (United States)

    2011-12-15

    As fuel cycle technology becomes more prevalent around the world, international safeguards have become increasingly important in verifying that nuclear materials have not been diverted. Uranium enrichment technology is a critical pathway to nuclear weapons development, making safeguards of enrichment facilities especially important. Independently-verifiable material accountancy is a fundamental measure in detecting diversion of nuclear materials. This paper is about a new instrument for uranium cylinder assay for enrichment plant safeguards called the Passive Neutron Enrichment Meter (PNEM). The measurement objective is to simultaneously verify uranium mass and enrichment in Uf6 cylinders. It can be used with feed, product, and tails cylinders. Here, we consider the enrichment range up to 5% {sup 235}U. The concept is to use the Doubles-to-Singles count rate to give a measure of the {sup 235}U enrichment and the Singles count rate to provide a measure of the total uranium mass. The cadmium ratio is an additional signature for the enrichment that is especially useful for feed and tails cylinders. PNEM is a {sup 3}He-based system that consists of two portable detector pods. Uranium enrichment in UF{sub 6} cylinders is typically determined using a gamma-ray-based method that only samples a tiny volume of the cylinder's content and requires knowledge of the cylinder wall thickness. The PNEM approach has several advantages over gamma-ray-based methods including a deeper penetration depth into the cylinder, meaning it can be used with heterogeneous isotopic mixtures of UF{sub 6}. In this paper, we describe a Monte Carlo modelling study where we have examined the sensitivity of the system to systematic uncertainties such as the distribution of UF{sub 6} within the cylinder. We also compare characterization measurements of the PNEM prototype to the expected measurements calculated with Monte Carlo simulations.

  17. Data base for a CANDU-PHW operating on a once-through, slightly enriched uranium cycle (AECL-6594)

    International Nuclear Information System (INIS)

    1979-07-01

    This report, prepared for INFCE, gives data for an extrapolated 1000 MW(e) CANDU-PHW design operating on a once-through fuel cycle with a feed fuel of slightly enriched uranium - 1.2 weight % U-235 in uranium. The effects of varying fuel enrichment, maximum channel power, and economic parameters are also discussed

  18. Uranium enrichment: investment options for the long term

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    The US government supplies a major portion of the enriched uranium used to fuel most of the nuclear power plants that furnish electricity in the free world. As manager of the US uranium enrichment concern, the Department of Energy (DOE) is investigating a number of technological choices to improve enrichment service and remain a significant world supplier. The Congress will ultimately select a strategy for federal investment in the uranium enrichment enterprise. A fundamental policy choice between possible future roles - that of the free world's main supplier of enrichment services, and that of a mainly domestic supplier - will underlie any investment decision the Congress makes. The technological choices are gaseous diffusion, gas centrifuge, and atomic vapor laser isotope separation (AVLIS). A base plan and four alternatives were examined by DOE and the Congressional Budget Office. In terms of total enterprise costs, Option IV, ultimately relying on advanced gas centrifuges for enrichment services, would offer the most economic approach, with costs over the full projection period totaling $123.5 billion. Option III, ultimately relying on AVLIS without gas centrifuge enrichment or gaseous diffusion, falls next in the sequence, with costs of $128.2 billion. Options I and II, involving combinations of the gas centrifuge and AVLIS technologies, follow closely with costs of $128.7 and $129.6 billion. The base plan has costs of $136.8 billion over the projection period. 1 figure, 22 tables

  19. 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...

  20. The evolution of the enriched uranium markets

    International Nuclear Information System (INIS)

    Arnaiz, J.; Moleres, C.; Tarin, F.

    2004-01-01

    This paper deals with the evolution of the enriched uranium component markets (uranium concentrates, conversion and enrichment), starting with the situation of historically low prices that occurred during 2000. The situation that has been reached as on December 2003, when the concentrates and conversion markets were 44% and 70% (current US$) respectively, and the enrichment prices 30%, higher, is analysed. Finally, the negative impact of the 90's depressed prices, due to abundant alternative sources of uranium components, on the primary production of all three components and, as a conclusion, the impact of the new situation on the transport logistics, and the need of appropriate economic conditions to make the future primary production sustainable, is commented. (Author)

  1. 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)

  2. Industrial aspects in uranium enrichment

    International Nuclear Information System (INIS)

    Mezin, M.

    1982-05-01

    Characteristics of isotope separation processes in operation and under development are discussed. These include the number of stages in series, the number of components, the component unit capacity and enery requirements. The implementation of an enrichment process and the question of an enrichment plant in Australia are also considered

  3. Uranium enrichment: heading for the abyss

    International Nuclear Information System (INIS)

    Norman, C.

    1983-01-01

    This article discusses the federal government's $2.3 billion a year business enriching uranium for nuclear power plants which is heading toward a major crisis. Due to miscalculations by the Department of Energy, it is caught with billions of dollars of construction in progress just as projected demand for enriched uranium is decreasing. At the center of the controversy is the Gas Centrifuge Plant at Portsmouth, Ohio - estimated to cost $10 billion dollars. A review of how DOE got into this situation and how they plan to solve it is presented

  4. Distribution of uranium supply and enrichment

    International Nuclear Information System (INIS)

    Bamford, F.W.

    1982-01-01

    Uranium supply and demand is examined from the perspective of companies in the uranium hexafluoride (UF6) conversion business whose main interest is their sources of uranium supply and UF6 destinations because of transportation costs. Because of the variations in yellowcake transport, charges for conversion, and UF6 transport costs, most converters don't have standard prices. Companies try to look ahead to determine patterns of supplies and delivery points when they analyze the market and estimate future prices. Market analyses must take into account the purchasing policies of utilities around the world. The presentation shows North America supplying about 40% of world uranium, with about 13% of the enrichment done elsewhere. It also shows North American converters getting 53% of the business, but that will require importing uranium from outside North America. 6 tables

  5. 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.)

  6. Airborne uranium, its concentration and toxicity in uranium enrichment facilities

    International Nuclear Information System (INIS)

    Thomas, J.; Mauro, J.; Ryniker, J.; Fellman, R.

    1979-02-01

    The release of uranium hexafluoride and its hydrolysis products into the work environment of a plant for enriching uranium by means of gas centrifuges is discussed. The maximum permissible mass and curie concentration of airborne uranium (U) is identified as a function of the enrichment level (i.e., U-235/total U), and chemical and physical form. A discussion of the chemical and radiological toxicity of uranium as a function of enrichment and chemical form is included. The toxicity of products of UF 6 hydrolysis in the atmosphere, namely, UO 2 F 2 and HF, the particle size of toxic particulate material produced from this hydrolysis, and the toxic effects of HF and other potential fluoride compounds are also discussed. Results of an investigation of known effects of humidity and temperature on particle size of UO 2 F 2 produced by the reaction of UF 6 with water vapor in the air are reported. The relationship of the solubility of uranium compounds to their toxic effects was studied. Identification and discussion of the standards potentially applicable to airborne uranium compounds in the working environment are presented. The effectiveness of High Efficiency Particulate (HEPA) filters subjected to the corrosive environment imposed by the presence of hydrogen fluoride is discussed

  7. Critical experiments in AQUILON with fuels slightly enriched in uranium 235 or in plutonium; Experiences critiques dans aquilon portant sur des combustibles legerement enrichis en uranium 235 et en plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Chabrillac, M; Ledanois, G; Lourme, P; Naudet, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Reactivity comparisons have been, made in Aquilon II between geometrically identical lattices differing only by the composition of the fuel. The fuel elements consist in metallic uranium single rods with either slight differences of the isotopic composition (0.69 - 0.71 - 0.83 - 0.86 per cent of uranium 235) or slight additions of plutonium (0.043 per cent). Five lattices pitches have bean used, in order to produce a large variation of spectrum. Two additional sets of plutonium fuels are prepared to be used in the same conditions. The double comparisons: natural enriched 235 versus natural-enriched plutonium are made in such a way that a very precise interpretation is permitted. The results are perfectly consistent which seems to prove that the calculation methods are convenient. Further it can been inferred that the usual data, namely for the ratio of the {eta} of {sup 235}U and {sup 239}Pu seem reliable. (authors) [French] On a compare neutroniquement dans Aquilon II des reseaux geometriquement identiques mais comportant de petites differences de composition du combustible. EL s'agit de barres d'uranium metallique, les unes avec des teneurs differentes en isotopes 235 (0,69 - 0,71 - 0,83 - 0,86 pour cent) les autres comportant une legere addition de plutonium (0,043 pour cent). Les comparaisons ont ete faites a cinq pas differents, de maniere a mettre en jeu une assez large variation de spectre. Deux autres jeux de combustible au plutonium seront utilises ulterieurement dans les memes conditions. Les resultats des mesures se presentent sous forme de doubles comparaisons: naturel-enrichi 235/naturel-enrichi plutonium. On s'est place dans des conditions qui permettent des interpretations tres precises. Les resultats sont remarquablement coherents, ce qui semble montrer que les methodes de calcul sont bien adaptees, Ils tendent d'autre part a prouver que les valeurs numeriques admises dans la litterature, notamment pour le rapport des {eta} de l'U 235 et de Pu 239

  8. NRC licensing of uranium enrichment plants

    International Nuclear Information System (INIS)

    Moran, B.W.

    1991-01-01

    The US Nuclear Regulatory Commission (NRC) is preparing a rule making that establishes the licensing requirements for low-enriched uranium enrichment plants. Although implementation of this rule making is timed to correspond with receipt of a license application for the Louisiana Energy Services centrifuge enrichment plant, the rule making is applicable to all uranium enrichment technologies. If ownership of the US gaseous diffusion plants and/or atomic vapor laser isotope separation is transferred to a private or government corporation, these plants also would be licensable under the new rule making. The Safeguards Studies Department was tasked by the NRC to provide technical assistance in support of the rule making and guidance preparation process. The initial and primary effort of this task involved the characterization of the potential safeguards concerns associated with a commercial enrichment plant, and the licensing issues associated with these concerns. The primary safeguards considerations were identified as detection of the loss of special nuclear material, detection of unauthorized production of material of low strategic significance, and detection of production of uranium enriched to >10% 235 U. The primary safeguards concerns identified were (1) large absolute limit of error associated with the material balance closing, (2) the inability to shutdown some technologies to perform a cleanout inventory of the process system, and (3) the flexibility of some technologies to produce higher enrichments. Unauthorized production scenarios were identified for some technologies that could prevent conventional material control and accounting programs from detecting the production and removal of 5 kg 235 U as highly enriched uranium. Safeguards techniques were identified to mitigate these concerns

  9. Operational experience with the first eighteen slightly enriched uranium fuel assemblies in the Atucha-1 nuclear power plant

    International Nuclear Information System (INIS)

    Higa, M.; Perez, R.; Pineyro, J.; Sidelnik, J.; Fink, J.; Casario, J.A.; Alvarez, L.

    1997-01-01

    Atucha I is a 357 Mwe nuclear station, moderated and cooled with heavy water, pressure vessel type of German design, located in Argentina. Fuel assemblies (FA) are 36 active natural UO2 rod clusters, 5.3 meters long and fuel is on power. Average FA exit burnup is 6 MWd/kg U. The reactor core contains 252 FA. To reduce the fuel costs about 6 MU$S/yr a program of utilization of SEU (0.85 %w U235) fuel was started at the beginning of 1995 with the introduction of 12 FA in the first step. The exit burnup of FA is approx. 10 MWd/kgU. It is planned to increase gradually the number of them up to having a full core with SEU fuel with an expected FA average exit burnup of 11 MWd/kgU. The SEU program has also the advantage of a strong reduction of spent fuel volume, and a moderate reduction of fuelling machine use. This paper presents the satisfactory operation experience with the introduction of the first 12 SEU fuel assemblies and the planned activities for the future. The fresh SEU fuel assemblies were introduced in six fuel channels located in an intermediate zone located 136 cm from the center of the reactor and selected because they have higher margins to the channel powers limits to accommodate the initial 15 to 20 % relative channel power increase. To verify the design and fuel management calculations, comparisons have been made of the calculated and measured values of the variation of channel ΔT, regulating rods insertion and flux reading in in-core detectors near to the refueled channel. The agreement was good and in most of the cases within the measurement errors. Cell calculations were made with WIMS-D4, and reactor calculations with PUMA. a fuel management 3D diffusion program developed in Argentina. With SEU fuel with a greater burnup in the central high power core region, new operating procedures were developed to prevent PCI failures in fuel power ramps that arise during operation. Some fuel rod and structural assembly design changes were introduced on the

  10. Contribution of CERCA to the US DOE conference on the use of 20% and 45% enriched uranium as fuel for research reactors [contributed by J. Doumerc, CERCA

    International Nuclear Information System (INIS)

    Doumerc, J.

    1993-01-01

    This paper speaks only of prices. Some basic statements can be provided. All the results which have been displayed by Mr. Dewez represent the CERCA work performed within the last twelve months. We have invested in this a little less than million French francs, which is roughly 220 000. As far as prices are concerned, for the time being, we have to compare the prices of a steady state situation, which is represented by a well known process that has been in use for many years, with a transition situation which is the achievement of the same expertise in new, extrapolated fuels. That is why the comparison has to be corrected for the results within the next 2 to 3 years. Obviously, it seems that there are other factors which contribute to the price increases. I think there is a very significant example for this. When you have to introduce a given amount for 235-U in the fuel, either 93 or 20% enrichment, you need in the second case a higher total uranium content, which means that you have to convert two to four times more uranium from UF 6 to a uranium compound then from uranium compound to powder. Obviously, you cannot prepare Kg of some product at one price and 250 g of the same product at the same price. Moreover, there is some chance that the new process fabrication will be slightly more difficult to achieve than the previous one. We have observed in the past that progress was continuously improved, until we reached something like a steady state situation. Now alloy yields, for instance, are the same for all manufacturers, except ± 1% depending on the day-to-day events of the manufacturer itself. There is a very good reason to consider that the progress in the yields will be the same for the manufacturer of extrapolated fuels. In that event it is rather easy to foresee what it will be in the near future. In consideration, besides explaining some reasons as to why prices will become higher, there are many reasons of remaining confident with the final result

  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. Sogin enriched uranium extraction (EUREX) plant spent fuel pool cleaning and decontamination utilizing the Smart Safe solution

    International Nuclear Information System (INIS)

    Denton, M.S.; Gili, M.; Nasta, M.; Quintiliani, R.; Caccia, G.; Botzen, W.; Forrester, K.

    2009-01-01

    SOGIN's EUREX facility in Italy was developed as a pilot plant functional testing laboratory for spent fuel reprocessing. This facility was operated successfully for many years since 1970 and was eventually shutdown consistent with Italy's suspension of all nuclear operations. At the time of suspension, the EUREX facility still had spent nuclear fuel assemblies in storage from a nearby PWR. Other fuel assemblies from an Italian AGR had remained stored in the spent fuel pool for the 20 years or so waiting for removal and reprocessing abroad. Being Magnox fuel elements, their recovery for the transport produced a huge amount of sludge in the pool. During this time, sediment, dirt, corrosion products, fuel cladding, etc. has collected within the fuel pool as a crud layer dispersed throughout. Most of this crud has accumulated on the horizontal surfaces of the pool and fuel element assemblies, while some remains as a suspended colloidal material. Furthermore many other contaminated metal components, used during the operation years, where still inside the pool. Due to a pool leak discovered in 2006, SOGIN speeded up its pool decommissioning program, making also available the transfer of the spent fuel to a nearby interim repository, with the goal to completely drain the pool in the shortest period of time. In order for SOGIN to successfully transfer the fuel assemblies from their current storage basket locations to the spent fuel transfer cask, the bulk of the crud needed to be removed. This cleanup operation was deemed necessary to minimize the suspension of contamination in the water during underwater handling operations. This would reduce the decontamination efforts on the transfer cask upon removal, once loaded with the spent fuel, and enhance safety by reducing potential underwater visibility issues. The operations were completed in July 2008 with the release to the environment of the pool water, thoroughly purified and without any relevant radiological impact. The

  13. Analysis of a PHWR slightly enriched fuel

    International Nuclear Information System (INIS)

    Notari, C.; Marajofsky, A.

    1994-01-01

    It is widely known that the use of slightly enriched uranium in PHWR reactors presents economic advantages derived from the fact that less uranium is required for producing the same amount of energy. Several studies related with the use of this alternative in Atucha I NPP have been performed. The fuel assembly geometry considered up to now has been almost identical to the natural uranium one. In this work a modification consisting in the use of annular pellets in the outer ring of the cluster is analyzed. This design produces several performance benefits. The redistribution of the power in the fuel improves the maximum to average bundle power ratio. The improvement achieved depends on the void volume in the pellets which at the same time represents a certain burnup decrease. These parameters (power ratios and burnup loss) are quantified for the Atucha I and Embalse NPPs. This design improves the fuel behaviour with respect to the burnup extension derived from the slight enrichment. It is also interesting in case an overall power increase is considered. (author). 16 refs, 8 figs, 1 tab

  14. Calculation of the fissile mass of a graphite moderated critical assembly using 93% enriched uranium

    International Nuclear Information System (INIS)

    Correa, F.; Marzo, M.A.S.; Collussi, I.; Ferreira, A.C.A.

    1976-01-01

    The critical mass of uranium has been calculated for a graphite moderated set fueled with 93% enriched uranium to be mounted on the Instituto de Energia Atomica split table Zero Power Reactor. The core composition was optimized to permit the maximum number of configurations to be studied. Analysis of three core compositions shows that 8 Kg of uranium enriched to 93% - U-235 (by weight) and 100 Kg of thorium would be sufficient for criticality experiments [pt

  15. The supply of the European community countries with enriched uranium

    International Nuclear Information System (INIS)

    1975-02-01

    A discussion is given of a survey regarding the supply of enriched uranium to the countries of the European Community. Costs of enriched uranium imports were not available but import values were calculated using world market prices. (R.L.)

  16. DOE hands over uranium enrichment duties to government corporation

    International Nuclear Information System (INIS)

    Simpson, J.

    1993-01-01

    In an effort to renew the United States' competitiveness in the world market for uranium enrichment services, the Department of Energy (DOE) is turning over control of its Paducah, KY, and Portsmouth, OH, enrichment facilities to a for-profit organization, the United States Enrichment Corp. (USEC), which was created by last year's Energy Policy Act. William H. Timbers, Jr., a former investment banker who was appointed acting CEO in March, said the Act's mandate will mean more competitive prices for enriched reactor fuel and greater responsiveness to utility customers. As a government corporation, USEC, with current annual revenues estimated at $1.5 billion, will no longer be part of the federal budget appropriations process, but will use business management techniques, set market-based prices for enriched uranium, and pay annual dividends to the US Treasury-its sole stockholder-from earnings. The goal is to finish privatizing the corporation within two years, and to sell its stock to investors for an estimated $1 to $3 billion. USEC's success will depend in part on developing short- and long-term marketing plants to help stanch the flow of enriched-uranium customers to foreign suppliers. (DOE already has received notice from a number of US utilities that they want to be let out of their long-term enrichment contracts as they expire over the next several years).USEC's plans likely will include exploring new joint ventures with other businesses in the nuclear fuel cycle-such as suppliers, fabricators, and converters-and offering a broader range of enrichment services than DOE provided. The corporation will have to be responsive to utilities on an individual basis

  17. Feasibility of uranium enrichment in Australia

    International Nuclear Information System (INIS)

    1979-10-01

    The Council considered that provided the balance between costs and markets was found to be acceptable, there was no valid reason against the Government proceeding with a study on the feasibility of, and perhaps participating in the establishment of a commercial uranium enrichment industry in Australia. Areas covered include technical expertise and industrial structure in Australia, environmental aspects and safeguards

  18. Topical papers on uranium conversion and enrichment

    International Nuclear Information System (INIS)

    Uranium conversion and enrichment are discussed in 5 papers by representatives of the USA, Great Britain and Switzerland. The state of the art is reviewed, and future prospects are given. Supply assurance is directly related to the necessary production capacities and the supply agreements

  19. Multinational uranium enrichment in the Middle East

    International Nuclear Information System (INIS)

    Ahmad, Ali; Salahieh, Sidra; Snyder, Ryan

    2017-01-01

    The Joint Comprehensive Plan of Action (JCPOA) agreed to by Iran and the P5+1 in July 2015 placed restrictions on Iran’s nuclear program while other Middle Eastern countries– Egypt, Jordan, Saudi Arabia, Turkey, and the United Arab Emirates–are planning to build their own nuclear power plants to meet increasing electricity demands. Although the JCPOA restricts Iran's uranium enrichment program for 10–15 years, Iran's neighbors may choose to develop their own national enrichment programs giving them a potential nuclear weapons capability. This paper argues that converting Iran's national enrichment program to a more proliferation-resistant multinational arrangement could offer significant economic benefits–reduced capital and operational costs–due to economies of scale and the utilization of more efficient enrichment technologies. In addition, the paper examines policy aspects related to financing, governance, and how multinational enrichment could fit into the political and security context of the Middle East. A multinational enrichment facility managed by regional and international partners would provide more assurance that it remains peaceful and could help build confidence between Iran and its neighbors to cooperate in managing other regional security challenges. - Highlights: • Freezing Iran's nuclear program is an opportunity to launch joint initiatives in ME. • A joint uranium enrichment program in the Middle East offers economic benefits. • Other benefits include improved nuclear security and transparency in the region.

  20. 31 CFR 540.306 - Highly Enriched Uranium (HEU).

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Highly Enriched Uranium (HEU). 540...) OFFICE OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.306 Highly Enriched Uranium (HEU). The term highly...

  1. Long term assurance of supply of uranium enrichment

    International Nuclear Information System (INIS)

    1978-01-01

    After elaborating a number of key questions on uranium enrichment, the representatives of 10 countries and of the EC commission present their answers. Attention is paid to the assurance of uranium supply, to uranium enrichment, market trends and flexibility in enrichment agreements

  2. 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)

  3. 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

  4. Uranium enrichment management review: summary of analysis

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-01

    In May 1980, the Assistant Secretary for Resource Applications within the Department of Energy requested that a group of experienced business executives be assembled to review the operation, financing, and management of the uranium enrichment enterprise as a basis for advising the Secretary of Energy. After extensive investigation, analysis, and discussion, the review group presented its findings and recommendations in a report on December 2, 1980. The following pages contain background material on which that final report was based. This report is arranged in chapters that parallel those of the uranium enrichment management review final report - chapters that contain summaries of the review group's discussion and analyses in six areas: management of operations and construction; long-range planning; marketing of enrichment services; financial management; research and development; and general management. Further information, in-depth analysis, and discussion of suggested alternative management practices are provided in five appendices.

  5. Uranium enrichment management review: summary of analysis

    International Nuclear Information System (INIS)

    1981-01-01

    In May 1980, the Assistant Secretary for Resource Applications within the Department of Energy requested that a group of experienced business executives be assembled to review the operation, financing, and management of the uranium enrichment enterprise as a basis for advising the Secretary of Energy. After extensive investigation, analysis, and discussion, the review group presented its findings and recommendations in a report on December 2, 1980. The following pages contain background material on which that final report was based. This report is arranged in chapters that parallel those of the uranium enrichment management review final report - chapters that contain summaries of the review group's discussion and analyses in six areas: management of operations and construction; long-range planning; marketing of enrichment services; financial management; research and development; and general management. Further information, in-depth analysis, and discussion of suggested alternative management practices are provided in five appendices

  6. Energy consumption of chemical uranium enrichment

    International Nuclear Information System (INIS)

    Miyake, T.; Takeda, K.; Obanawa, H.

    1987-01-01

    A quantitative study of chemical separation energy for enriching uranium-235 by the redox chromatography was conducted. Isotope exchange reactions between U 4+ -UO 2 2+ ions in the enrichment column are maintained by the redox reactions. The chemical separation energy is ultimately supplied by hydrogen and oxygen gas for regenerating redox agents. The redox energy for the isotope separation is theoretically predicted as a function of the dynamic enrichment factor observed in the chromatographic development of uranium adsorption band. Thermodynamic treatments of the equilibrium reactions implies and inverse redox reaction which can be enhanced by the chemical potential of the ion-exchange reaction of oxidant. Experimental results showed 30 to 90% recovery of the redox energy by the inverse reaction. These results will devise a simplified redox chromatography process where a number of columns in one module is reduced

  7. 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.

  8. Use of highly enriched uranium in the material testing reactor BR2

    International Nuclear Information System (INIS)

    Beeckmans de West-Meerbeeck, A.

    1979-05-01

    In the material testing reactor BR2, the use of highly enriched uranium is determined by the consideration of the fast, epithermal and thermal neutron flux effectively available for the experimental devices. The choice of the core configuration is defined by combining the localisation of the experimental devices and of fuel elements of various burnup, such as to satisfy the irradiation conditions of the experimental load, compatible with an economic use of the fuel elements and safe operation of the reactor. Taking into account the present manufacturing technology for MTR fuels (37 Wt % uranium density in the fuel meat) the highly enriched uranium cannot be avoided; if higher concentration of uranium could be realised by some new manufacturing technology, the 235 U density of fuel elements at elimination should be kept at the required level and the enrichment could be reduced accordingly

  9. Use of highly enriched uranium in the material testing reactor BR2

    International Nuclear Information System (INIS)

    Beeckmans de West-Meerbeeck, A.

    1979-05-01

    In the material testing reactor BR2, the use of highly enriched uranium is determined by the consideration of the fast, epithermal and thermal neutron flux effectively available for the experimental devices. The choice of the core configuration is defined by combining the localisation of the experimental devices and of fuel elements of various burnup, such as to satisfy the irradiation conditions of the experimental load, compatible with an economic use of the fuel elements and safe operation of the reactor. Taking into account the present manufacturing technology for MTR fuels (37 Wt % uranium density in the fuel meat) the highly enriched uranium cannot be avoided: if higher concentration of uranium could be realised by some new manufacturing technology, the 235 U density of fuel elements at elimination should be kept at the required level and the enrichment could be reduced accordingly. (author)

  10. Effect of reduced enrichment on the fuel cycle for research reactors

    International Nuclear Information System (INIS)

    Travelli, A.

    1982-01-01

    The new fuels developed by the RERTR Program and by other international programs for application in research reactors with reduced uranium enrichment (<20% EU) are discussed. It is shown that these fuels, combined with proper fuel-element design and fuel-management strategies, can provide at least the same core residence time as high-enrichment fuels in current use, and can frequently significantly extend it. The effect of enrichment reduction on other components of the research reactor fuel cycle, such as uranium and enrichment requirements, fuel fabrication, fuel shipment, and reprocessing are also briefly discussed with their economic implications. From a systematic comparison of HEU and LEU cores for the same reference research reactor, it is concluded that the new fuels have a potential for reducing the research reactor fuel cycle costs while reducing, at the same time, the uranium enrichment of the fuel

  11. Technical problems in case of utilizing uranium of medium enrichment for a research reactor

    International Nuclear Information System (INIS)

    Kanda, Keiji; Shibata, Shun-ichi

    1979-01-01

    Usually, highly enriched uranium of 90 - 93% is used for research reactors, but the US government proposed the strong policy to use low enriched uranium of the uranium of medium enrichment in unavoidable case from the viewpoint of the resistance to nuclear proliferation in November, 1977. This policy is naturally applied to Japan also. The export of highly enriched uranium will be permitted only when the President approves it after the technical and economical evaluations by the government. The Kyoto University high flux reactor has the features which are not seen in other research reactors, such as medical irradiation, and it is hard to attain the objectives of researches unless HEU is used. The application for the export of HEU was accepted in February, 1978. The nuclear characteristics of the KUHFR when medium or low enriched uranium is used, the criticality experiment in the KUCA using the uranium of medium enrichment, and the burning test on the uranium fuel plates of medium enrichment are described. The research project to lower the degree of enrichment in the fuel for research and test reactors is expected to be continued down to less than 20%. The MEU of 45% enrichment will be actually used in 1983. (Kako, I.)

  12. Critical experiments on enriched uranium graphite moderated cores

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko; Akino, Fujiyoshi; Kitadate, Kenji; Kurokawa, Ryosuke

    1978-07-01

    A variety of 20 % enriched uranium loaded and graphite-moderated cores consisting of the different lattice cells in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments systematically. In the present report, the experimental results for homogeneously or heterogeneously fuel loaded cores and for simulation core of the experimental reactor for a multi-purpose high temperature reactor are filed so as to be utilized for evaluating the accuracy of core design calculation for the experimental reactor. The filed experimental data are composed of critical masses of uranium, kinetic parameters, reactivity worths of the experimental control rods and power distributions in the cores with those rods. Theoretical analyses are made for the experimental data by adopting a simple ''homogenized cylindrical core model'' using the nuclear data of ENDF/B-III, which treats the neutron behaviour after smearing the lattice cell structure. It is made clear from a comparison between the measurement and the calculation that the group constants and fundamental methods of calculations, based on this theoretical model, are valid for the homogeneously fuel loaded cores, but not for both of the heterogeneously fuel loaded cores and the core for simulation of the experimental reactor. Then, it is pointed out that consideration to semi-homogeneous property of the lattice cells for reactor neutrons is essential for high temperature graphite-moderated reactors using dispersion fuel elements of graphite and uranium. (author)

  13. Detection of uranium enrichment activities using environmental monitoring techniques

    International Nuclear Information System (INIS)

    Belew, W.L.; Carter, J.A.; Smith, D.H.; Walker, R.L.

    1993-01-01

    Uranium enrichment processes have the capability of producing weapons-grade material in the form of highly enriched uranium. Thus, detection of undeclared uranium enrichment activities is an international safeguards concern. The uranium separation technologies currently in use employ UF 6 gas as a separation medium, and trace quantities of enriched uranium are inevitably released to the environment from these facilities. The isotopic content of uranium in the vegetation, soil, and water near the plant site will be altered by these releases and can provide a signature for detecting the presence of enriched uranium activities. This paper discusses environmental sampling and analytical procedures that have been used for the detection of uranium enrichment facilities and possible safeguards applications of these techniques

  14. Uranium enrichment by centrifuge in Japan

    International Nuclear Information System (INIS)

    Watanabe, T.; Murase, T.

    1977-01-01

    The demand for enriched uranium is on the increase with nuclear power capacity in which the LWR predominates and is estimated to exceed the supply from the present facilities in the world in less than ten years. Therefore, the basic strategy for enriched uranium is investigated on the following three-point long-range program in Japan: 1. To continue negotiations to extend the current allocation by the long-term contract; 2. To seek active participation in international enrichment projects; and 3. To make efforts to develop uranium enrichment technology and to construct inland facilities. On this basis, a vigorous development program of gas centrigue process for industrialization was launched out in 1972 as a national project. Ever since substantial progress in this field has been made and development works have been increased year after year. At present, a concrete plan of a pilot plant is taking shape. Up to now, several types of centrifuges were developed, of which some were completed as prototype models, and subjected to life tests and also to extensive earthquake-resistivity tests for the characteristics of Japanese geological condition. An enrichment plant is composed of so many centrifuges that the installation and piping system of centrifuges is an important factor which has an effect on plant economy and reliability. Two types of the experimental cascade were constructed in Japan. One has been in operation since 1973, and the other since 1975. Valuable empirical data have been accumulated on cascade characteristics, maintenance scheme and so on. It will be important for the coming plants to have a flexibility to escalation of labor and energy cost, or to variation of the separative work requirement and further. An economic prospect of centrifuge enrichment process is presented

  15. Uranium enrichment activities: the SILVA program

    International Nuclear Information System (INIS)

    Guyot, J.; Cazalet, J.; Camarcat, N.; Figuet, J.

    1994-01-01

    Through its commitment to a nuclear electricity generation policy, France holds today a specific position in the uranium enrichment market thanks to the modern multinational EURODIF gaseous diffusion plant. France has, altogether, a long-term goal in developing SILVA, a laser uranium enrichment process, based on the selective photo-ionization of U-235. After reviewing the fundamentals of SILVA (the laser system with copper vapor lasers and dye lasers and the separator system), a description of the general organization of the R and D program is provided going through basic research, subsystems assessment, production demonstrations and simulations (with the LACAN code), plant design and economics. The general schedule of SILVA is outlined, leading to the possible construction of a commercial plant. 7 figs., 11 refs

  16. Computational fluid mechanics in R and D on uranium enrichment

    International Nuclear Information System (INIS)

    Soubbaramayer, O.

    1988-01-01

    Uranium enrichment represents an essential link in the cycle of nuclear fuels for power production. There are many processes of uranium enrichment, but three of them dominate the nuclear history as well in the past (Gaseous diffusion and centrifugation) as in the present (Laser process). The important role played by the Numerical Fluid Mechanics in the three processes is pointed out. The type of problem raised by Gaseous Diffusion is Channel Flow with wall suction, by Centrifugation, flow of a Compressible gas in a strongly rotating cylinder (Stewartson and Ekman layers) and by Laser process, Thermocapillary-buoyancy flow of a molten metal in an evaporation crucible. The methods and results in these problems are reviewed. 18 refs, 11 figs

  17. Enriching services and uranium markets from an international supplier's perspective

    International Nuclear Information System (INIS)

    Guais, J.C.; Peterson, C.H.

    1987-01-01

    In the past 10 yr, the international market for nuclear fuel cycle products and services, mainly yellowcake and separative work units (SWUs), has been characterized by shifting patterns in global demand and resource allocation. For suppliers as well as buyers, the analysis of the supply and demand pattern over the next 10 yr can be a useful exercise, especially if it highlights geopolitical realities and the influence of new technologies. This paper will attempt to answer two questions related to the marketing of natural uranium and enrichment services: (1) on a global basis (i.e., for the Western World), are uranium and enrichment capacities now on line sufficient to meet long-term demand. (2) On a regional basis (the US, Europe, and Far East), what is the short-to-medium term pattern of the supply-and-demand balance

  18. Use of enriched uranium in Canada's power reactors

    International Nuclear Information System (INIS)

    Dormuth, K.W.; Jackson, D.P.

    2011-01-01

    Recent trends in Canadian nuclear power reactor design and proposed development of nuclear power in Canada have indicated the possibility that Canada will break with its tradition of natural uranium fuelled systems, designed for superior neutron economy and, hence, superior uranium utilization. For instance, the Darlington B new reactor project procurement process included three reactor designs, all employing enriched fuel, although a natural uranium reactor design was included at a late stage in the ensuing environmental assessment for the project as an alternative technology. An evaluation of the alternative designs should include an assessment of the environmental implications through the entire fuel cycle, which unfortunately is not required by the environmental assessment process. Examples of comparative environmental implications of the reactor designs throughout the fuel cycle indicate the importance of these considerations when making a design selection. As Canada does not have enrichment capability, a move toward the use of enriched fuel would mean that Canada would be exporting natural uranium and buying back enriched uranium with value added. From a waste management perspective, Canada would need to deal with mill, refinery, and conversion tailings, as well as with the used fuel from its own reactors, while the enrichment supplier would retain depleted uranium with some commercial value. On the basis of reasoned estimates based on publicly available information, it is expected that enrichment in Canada is likely to be more profitable than exporting natural uranium and buying back enriched uranium. Further, on the basis of environmental assessments for enrichment facilities in other countries, it is expected that an environmental assessment of a properly sited enrichment facility would result in approval. (author)

  19. Uranium enrichment: a competitive market in the future?

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Andre Ferreira; Honaiser, Eduardo Henrique Rangel [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil)]. E-mail: 20-1@ctemsp.mar.mil.br

    2005-07-01

    Uranium enrichment is the costly step in the nuclear fuel cycle. It has born as a an activity for the military in the 40s, financed by governments, such as the United States (US) and the former Soviet Union. Later, other major nations have joined them in the nuclear weapons development. The activity of enrichment was done in each country that developed nuclear weapons, and the nuclear weapons countries, especially the US and Soviet Union, dictated the mined uranium market. In the 70s, with the growth of the commercial use of nuclear energy, uranium enrichment started to be treated as a market, which gradually have structured itself, strongly influenced by the historical background. Today, the market is an oligopoly of four major government-owned (or government-influenced) companies. In this paper, the trends in the enrichment market are identified, focusing on competitiveness. Through the conduction of a market analysis (past and future), and the study of the market structure evolution, a more competitive market is shown, but still influenced by the governmental participation. Competitiveness is dictated by government support, verticalization capacity, and, mainly by technological advantages. (author)

  20. Uranium enrichment: a competitive market in the future?

    International Nuclear Information System (INIS)

    Marques, Andre Ferreira; Honaiser, Eduardo Henrique Rangel

    2005-01-01

    Uranium enrichment is the costly step in the nuclear fuel cycle. It has born as a an activity for the military in the 40s, financed by governments, such as the United States (US) and the former Soviet Union. Later, other major nations have joined them in the nuclear weapons development. The activity of enrichment was done in each country that developed nuclear weapons, and the nuclear weapons countries, especially the US and Soviet Union, dictated the mined uranium market. In the 70s, with the growth of the commercial use of nuclear energy, uranium enrichment started to be treated as a market, which gradually have structured itself, strongly influenced by the historical background. Today, the market is an oligopoly of four major government-owned (or government-influenced) companies. In this paper, the trends in the enrichment market are identified, focusing on competitiveness. Through the conduction of a market analysis (past and future), and the study of the market structure evolution, a more competitive market is shown, but still influenced by the governmental participation. Competitiveness is dictated by government support, verticalization capacity, and, mainly by technological advantages. (author)

  1. Civilian inventories of plutonium and highly enriched uranium

    International Nuclear Information System (INIS)

    Albright, D.

    1987-01-01

    In the future, commercial laser isotope enrichment technologies, currently under development, could make it easier for national to produce highly enriched uranium secretly. The head of a US firm that is developing a laser enrichment process predicts that in twenty years, major utilities and small countries will have relatively small, on-site, laser-based uranium enrichment facilities. Although these plants will be designed for the production of low enriched uranium, they could be modified to produce highly enriched uranium, an option that raises the possibility of countries producing highly enriched uranium in small, easily hidden facilities. Against this background, most of this report describes the current and future quantities of plutonium and highly enriched uranium in the world, their forms, the facilities in which they are produced, stored, and used, and the extent to which they are transported. 5 figures, 10 tables

  2. Uranium enrichment export control guide: Gaseous diffusion

    Energy Technology Data Exchange (ETDEWEB)

    1989-09-01

    This document was prepared to serve as a guide for export control officials in their interpretation, understanding, and implementation of export laws that relate to the Zangger International Trigger List for gaseous diffusion uranium enrichment process components, equipment, and materials. Particular emphasis is focused on items that are especially designed or prepared since export controls are required for these by States that are party to the International Nuclear Nonproliferation Treaty.

  3. 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

  4. Status report on uranium enrichment associates

    International Nuclear Information System (INIS)

    Langley, R.A. Jr.; O'Donnell, A.J.; Garrett, G.A.

    1977-01-01

    Uranium Enrichment Associates (UEA) had as its priority project financing, an approach in which the total project is financially self-liquidating. UEA worked with financial institutions to define the combination of assurances and guarantees required by lenders in order to provide the required debt funding. UEA's assets against which the debt liability for the plant would be balanced would be the facilities under construction and the equipment on order. On the customer side, there was major concern on the part of the utilities of whether private industry would be able to complete and operate the plant owing to many of the same possibilities which concerned financial institutions. The disparity between the conditions under which financing could be obtained and the terms acceptable to utilities was a significant element in EUA's choice of process to use for its enrichment plants. UEA's technical staff then began to parallel conceptual designs of gaseous diffusion and gas cenrifuge plants. UEA negotiated with ERDA on the terms of a Cooperative Arrangement, within the provisions of the NFAA, providing the minimum conditions necessary to obtain financing and contracts with utilities for enrichment sources. The UEA plant has several features different from the ERDA plants. The UEA plant used only two basic stage sizes. The UEA design employed four main process buildings. The partners in UEA have mutually agreed to follow the private uranium enrichment project to a logical conclusion. 6 figures

  5. Standard specification for uranium hexafluoride enriched to less than 5 % 235U

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This specification covers nuclear grade uranium hexafluoride (UF6) that either has been processed through an enrichment plant, or has been produced by the blending of Highly Enriched Uranium with other uranium to obtain uranium of any 235U concentration below 5 % and that is intended for fuel fabrication. The objectives of this specification are twofold: (1) To define the impurity and uranium isotope limits for Enriched Commercial Grade UF6 so that, with respect to fuel design and manufacture, it is essentially equivalent to enriched uranium made from natural UF6; and (2) To define limits for Enriched Reprocessed UF6 to be expected if Reprocessed UF6 is to be enriched without dilution with Commercial Natural UF6. For such UF6, special provisions, not defined herein, may be needed to ensure fuel performance and to protect the work force, process equipment, and the environment. 1.2 This specification is intended to provide the nuclear industry with a standard for enriched UF6 that is to be used in the pro...

  6. 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.

  7. 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

  8. Candidate processes for diluting the 235U isotope in weapons-capable highly enriched uranium

    International Nuclear Information System (INIS)

    Snider, J.D.

    1996-02-01

    The United States Department of Energy (DOE) is evaluating options for rendering its surplus inventories of highly enriched uranium (HEU) incapable of being used to produce nuclear weapons. Weapons-capable HEU was earlier produced by enriching uranium in the fissile 235 U isotope from its natural occurring 0.71 percent isotopic concentration to at least 20 percent isotopic concentration. Now, by diluting its concentration of the fissile 235 U isotope in a uranium blending process, the weapons capability of HEU can be eliminated in a manner that is reversible only through isotope enrichment, and therefore, highly resistant to proliferation. To the extent that can be economically and technically justified, the down-blended uranium product will be made suitable for use as commercial reactor fuel. Such down-blended uranium product can also be disposed of as waste if chemical or isotopic impurities preclude its use as reactor fuel

  9. Aerodynamic isotope separation processes for uranium enrichment: process requirements

    International Nuclear Information System (INIS)

    Malling, G.F.; Von Halle, E.

    1976-01-01

    The pressing need for enriched uranium to fuel nuclear power reactors, requiring that as many as ten large uranium isotope separation plants be built during the next twenty years, has inspired an increase of interest in isotope separation processes for uranium enrichment. Aerodynamic isotope separation processes have been prominently mentioned along with the gas centrifuge process and the laser isotope separation methods as alternatives to the gaseous diffusion process, currently in use, for these future plants. Commonly included in the category of aerodynamic isotope separation processes are: (a) the separation nozzle process; (b) opposed gas jets; (c) the gas vortex; (d) the separation probes; (e) interacting molecular beams; (f) jet penetration processes; and (g) time of flight separation processes. A number of these aerodynamic isotope separation processes depend, as does the gas centrifuge process, on pressure diffusion associated with curved streamlines for the basic separation effect. Much can be deduced about the process characteristics and the economic potential of such processes from a simple and elementary process model. In particular, the benefit to be gained from a light carrier gas added to the uranium feed is clearly demonstrated. The model also illustrates the importance of transient effects in this class of processes

  10. Comments on Smith Barney's uranium enrichment analysis

    International Nuclear Information System (INIS)

    Rezendes, V.S.

    1990-07-01

    In a May 1990 report, Smith Barney, Harris Upham and Co. concluded that DOE's uranium enrichment program should be restructured as a government corporation; all past costs have been recovered, and DOE's customers have been overcharged about $1.2 billion; the government should retain responsibility for environment and decommissioning costs associated with enriched uranium production before the corporation's formation; and at some future time the corporation could be sold to the private sector. This report agrees with Smith Barney's recommendation to restructure the enrichment program as a government corporation, but disagrees that DOE's customers have paid for all past costs. According to the author, Smith Barney did not identify the total environmental or decommissioning costs between the government and the corporation. Since these costs are largely undefined, but could amount to billions, Congress should immediately require the program to begin setting aside funds for these costs. DOE estimates that government purchases are responsible for 50 percent of the decommissioning costs; therefore, the government should share these costs by matching the corporation's fund contributions. This requirement should continue until the existing plants have been decommissioned

  11. Status report on the cost and availability of enriched uranium for research reactors

    International Nuclear Information System (INIS)

    Mueller, Hans; Laucht, Juergen

    2005-01-01

    Availability and price development of enriched uranium contained in fuel elements for research reactors plays an important role with regard to reliability and economic and planning reasons. The leading price factors of LEU (19.75% enriched uranium metal), are the contained natural uranium equivalent in the form of UF6 (feed component), the separative work of the enrichment (SWU), conversion of the enriched uranium into metal form and associated services, such as transportation. World market price of feed material for enrichment was more or less stable in the last decades. After very moderate feed price increases between 2001 and mid-2003, the price gained momentum and almost doubled in the short period between the 2nd half of 2003 and year-end 2004. (author)

  12. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle for use in establishing ''as low as practicable'' guides: fabrication of light-water reactor fuel from enriched uranium dioxide

    International Nuclear Information System (INIS)

    Pechin, W.H.; Blanco, R.E.; Dahlman, R.C.; Finney, B.C.; Lindauer, R.B.; Witherspoon, J.P.

    1975-05-01

    A cost-benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model enriched-uranium, light-water reactor (LWR) fuel fabrication plant, and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as practicable'' in relation to limiting the release of radioactive materials from nuclear facilities. The base case model plant is representative of current plant technology and has an annual capacity of 1500 metric tons of LWR fuel. Additional radwaste treatment equipment is added to the base case plants in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The cost for the added waste treatment operations and the corresponding dose commitment are calculated for each case. In the final analysis, radiological dose is plotted vs the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Some of the technology used in the advanced cases is in an early stage of development and is not suitable for immediate use. The methodology used in estimating the costs and the radiological doses, detailed calculations, and tabulations are presented in Appendix A and ORNL-4992. (U.S.)

  13. The case for enrichment of uranium in Australia

    International Nuclear Information System (INIS)

    George, D.W.

    1981-01-01

    Information is presented on the number of nuclear power plants in operation and under construction and on the extent of the use of uranium. The case for enrichment of uranium in Australia is then considered in detail and the status of feasbility studies being carried out is discussed. Arguments to support an enrichment industry include: the need for additional enrichment capacity; added value; potential profitability; increased employment and industrial opportunities; and retention of depleted uranium

  14. Legal and regulatory framework of Uranium's enrichment

    International Nuclear Information System (INIS)

    Antelo, Josefina; Figueredo, Micaela S.; Mangone, Gisela P.; Manin, Maria L.; Pota, Luciana F.

    2009-01-01

    The object of this paper is to develop the legal aspects referred to the activities of uranium's enrichment, in order to achieve the pacific use of nuclear energy and to obey treatments, agreements and international conventions in which Argentine is party and through them assumes the non proliferation's commitment. In this context, we will develop the rights and obligations established in those legal instruments, as well as the juridical concerns of the eventual subscription of Argentine to the Additional Protocol approved by the Board of Governors in 1997. (author)

  15. Active interrogation of highly enriched uranium

    Science.gov (United States)

    Fairrow, Nannette Lea

    Safeguarding special nuclear material (SNM) in the Department of Energy Complex is vital to the national security of the United States. Active and passive nondestructive assays are used to confirm the presence of SNM in various configurations ranging from waste to nuclear weapons. Confirmation measurements for nuclear weapons are more challenging because the design complicates the detection of a distinct signal for highly enriched uranium. The emphasis of this dissertation was to investigate a new nondestructive assay technique that provides an independent and distinct signal to confirm the presence of highly enriched uranium (HEU). Once completed and tested this assay method could be applied to confirmation measurements of nuclear weapons. The new system uses a 14-MeV neutron source for interrogation and records the arrival time of neutrons between the pulses with a high efficiency detection system. The data is then analyzed by the Feynman reduced variance method. The analysis determined the amount of correlation in the data and provided a unique signature of correlated fission neutrons. Measurements of HEU spheres were conducted at Los Alamos with the new system. Then, Monte Carlo calculations were performed to verify hypothesis made about the behavior of the neutrons in the experiment. Comparisons of calculated counting rates by the Monte Carlo N-Particle Transport Code (MCNP) were made with the experimental data to confirm that the measured response reflected the desired behavior of neutron interactions in the highly enriched uranium. In addition, MCNP calculations of the delayed neutron build-up were compared with the measured data. Based on the results obtained from this dissertation, this measurement method has the potential to be expanded to include mass determinations of highly enriched uranium. Although many safeguards techniques exist for measuring special nuclear material, the number of assays that can be used to confirm HEU in shielded systems is

  16. Spectrographic determination of impurities in enriched uranium solutions

    International Nuclear Information System (INIS)

    Capdevila, C.; Roca, M.

    1980-01-01

    A spectrographic procedure for the determination of trace amounts of Al, B, Ba, Be, Bi, Ca, Cd, Co, Cr, Cu, Fe, K, L i , Hg, Mn, Mo, Na, Nb, Ni, P, Pb, Ru, Sb, Sn, Sr, Ti, V, Zn, and Zr in enriched uranyl nitrate solutions from the reprocessing of spent nuclear fuels is described. After removal of uranium by either TBP or TNOA solvent extraction, the aqueous phase Is analysed by the graphite spark technique. TBP is adequate for all impurities, excepting boron and phosphorus; both of these elements can sat is factory be determined by using TNOA after the addition of mannitol to avoid boron losses. (Author) 4 refs

  17. Report of the Uranium Enrichment Technology Evaluation Committee

    International Nuclear Information System (INIS)

    1977-01-01

    The centrifuge method of uranium enrichment is being pushed forward in Japan as a national project. The results of the technology evaluation are described regarding the efforts heretofore, the plan for pilot plant construction, internationally competitive ability, and the schedule after the pilot plant. In conclusion, the basic technology for proceeding to a pilot plant is now mature, and the pilot plant planning by PNC (Power Reactor and Nuclear Fuel Development Corporation) is generally appropriate. The outlook for constructing and internationally competitive practical plant is reasonably good. (Mori, K.)

  18. Evaluation of economical at a uranium enrichment demonstration plant

    International Nuclear Information System (INIS)

    Sugitsue, Noritake

    2001-01-01

    In this report, the economy of technical achievement apply in the uranium enrichment demonstration plant is evaluated. From the evaluation, it can be concluded that the expected purpose was achieved because there was a definite economic prospect to commercial plant. The benefit analysis of thirteen years operation of the uranium enrichment demonstration plant also provides a financial aspect of the uranium enrichment business. Therefore, the performance, price and reliability of the centrifuge is an important factor in the uranium enrichment business. And the continuous development of a centrifuge while considering balance with the development cost is necessary for the business in the future. (author)

  19. 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.)

  20. 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)

  1. Contribution of CERCA to the US DOE conference on the use of 20% and 45% enriched uranium as fuel for research reactors [contributed by Ph. Dewez and J. Doumerc, CERCA

    Energy Technology Data Exchange (ETDEWEB)

    Dewez, Ph; Doumerc, J [CERCA, Paris (France)

    1993-08-01

    CERCA (Compagnie pour l'Etude et la Realisation de Combustibles Atomiques) is a private French Company that was set up more than 20 years ago, in 1957. The head office of our Company is located in Paris. We have an industrial center at Bonneuil-sur-Marne near Paris, and a research and production center at Romans, between the cities of Lyon and Marseille in southern France. Throughout its existence, CERCA has many times brought a significant contribution to the design and manufacturing procedures of fuels of all kinds: - graphite-gas - heavy water, gas cooled (El-4) - heavy water, water cooled (El-3 previous tube and snow crystal designs) - light water - high temperatures and also sophisticated control rods, fuel followers, etc. Among a wide variety of other type of fuel elements CERCA has been involved since 1960 in MTR fuel element production and operates in Romans a production unit capable of 15 to 20,000 fuel plates per year. The backing of our strong R and D department enabled not only to follow the regularly increasing demand for improved characteristics by the reactor designers and users, but also to remain always a little bit ahead of these requirements in order to be able at all times to face new demands. The present international concern about non-proliferation of weapons-grade enriched uranium is now focusing interest on fuel elements with high total uranium contents. The current performance level of our products, that was made possible by our previous R and D programs, is such that we will experience no problems in keeping many low or medium power research reactors in operation. About one year ago, CERCA decided to initiate a program of technological development, to face the problem of the reduction of available enrichment.

  2. RERTR program progress in qualifying reduced-enrichment fuels

    International Nuclear Information System (INIS)

    Snelgrove, James L.

    1983-01-01

    In order to provide the technical means for reducing the enrichment of uranium used to fuel research and test reactors, the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program has been engaged in the development and testing of higher-uranium-density fuels than had been used previously. This fuel development effort included work to increase the density of fuels which were being used at the time the Program began and work on a fuel with the potential for much higher density. The ultimate goal of the fuel development and testing phase of the Program is to 'qualify' the fuel for use. A fuel is considered qualified when a sufficient data base for the fuel exists that it can be approved by regulating bodies for use in reactors. To convert a core to the use of reduced-enrichment fuel it is necessary to show that the core will behave properly during normal and off-normal operating conditions and to show that the fuel will behave properly to a reasonable margin beyond the conditions expected during normal operation. It is this latter area that this paper will address. The main characteristics to be considered in evaluating the performance of a fuel are its swelling, its blister-threshold temperature, and its metallurgical appearance. Data for the qualification of the reduced-enrichment fuels being developed by the RERTR Program are obtained from examination of miniature fuel plates (miniplates) which successfully pass the irradiation screening tests and from examinations of full-sized fuel elements. This paper will summarize the miniplate data reported in other papers presented during this meeting and will give the status of full-sized element irradiations. Finally, the current status of qualification of the various fuel types will be discussed and some projections of the future will be given

  3. 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.

  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. Uranium enrichment by laser: a technology for the future

    International Nuclear Information System (INIS)

    Cazalet, J.

    1999-01-01

    The SILVA (Isotopic Separation by Laser on atomic Vapor of uranium) process, developed by CEA and COGEMA, is an innovative system of production of enriched uranium, to be used as the fuel of nuclear reactors. It is a sound research program, calling on advanced technologies that are quickly changing. The goal is to cut drastically the production cost in comparison with the operating cost of the present plants based on gaseous diffusion. its industrialization is forecast for the beginning of next century. The SILVA process consists in putting a vapor of uranium through a beam of photons emitted by finely tuned lasers capable of ionising selectively the isotopes 235. The ionised isotopes are attracted on plates by an electric field, they are condensed and collected on these plates. The very high selectivity of enrichment technologies by laser, which are quite new, pave the way for compact and modular plants, which will consume little energy. Accordingly their production cost will be very low. So a new process could take a significant part of the uranium enrichment market after 2010. Even if the multinational EURODIF gaseous diffusion plant is modern and performing, it will be necessary to strengthen the French industry of uranium enrichment to maintain or improve its competitive position on the world market. This could be achieved by smoothly replacing EURODIF by a high performance laser plant. This is the common goal of CEA and COGEMA: all the efforts are concentrated on SILVA, the qualities of which (high selectivity, separation in one single step) have been demonstrated in the facilities of Saclay and Pierrelatte. 400 researchers and technicians are involved, as well as many industrial firms. The budget is equally by CEA and COGEMA through a cooperation agreement. The program includes: a phase of scientific and technical research, which has been highlighted in 1997-1998 by a demonstration of feasibility of the process; a phase of technological development, with

  6. 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

  7. Integrated design of SIGMA uranium enrichment plants

    International Nuclear Information System (INIS)

    Rivarola, Martin E.; Brasnarof, Daniel O.

    1999-01-01

    In the present work, we describe a preliminary analysis of the design feedbacks in a Uranium Enrichment Plant, using the SIGMA concept. Starting from the result of this analysis, a computer code has been generated, which allows finding the optimal configurations of plants, for a fixed production rate. The computer code developed includes the model of the Thermohydraulic loop of a SIGMA module. The model contains numerical calculations of the main components of the circuit. During the calculations, the main components are dimensioned, for a posterior cost compute. The program also makes an estimation of the enrichment gain of the porous membrane, for each separation stage. Once the dimensions of the main components are known, using the enrichment cascade calculation, the capital and operation costs of the plant could be determined. At this point it is simple to calculate a leveled cost of the Separative Work Unit (SWU). A numerical optimizer is also included in the program. This optimizer finds the optimal cascade configuration, for a given set of design parameters. The whole-integrated program permits to investigate in detail the feedback in the component design. Therefore, the sensibility of the more relevant parameters can be computed, with respect of the economical variables of the plant. (author)

  8. Low-resolution gamma-ray measurements of uranium enrichment

    International Nuclear Information System (INIS)

    Sprinkle, J.K. Jr.; Christiansen, A.; Cole, R.; Collins, M.L.

    1996-01-01

    Facilities that process special nuclear material perform periodic inventories. In bulk facilities that process low-enriched uranium, these inventories and their audits are based primarily on weight and enrichment measurements. Enrichment measurements determine the 211 U weight fraction of the uranium compound from the passive gamma-ray emissions of the sample. Both international inspectors and facility operators rely on the capability to make in-field gamma-ray measurements of uranium enrichment. These users require rapid, portable measurement capability. Some in-field measurements have been biased, forcing the inspectors to resort to high-resolution measurements or mass spectrometry to accomplish their goals

  9. Uranium enrichment decontamination and decommissioning fund

    International Nuclear Information System (INIS)

    1994-01-01

    One of the most challenging issues facing the Department of Energy's Office of Environmental Management is the cleanup of the three gaseous diffusion plants. In October 1992, Congress passed the Energy Policy Act of 1992 and established the Uranium Enrichment Decontamination and Decommissioning Fund to accomplish this task. This mission is being undertaken in an environmentally and financially responsible way by: devising cost-effective technical solutions; producing realistic life-cycle cost estimates, based on practical assumptions and thorough analysis; generating coherent long-term plans which are based on risk assessments, land use, and input from stakeholders; and, showing near-term progress in the cleanup of the gaseous diffusion facilities at Oak Ridge

  10. Chemical uranium enrichment with ion exchanger

    International Nuclear Information System (INIS)

    Takeda, Kunihiko; Onitsuka, Hatsuki; Obanawa, Heiichiro

    1991-01-01

    The uranium enrichment by using ion-exchange has been studied and developed since 1972. The ion-exchange rate has been improved approx. 3000 times and the electron exchange reaction, which occurs with ion-exchange reaction, was also accelerated with catalyst. Flow disturbance in a ion-exchange column has been fully studied and the value of turbulence has been reduced to 150μm. These results allowed us to design a very fine separation column, in which about 10000 stages can be obtained even when the column is more than 1 m in diameter. In the course of the development, a self-regenerating reaction between the redox agents was discovered and incorporated into the process, and has resulted in a reduction of 70 % in the separation energy requirement. (author)

  11. 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.

  12. 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.)

  13. A 2000-2010 years outlook of isotopic uranium enrichment

    International Nuclear Information System (INIS)

    Vasaru, G.

    1998-01-01

    The increase of the installed power in nuclear plants implies the following steps to be achieved: - developing a parallel industry for the nuclear fuel cycle able to ensure a rhythmic supply of natural uranium, possibly an isotopic enrichment of 235 U of around 1.2 - 3.2%, depending on the reactor system; - manufacturing the fuel elements and the operation of cycle back-end, which may, possibly, include a temporary storage of the irradiated fuel; - reprocessing the spend fuel; - radioactive waste processing in view of final disposal, as well as the recovery of un-spent uranium and of plutonium formed. The heavy water reactors of CANDU-PHW does not imply any isotopic enrichment but provides a lower burnup of only 7,000 MW day/tone. An enrichment to 1.2% in 235 U for this type of reactors could increase the burnup up to 20,000 MW day/tone. An advanced method of enriching 235 U is based on the Atomic Vapor Laser Isotop Separation (AVLIS). This procedure called AVLIS has several advantages which are pointed out in this paper, among which: a very high selectivity; high separation factors; a low energy consumption due to the fact that in the conditions of a selective photo ionization, the energy necessary to the process is only 6.2 eV for the separated 235 U atom vs 0.3 MeV in case of inertial separators or 3 MeV in case of gaseous diffusion procedure. With the current laser yields an energy consumption of 100 kWh/SWU is estimated for AVLIS procedures as compared with 2,400 kWh/SWU in case of gaseous diffusion; an almost entire extraction of 235 U, what ensures a more efficient utilisation of nuclear fuel. Due to its modular character and to potential improvement in the equipment which could be achieved, this procedure will ensure a reduction in the investment costs in the construction stage what will make AVLIS a substitute of the classical separation procedures

  14. 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.

  15. Uranium exploration, mining and ore enrichment techniques

    International Nuclear Information System (INIS)

    Fuchs, H.D.; Wentzlau, D.

    1985-01-01

    The paper describes the different types of uranium deposits and their importance. It is shown that during the present depressed uranium market situation, mainly high grade deposits such as unconformity-related deposits can be mined economically. The different successive exploration steps are outlined including methods used for uranium. Uranium mining does not greatly differ from normal mining, but the uranium metallurgy needs its own specialized but already classic technology. Only a relative small amount of uranium can be expected from projects where uranium is produced by in situ leach methods or by extraction from phosphoric acid. A short summary of investment costs and operating costs is given for an average uranium mine. The last chapter deals with the definition of different reserve categories and outlines the uranium reserves of the western world including the uranium production (1983) and the expected uranium production capacity for 1985 and 1990. (orig.) [de

  16. 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.

  17. Alpha spectrometry enriched uranium urinalysis results from IPEN

    International Nuclear Information System (INIS)

    Lima, Marina Ferreira

    2008-01-01

    Full text: IPEN (Instituto de Pesquisas Energeticas e Nucleares) manufactures the nuclear fuel to its research reactor, the IEA-R1. The CCN (Centro do Ciclo do Combustivel) facility produces the fuel cermets from UF 6 (uranium hexafluoride) enriched to 19.75% in 235 U. The production involves the transformation of the gaseous form in oxides and silicates by ceramic and metallurgical processing. The workers act in more than one step that involves exposition to types F, S and M compounds of uranium. Until 2003, only fluorimetric analysis was carried out by the LRT (Laboratorio de Radiotoxicologia - IPEN) in order to evaluate the intake of uranium, in spite of the sub estimation of the 234 U contribution to the internal doses. Isotopic uranium determination in urine by alpha spectrometry is the current method to monitoring the contribution of 234 U, 235 U and 238 U. Alpha spectrometry data of 164 samples from 84 individuals separate in three categories of workers: routinely work group; special operation group and control group - were analyzed how the isotopic composition excreted by urinary tract corresponds with the level of enrichment and isotopic composition of the plant products. Results show that is hard to estimate these intakes of 234 U and 235 U since these isotopes alpha activities are below the limit of detection or minimum detectable activity (MAD) of this method in the most part of the samples. Only in 22 samples it was possibly to measure the three radionuclides. Not expected high contribution of 234 U activity was found in samples of the control group. No one result over the 234 U and 235 U MAD was found in the samples from the special operation group. Only in 5 samples from the routinely group the levels of 235 U was higher than the levels of others groups. In a complementary study, 3 solid samples of UF 6 , U 2 O 8 and U 3 Si 2 from CCN plant were analyzed to determinate the isotopic uranium composition in these salts, since this composition varies

  18. 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)

  19. 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

  20. Open / Unipede 1995 survey on uranium and enrichment requirements and supplies in Western Europe

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    This report describes the current global situation with respect to uranium and enrichment supply and demand in Europe, as well as future prospects, based upon a survey conducted in mid-1995 by Unipede and Open among their member companies. The nuclear fuel cycle committees of Unipede and Open maintain an interest in all aspects of the nuclear fuel cycle from uranium procurement up to final disposal of radioactive waste. Regarding the front-end of the fuel cycle, the work within those committees currently concentrates of uranium and enrichment supply and demand in Europe, including the recycling of uranium and of plutonium recovered from reprocessing. For the past fifteen years, the nuclear fuel cycle committees of Unipede and Open have jointly conducted an annual survey among their members about uranium and enrichment requirements and supplies. The 1995 survey involved all nuclear electric utilities in Belgium, Finland, France, Germany, the Netherlands, Spain, Sweden, Switzerland and the United Kingdom. Exclusively electric utilities are surveyed, i.e. excluding national procurement organizations, traders, brokers, financial institutions, etc. The data obtained from the individual utilities in the same format, are aggregated and form the basis of the report. The quality of these data is guaranteed by the fact that the respondents are generally those people who are responsible for the day-to-day management of the nuclear fuel cycle in their company

  1. The world market-situation for uranium and its enrichment

    International Nuclear Information System (INIS)

    Lurf, G.

    1977-01-01

    The development of the uranium market is described as well as all pertinent facts which may have contributed to the strong rise in uranium prices of the past three years. The policies of countries which may in the future become major uranium exporters are discussed. For the conversion of uranium there is sufficient capacity. However, if construction of new plants is not started soon shortages could occur in the early 80ies. The market for enrichment has characterized in past years by substantial overcapacities. If new enrichment plants are constructed according to present schedules this overcapacity may prevail into the early 90ies. (orig.) [de

  2. Equations of state for enriched uranium and uranium alloy to 3500 MPa

    International Nuclear Information System (INIS)

    Bai Chaomao; Hai Yuying; Liu Jenlong; Li Zhenrong

    1990-04-01

    The volume compressions of 6 kinds of cast materials including enriched uranium, poor uranium, U-0.57 wt% Ti, U-0.33 wt% Nb, U-2.85 wt% Nb and U-7.5 wt% Nb-3.3 wt% Zr have been determined by monitoring piston displacements in a piston cylinder apparatus with double strengthening rings to 3500 MPa at room temperature. The dilation of the cylinder vessel and the press deformation were corrected by some experiments. The calculational data free from using the standard sample closed with used standard sample. The volume compressions of enriched uranium and poor uranium are nearly coincident. Pure uranium is more compressible than uranium alloys. These values of enriched uranium are in close agreement with values of Bridgman's pure uranium. The fitting coefficients of Bridgman's polynomial and Anderson's equation of state and isothermal bulk modules for the above materials are given

  3. Determination of uranium enrichment by using gamma-spectrometric methods

    International Nuclear Information System (INIS)

    Kutnyj, D.V.; Telegin, Yu.N.; Odejchuk, N.P.; Mikhailov, V.A.; Tovkanets, V.E.

    2009-01-01

    By using commercial analysis programs MGAU (LLNL, USA) and FRAM (LANL, USA) the summary error of gamma-spectrometric uranium enrichment measurements was investigated. Uranium samples with enrichments of 0,71; 4,46 and 20,1 % were measured. The coaxial high purity germanium detector (type GC) and the planar germanium detector (type LEGe) were used as gamma-radiation detectors. It was shown that experimental equipment and mathematical software available in NSC KIPT allow us to measure uranium enrichment by nondestructive method with accuracy of not worse than 2%.

  4. 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

  5. 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

  6. Fuel enrichment reduction for heavy water moderated research reactors

    International Nuclear Information System (INIS)

    McCulloch, D.B.

    1984-01-01

    Twelve heavy-water-moderated research reactors of significant power level (5 MW to 125 MW) currently operate in a number of countries, and use highly enriched uranium (HEU) fuel. Most of these reactors could in principle be converted to use uranium of lower enrichment, subject in some cases to the successful development and demonstration of new fuel materials and/or fuel element designs. It is, however, generally accepted as desirable that existing fuel element geometry be retained unaltered to minimise the capital costs and licensing difficulties associated with enrichment conversion. The high flux Australian reactor, HIFAR, at Lucas Heights, Sydney is one of 5 Dido-class reactors in the above group. It operates at 10 MW using 80% 235 U HEU fuel. Theoretical studies of neutronic, thermohydraulic and operational aspects of converting HIFAR to use fuels of reduced enrichment have been made over a period. It is concluded that with no change of fuel element geometry and no penalty in the present HEU fuel cycle burn-up performance, conversion to MEU (nominally 45% 235 U) would be feasible within the limits of current fully qualified U-Al fuel materials technology. There would be no significant, adverse effects on safety-related parameters (e.g. reactivity coefficients) and only small penalties in reactor flux. Conversion to LEU (nominally 20% 235 U) a similar basis would require that fuel materials of about 2.3 g U cm -3 be fully qualified, and would depress the in-core thermal neutron flux by about 15 per cent relative to HEU fuelling. In qualitative terms, similar conclusions would be expected to hold for a majority of the above heavy water moderated reactors. (author)

  7. 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.

  8. 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.

  9. Developments on uranium enrichment processes in France

    International Nuclear Information System (INIS)

    Frejacques, C.; Gelee, M.; Massignon, D.; Plurien, P.

    1977-01-01

    Gaseous diffusion has so far been the main source of supply for enriched uranium and it is only recently that the gas centrifuge came into the picture. Numerous other isotope separation processes have been considered or are being assessed, and there is nothing to exclude the future use of a new process. Developments on likely new processes have been carried out by many organizations both governmental and private. The French Commissariat a l'energie atomique, besides their very extensive endeavours already devoted to gaseous diffusion, have studied and developed the gas centrifuge, chemical exchange, aerodynamic and selective photoexcitation processes. The gaseous diffusion process, selected by Eurodif for the Tricastin plant, and which will also be used by Coredif, is discussed in another paper in these Proceedings. This process is the technico-economic yardstick on which our comparisons are based. Within the limits of their development level, processes are compared on the basis of the separative work cost components: specific investment, specific power consumption and power cost, and specific operating and maintenance costs. (author)

  10. Development of uranium enrichment technology by gas centrifugation

    International Nuclear Information System (INIS)

    Sibata, Tomofumi; Kai, Tsunetoshi

    1996-01-01

    The development of a gas-centrifuge for uranium enrichment has been conducted by Power Reactor and Nuclear Fuel Development Corporation in Japan after the first several years' fruitless works, the R and D works came to the point and continuing rapid improvements of centrifuges have started, Cascade tests were given with C-1 and C-2 cascade experimental facilities. Life, reliability and feasibility tests were given with the pilot plant and the demonstration plant. As a result of these works, the private commercial plant has started the operation. Although the main efforts were devoted to the development of metal rotor centrifuges in the course mentioned above, composite material rotor centrifuges have also been developed in parallel to achieve higher performance. Promising results have been being obtained with cascade test facilities on the pilot plant scale. Furthermore, R and D works are being proceeded on more excellent and advanced centrifuges. (author)

  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. The assisting system for uranium enrichment plant operation

    International Nuclear Information System (INIS)

    Nakazawa, Hiroaki; Yamamoto, Fumio

    1990-01-01

    We have been developing an operation assisting system, partially supported by AI system, for uranium enrichment plant. The AI system is a proto-type system aiming a final one which can be applied to any future large uranium enrichment plant and also not only to specific operational area but also to complex and multi-phenomenon operational area. An existing AI system, for example facility diagnostic system that utilizes the result of CCT analysis as knowledge base, has weakness in flexibility and potentiality. To build AI system, we have developed the most suitable knowledge representations using deep knowledge for each facility or operation of uranium enrichment plant. This paper describes our AI proto-type system adopting several knowledge representations that can represent an uranium enrichment plant's operation with deep knowledge. (author)

  13. Review of uranium enrichment prospects in Canada, 1976

    International Nuclear Information System (INIS)

    Developments since 1971 which affect the prospects for uranium enrichment in Canada from the federal government point of view are reviewed. The market for enriched uranium to the year 2000 is similar to that projected in 1971. The committed enrichment capacity of the world will be sufficient until 1990. The Canadian uranium mining capability may be adequate to supply an enrichment plant, but the present reserves policy along with the currently known resources are likely to restrict exports of its products during the plant life. Prices for enriched uranium produced in Canada would be higher than those reported by other proposed new plants; however, newer enrichment techniques have some potential for cost reductions. Application of enrichment with U235 (or plutonium and U233/thorium) to CANDU offers some uranium resource conservation and possible slight power cost reductions. Construction of an enrichment plant in Canada to supply the export market is less attractive in 1976 than in 1971, but there is potential for such a business in the future. (L.L.)

  14. Experimental determination of lattice parameters for 2% enriched uranium heavy water reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N; Takac, S; Markovic, H; Bosevski, T [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1963-04-15

    Systematic measurements of the buckling, infinite multiplication factor and the thermal utilization factor were made on a series of lattices for 2% enriched uranium tubular fuel elements in heavy water. This work represents the first phase of experimental verification of standard theoretical methods used for the determination of reactor parameters.

  15. Minimizing civilian use of highly enriched uranium - FRM II and global developments

    Energy Technology Data Exchange (ETDEWEB)

    Englert, Matthias [Oeko-Institut e.V., Darmstadt (Germany)

    2016-07-01

    The need to use highly enriched uranium (HEU) in civil nuclear applications is shrinking due to international efforts worldwide in the last three decades. Today low enriched uranium (LEU) that is not suitable for nuclear weapon purposes can be used instead in almost all civil applications. An overview of the current HEU use worldwide will be presented before focusing more on the use of HEU in research reactors and the conversion of existing reactors to LEU. Specifically interesting is the case of the German research reactor in Munich, the FRM-II. The reactor operates since ten years after intense national and international discussions over the use of weapon usable HEU to fuel the reactor. Since its construction the reactor is therefore obliged to convert to lower enrichment levels as soon as a suitable fuel becomes available. Despite huge international efforts to develop new fuels it is still not clear if and when the reactor can be converted.

  16. Apparatus for enrichment of uranium by double photoionization

    International Nuclear Information System (INIS)

    Laude, J.P.

    1983-11-01

    The present invention concerns enrichment of uranium by double photoionization. The use of a beam from a dye laser for excitation of gaseous uranium is known and the present invention concerns an apparatus of this type. The purpose of the invention is essentially to produce an apparatus having high energy efficiency. This is achieved according to the invention by using a continuous wave laser

  17. 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

  18. U.S. Non-proliferation policy and programs regarding use of high-enriched uranium in research reactors

    International Nuclear Information System (INIS)

    Lewis, R.A.

    1993-01-01

    Uranium enriched to 90-93%, supplied by the U.S., is now used in 141 research and test reactors in 35 countries around the world with a cumulative power of 1714 mw. Since of the order of 3 kg of 235 U is involved annually in fuel fabrication, fresh fuel transport and storage, reactor operation, and spent fuel cooling and return per megawatt of research reactor power, it is estimated that more than 5000 kg of very high-enriched uranium is handled each year to operate these reactors. Recent U.S. assessments have led to the tentative conclusion that in only approximately 11 of these reactors, generally those of highest power or power density, is the use of 90-93% enriched uranium currently a technical necessity. Universal use of the best state-of-the-art fuel technology would permit an estimated 90 of these reactors to use 20% enriched fuel, and estimated 40 others to use 45% enriched fuel, without significant performance degradation. If advanced research reactor fuel development programs currently under way in the U.S. and elsewhere are successful, it may, in fact, be possible to operate virtually all of these reactors on less than 20% enriched uranium in the longer term. The physical and economic practicality of these developmental fuels must, of course, await future assessments

  19. 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)

  20. Preliminary study of the economics of enriching PWR fuel with a fusion hybrid reactor

    International Nuclear Information System (INIS)

    Kelly, J.L.

    1978-09-01

    This study is a comparison of the economics of enriching uranium oxide for pressurized water reactor (PWR) power plant fuel using a fusion hybrid reactor versus the present isotopic enrichment process. The conclusion is that privately owned hybrid fusion reactors, which simultaneously produce electrical power and enrich fuel, are competitive with the gaseous diffusion enrichment process if spent PWR fuel rods are reenriched without refabrication. Analysis of irradiation damage effects should be performed to determine if the fuel rod cladding can withstand the additional irradiation in the hybrid and second PWR power cycle. The cost competitiveness shown by this initial study clearly justifies further investigations

  1. 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.

  2. HIGHLY ENRICHED URANIUM BLEND DOWN PROGRAM AT THE SAVANNAH RIVER SITE PRESENT AND FUTURE

    International Nuclear Information System (INIS)

    Magoulas, V; Charles Goergen, C; Ronald Oprea, R

    2008-01-01

    The Department of Energy (DOE) and Tennessee Valley Authority (TVA) entered into an Interagency Agreement to transfer approximately 40 metric tons of highly enriched uranium (HEU) to TVA for conversion to fuel for the Browns Ferry Nuclear Power Plant. Savannah River Site (SRS) inventories included a significant amount of this material, which resulted from processing spent fuel and surplus materials. The HEU is blended with natural uranium (NU) to low enriched uranium (LEU) with a 4.95% 235U isotopic content and shipped as solution to the TVA vendor. The HEU Blend Down Project provided the upgrades needed to achieve the product throughput and purity required and provided loading facilities. The first blending to low enriched uranium (LEU) took place in March 2003 with the initial shipment to the TVA vendor in July 2003. The SRS Shipments have continued on a regular schedule without any major issues for the past 5 years and are due to complete in September 2008. The HEU Blend program is now looking to continue its success by dispositioning an additional approximately 21 MTU of HEU material as part of the SRS Enriched Uranium Disposition Project

  3. Expansion of U. S. uranium enrichment capacity. Final environmental statement

    Energy Technology Data Exchange (ETDEWEB)

    1976-04-01

    Reasonably foreseeable environmental, social, economic, and technological costs and benefits of postulated expansion of U. S. enrichment capacity through the year 2000 and reasonably available alternatives to such expansion are described. Both the gas centrifuge and gaseous diffusion methods for the enrichment of uranium are considered in this impact assessment. (JGB)

  4. Development of an On-Line Uranium Enrichment Monitor

    International Nuclear Information System (INIS)

    Xuesheng, L.; Guorong, L.; Yonggang, Z.; Xueyuan, H. X.-Y.

    2015-01-01

    An on-line enrichment monitor was developed to measure the enrichment of UF6 flowing through the processing pipes in centrifuge uranium enrichment plant. A NaI(Tl) detector was used to measure the count rates of the 186 keV gamma ray emitted from 235U, and the total quantity of uranium was determined from thermodynamic characteristics of gaseous uranium hexafluoride. The results show that the maximum relative standard deviation is less than 1% when the measurement time is 120 s or more and the pressure is more than 2 kPa in the measurement chamber. There are two working models for the monitor. The monitor works normally in the continuous model, When the gas's pressure in the pipe fluctuates greatly, it can work in the intermittent model, and the measurement result is very well. The background of the monitor can be measured automatically periodically. It can control automatically electromagnetic valves open and close, so as to change the gas's quantity in the chamber. It is a kind of unattended and remote monitor, all of data can be transfer to central control room. It should be effective for nuclear materials accountability verifications and materials balance verification at uranium enrichment plant by using the monitor to monitor Uranium enrichment of gaseous uranium hexafluoride in the output end of cascade continuously. (author)

  5. 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)

  6. Control of uranium hazards - Portsmouth uranium enrichment plant

    International Nuclear Information System (INIS)

    Wagner, E.R.

    1985-01-01

    This report summarizes the Environmental, Safety and Health programs to control uranium hazards at the Portsmouth Gaseous Diffusion Plant. A description of the physical plant, the facility processes and the attendant uranium flows and effluents are provided. The hazards of uranium are discussed and the control systems are outlined. Finally, the monitoring programs are described and summaries of recent data are provided. 11 figs., 20 tabs

  7. Configuration of LWR fuel enrichment or burnup yielding maximum power

    International Nuclear Information System (INIS)

    Bartosek, V.; Zalesky, K.

    1976-01-01

    An analysis is given of the spatial distribution of fuel burnup and enrichment in a light-water lattice of given dimensions with slightly enriched uranium, at which the maximum output is achieved. It is based on the spatial solution of neutron flux using a one-group diffusion model in which linear dependence may be expected of the fission cross section and the material buckling parameter on the fuel burnup and enrichment. Two problem constraints are considered, i.e., the neutron flux value and the specific output value. For the former the optimum core configuration remains qualitatively unchanged for any reflector thickness, for the latter the cases of a reactor with and without reflector must be distinguished. (Z.M.)

  8. A committee report on the state-of-art of uranium isotope enrichment measurement techniques

    International Nuclear Information System (INIS)

    1982-06-01

    The activity of a sectional meeting (June, 1979 - February, 1982) for uranium enrichment measurement under the Committee on Analytical Chemistry of Nuclear Fuels and Reactor Materials is summarized. In the first part of this report, the object and present state of the measurement at the following organization are described; two development organizations (Power Reactor and Nuclear Fuel Development Corporation and Japan Nuclear Fuel Development), two nuclear fuel makers (Japan Nuclear Fuel and Mitsubishi Nuclear Fuel), one safeguards inspection organization (Nuclear Material Control Center) and one research organization (Japan Atomic Energy Research Institute). The second part contains the principle and technique of several measurement methods for uranium isotopic assay, such as mass spectrometry, passive and active assays and optical spectral method. Lastly, the concept of the reference materials and its practical information for mass spectrometry and non-destructive assay are described. (author)

  9. The gas centrifuge, uranium enrichment and nuclear proliferation

    International Nuclear Information System (INIS)

    Chapman, A.

    1988-01-01

    The author considers the consequences for controlling nuclear proliferation of the emergence of the gas centrifuge method for enriching uranium and succeeds in the difficult and delicate task of saying enough about gas centrifuge techniques for readers to judge, what may be involved in fully embracing gas centrifuge enrichment within the constraints of an anti-proliferation strategy, whilst at the same time saying nothing that could be construed as encouraging an interest in the gas centrifuge route to highly enriched uranium where none had before existed. (author)

  10. Highly enriched uranium, a dangerous substance that should be eliminated

    Energy Technology Data Exchange (ETDEWEB)

    Schaper, Annette

    2013-07-01

    Either highly enriched uranium (HEU) or plutonium is needed to construct a nuclear weapon. While plutonium is radioactive and hazardous in handling, HEU is far less dangerous. Furthermore, it is more difficult to detect by technical means. Therefore, in comparison to plutonium, HEU is much easier to divert, smuggle and hide. Moreover, a crude nuclear explosive made of HEU can be constructed in a much simpler way than one made using plutonium. For these reasons, HEU is the material most wanted by terrorists. A few tens of kilograms are sufficient for one explosive, but the quantities existing in the world add up to hundreds of tons. Due to the disarmament at the end of the Cold War, the NPT nuclear weapon states possess large quantities of HEU in excess of their needs for nuclear weapons. Therefore, these countries have not produced HEU for many years. Several international projects are working towards reducing the proliferation risks posed by HEU. The projects include the reduction of existing HEU by converting it to civilian reactor fuel that cannot be easily used for nuclear weapons. Other projects work towards reducing the number of countries and sites where HEU is stored by transferring it back to the countries of origin. And there are yet other projects which seek to minimize uses which would require new production of HEU. An international non-proliferation goal should be to eliminate all uses of HEU and thus to eliminate the need for any future production. Uses of HEU other than for nuclear weapons are as fuel in civilian research reactors, as base material for the production of special isotopes used in medical diagnostics, so-called medical targets and as fuel in military naval reactors. It is desirable to replace the HEU in all these applications with other materials and thus cease all HEU production forever. Use as fuel in civilian reactors has been greatly reduced during the last few decades. Within an international campaign, the Reduced Enrichment for

  11. Highly enriched uranium, a dangerous substance that should be eliminated

    International Nuclear Information System (INIS)

    Schaper, Annette

    2013-01-01

    Either highly enriched uranium (HEU) or plutonium is needed to construct a nuclear weapon. While plutonium is radioactive and hazardous in handling, HEU is far less dangerous. Furthermore, it is more difficult to detect by technical means. Therefore, in comparison to plutonium, HEU is much easier to divert, smuggle and hide. Moreover, a crude nuclear explosive made of HEU can be constructed in a much simpler way than one made using plutonium. For these reasons, HEU is the material most wanted by terrorists. A few tens of kilograms are sufficient for one explosive, but the quantities existing in the world add up to hundreds of tons. Due to the disarmament at the end of the Cold War, the NPT nuclear weapon states possess large quantities of HEU in excess of their needs for nuclear weapons. Therefore, these countries have not produced HEU for many years. Several international projects are working towards reducing the proliferation risks posed by HEU. The projects include the reduction of existing HEU by converting it to civilian reactor fuel that cannot be easily used for nuclear weapons. Other projects work towards reducing the number of countries and sites where HEU is stored by transferring it back to the countries of origin. And there are yet other projects which seek to minimize uses which would require new production of HEU. An international non-proliferation goal should be to eliminate all uses of HEU and thus to eliminate the need for any future production. Uses of HEU other than for nuclear weapons are as fuel in civilian research reactors, as base material for the production of special isotopes used in medical diagnostics, so-called medical targets and as fuel in military naval reactors. It is desirable to replace the HEU in all these applications with other materials and thus cease all HEU production forever. Use as fuel in civilian reactors has been greatly reduced during the last few decades. Within an international campaign, the Reduced Enrichment for

  12. Analysis of the performance of fuel cells PWR with a single enrichment and radial distribution of enrichments

    International Nuclear Information System (INIS)

    Vargas, S.; Gonzalez, J. A.; Alonso, G.; Del Valle, E.; Xolocostli M, J. V.

    2008-01-01

    One of the main challenges in the design of fuel assemblies is the efficient use of uranium achieving burnt homogeneous of the fuel rods as well as the burnt maximum possible of the same ones to the unload. In the case of the assemblies type PWR has been decided actually for fuel assemblies with a single radial enrichment. The present work has like effect to show the because of this decision, reason why a comparison of the neutronic performance of two fuel cells takes place with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The results shown in the present study of the behavior of the neutron flow as well as the power distribution through of assembly sustain the because of a single radial enrichment. (Author)

  13. 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

  14. 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)

  15. Enrichment plants. A survey of major new uranium enriching projects

    International Nuclear Information System (INIS)

    Kovan, D.

    1976-01-01

    The work enrichment situation is reported. The development of enrichment in the U.S. and in Europe is outlined. A brief description is given of the technology of separation by diffusion and by centrifugation and the advantages and disadvantages of the two processes are compared. Finally the supply and demand situation is briefly considered. (U.K.)

  16. Technical and economic aspects of new gaseous diffusion uranium enrichment capacity

    International Nuclear Information System (INIS)

    Langley, R.A. Jr.; O'Donnell, A.J.

    1977-01-01

    Work is well advanced on design and construction of the next major increment of U.S. uranium enrichment capacity. The plant will use the gaseous diffusion process to provide the required capacity and reliability at a competitive enrichment services cost. Gaseous diffusion technology is the base against which other processes are compared in order to assess their commercial viability. While it has generally been described as a mature technology with limited future development potential, work on design of the new U.S. plant has resulted in major improvement in plant design with corresponding decreases in plant capacity and operating costs. The paper describes major technological advances incorporated into the new plant design and their impact on enrichment costs. These include the effects of: - advanced barrier technology; - tandem compressor drive systems; - optimization of number of equipment sizes; - single level plant design; - development of rapid power level change capability; - electrical system simplification; - plant arrangement and layout. Resulting capital costs and projected enrichment costs are summarized. Enrichment costs are placed in the context of total nuclear fuel cycle costs. Trade-offs between uranium feed material quantities and enrichment plant tails assays are described, and optimization of this aspect of the nuclear fuel cycle is discussed. The effect on enrichment plant characteristics is described. Flexibility and capability of the new U.S. enrichment plant to meet these changing optimization conditions are described

  17. 77 FR 51579 - Application for a License To Export High-Enriched Uranium

    Science.gov (United States)

    2012-08-24

    ... NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant.... Complex, July 30, 2012, August Uranium (93.35%). uranium-235 high-enriched 1, 2012, XSNM3726, 11006037. contained in 7.5 uranium in the kilograms uranium. form of broken metal to the Atomic Energy of Canada...

  18. 77 FR 14838 - General Electric-Hitachi Global Laser Enrichment LLC, Commercial Laser-Based Uranium Enrichment...

    Science.gov (United States)

    2012-03-13

    ... Laser Enrichment LLC, Commercial Laser-Based Uranium Enrichment Facility, Wilmington, North Carolina... a license to General Electric-Hitachi Global Laser Enrichment LLC (GLE or the applicant) to authorize construction of a laser-based uranium enrichment facility and possession and use of byproduct...

  19. Reduced enrichment fuel and its reactivity effects in the University Training Reactor Moata

    International Nuclear Information System (INIS)

    Wilson, D.J.

    1983-08-01

    Concern for nuclear proliferation is likely to preclude future supply of highly enriched uranium fuel for research reactors such as the University Training Reactor Moata. This study calculates the fuel densities necessary to maintain the reactivity per plate of the present high enrichment (90 per cent 235 U) fuel for a range of lower enrichments assuming that no geometry changes are allowed. The maximum uranium density for commercially available aluminium-type research reactor fuels is generally considered to be about 1.7 g cm -3 . With this density limitation, the minimum enrichment to maintain present reactivity per plate is about 35 per cent 235 U. For low enrichment (max. 20 per cent 235 U) fuel, the required U density is about 2.9 g cm -3 , which is beyond the expected range for UAl/sub x/-Al but within that projected for the longer term development and full qualification for U 3 O 8 -Al. Medium enrichment (nominally 45 per cent 235 U) Al/sub x/-Al would be entirely satisfactory as an immediate replacement fuel, requiring no modifications to the reactor and operating procedures, and minimal reappraisal of safety issues. Included in this study are calculations of the fuel coefficients at various enrichments, the effect of replacing standard fuel plates or complete elements with 45 per cent enriched fuel, and the reactivity to be gained by replacing 12-plate with 13-plate elements

  20. Study of correcting the effect of daughter age on determining 235U enrichment of fuel rods

    International Nuclear Information System (INIS)

    Deng Jingshan; Zhou Chengfang; Luo Minxuan; Liu Yun

    1997-01-01

    Gamma-ray passive technique is a very effective method to assay and determine 235 U enrichment of nuclear power plant fuel rods. There is a weakness in this passive method, i.e. only after the uranium isotope daughters of UO 2 pellets have reached to equilibrium with uranium parent, then the 235 U enrichment can be determined. This weakness greatly restricts the application of the method. A new two-peak and two-window technique is developed that can overcome the interference of uranium daughter decay in determining 235 U enrichment of nuclear fuel rods, and the results are very satisfactory. The new technique will play an important role in the gamma-ray passive technique for determining 235 U enrichment of fuel rods. This new technique also makes the gamma-ray passive method perfectly. (11 figs., 6 tabs.)

  1. Research reactor preparations for the air shipment of highly enriched uranium from Romania

    International Nuclear Information System (INIS)

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

    2010-01-01

    In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation (RF) for conversion to low enriched uranium (LEU). The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR-S research reactor at Magurele, Romania, to Ozersk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation for Atomic Energy Rosatom and the International Atomic Energy Agency (IAEA). Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel. (author)

  2. Material control and accounting requirements for uranium enrichment facilities

    International Nuclear Information System (INIS)

    Ting, P.

    1991-01-01

    This paper reports that the U.S. Nuclear Regulatory Commission has defined material control and accounting (MC and A) requirement for low-enriched uranium enrichment plants licensed under 10 CFR parts 40 and 70. Following detailed assessment of potential safeguards issues relevant to these facilities, a new MC and A rule was developed. The primary safeguards considerations are detection of the loss of special nuclear material, detection of clandestine production of special nuclear material of low strategic significance for unauthorized use or distribution, and detection of unauthorized production of uranium enriched to ≥10 wt % U-235. The primary safeguards concerns identified were the large absolute limit of error associated with the material balance closing, the inability to shutdown some uranium enrichment technologies to perform a cleanout inventory of the process system, and the flexibility of some of these technologies to produce higher enrichments. Unauthorized production scenarios were identified for some technologies that could circumvent the detection of the production and removal of 5 kilograms of U-235 as high-enriched uranium through conventional material control and accounting programs. Safeguards techniques, including the use of production and process control information, measurements, and technical surveillance, were identified to compensate for these concerns

  3. Measurement system analysis (MSA) of the isotopic ratio for uranium isotope enrichment process control

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Josue C. de; Barbosa, Rodrigo A.; Carnaval, Joao Paulo R., E-mail: josue@inb.gov.br, E-mail: rodrigobarbosa@inb.gov.br, E-mail: joaocarnaval@inb.gov.br [Industrias Nucleares do Brasil (INB), Rezende, RJ (Brazil)

    2013-07-01

    Currently, one of the stages in nuclear fuel cycle development is the process of uranium isotope enrichment, which will provide the amount of low enriched uranium for the nuclear fuel production to supply 100% Angra 1 and 20% Angra 2 demands. Determination of isotopic ration n({sup 235}U)/n({sup 238}U) in uranium hexafluoride (UF{sub 6} - used as process gas) is essential in order to control of enrichment process of isotopic separation by gaseous centrifugation cascades. The uranium hexafluoride process is performed by gas continuous feeding in separation unit which uses the centrifuge force principle, establishing a density gradient in a gas containing components of different molecular weights. The elemental separation effect occurs in a single ultracentrifuge that results in a partial separation of the feed in two fractions: an enriched on (product) and another depleted (waste) in the desired isotope ({sup 235}UF{sub 6}). Industrias Nucleares do Brasil (INB) has used quadrupole mass spectrometry (QMS) by electron impact (EI) to perform isotopic ratio n({sup 235}U)/n({sup 238}U) analysis in the process. The decision of adjustments and change te input variables are based on the results presented in these analysis. A study of stability, bias and linearity determination has been performed in order to evaluate the applied method, variations and systematic errors in the measurement system. The software used to analyze the techniques above was the Minitab 15. (author)

  4. Idaho National Engineering Laboratory materials in inventory natural and enriched uranium management and storage costs

    International Nuclear Information System (INIS)

    Nebeker, R.L.

    1995-11-01

    On July 13, 1994, the Office of Environmental Management (EM) was requested to develop a planning process that would result in management policies for dealing with nuclear materials in inventory. In response to this request, EM launched the Materials In Inventory (MIN) Initiative. A Headquarters Working Group was established to develop the broad policy framework for developing MIN management policies. MIN activities cover essentially all nuclear materials within the DOE complex, including such items as spent nuclear fuel, depleted uranium, plutonium, natural and enriched uranium, and other materials. In August 1995, a report discussing the natural and enriched uranium portion of the Initiative for the Idaho National Engineering Laboratory (INEL) was published. That report, 'Idaho National Engineering Laboratory Materials-in-Inventory, Natural and Enriched Uranium'.' identified MIN under the control of Lockheed Idaho Technologies Company at the INEL. Later, additional information related to the costs associated with the storage of MIN materials was requested to supplement this report. This report provides the cost information for storing, disposing, or consolidating the natural and enriched uranium portion of the MIN materials at the INEL. The information consists of eight specific tables which detail present management costs and estimated costs of future activities

  5. The US uranium and enrichment industries: their fall and rise?

    International Nuclear Information System (INIS)

    Sewell, P.G.

    1988-01-01

    Strong government influence, monopolistic practices, free market forces and market orientation to customer needs are the conflicting forces which have shaped the evolution of the uranium and the uranium enrichment industries in the United States. These same factors are likely to continue to dictate to a large extent the future for each of these industries. Both the uranium and the uranium enrichment industries in the USA enjoyed the benefits and suffered the consequences of a monopolistic environment until the dynamics of a free market became prevalent in the 1980s. This resulted in the deterioration of both industries with respect to market share, sales and supply capacity needs. The history and environment of the two industries, the road to recovery for both, and the status and scope of legal and legislative initiatives to address the problems of each industry, are reviewed. (author)

  6. 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

  7. 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)

  8. Uranium enrichment : global view and Brazilian perspectives

    International Nuclear Information System (INIS)

    Zouain, D.M.; Sakamoto, L.H.

    1981-12-01

    A global view of isotope enrichment involving a general description of process (technical-economical aspects and policy) and status in developing countries is done. An enrichment demand in function of the Brazilian Nuclear Program is evaluated, analyzing a probable market and a low market. The perspectives to attend this demand, are studied. (E.G.) [pt

  9. Recent developments in the United States uranium enrichment enterprise

    International Nuclear Information System (INIS)

    Longenecker, J.R.

    1987-01-01

    In the near term, DOE is reducing production costs at the gaseous diffusion plants (GDPs), and we've made significant progress already. GDP production costs are expected to decline even further in the near future. DOE is also negotiating new power contracts for the GDPs. The new power contracts, capital improvements, and the use of more unfirm power should reduce our GDP average cost of production to about $60/SWU in the 1990s. Significant technical progress on the Atomic Vapor Laser Isotope Separation (AVLIS) advanced enrichment technology has been made recently. The highlight has been a series of half-scale integrated enrichment experiments using the Laser Demonstration Facility and the Mars separator. We are also ready to initiate testing in the full-scale Separator Demonstration Facility, including a 100 hour run that will vaporize over 5 tons of uranium. DOE is developing plans to restructure the enterprise into a more businesslike entity. The key objective in 1987 is to work with Congress to advance the restructuring of the U.S. uranium enrichment enterprise, to assure its long term competitiveness. We hope to establish in law the charter, objectives, and goals for the restructured enterprise. DOE expects that the world price for enrichment services will continue to decrease in the future. There should be sufficient excess enrichment capacity in the future to assure that competition will be keen. Such a healthy, competitive, world enrichment market will be beneficial to both suppliers and consumers of uranium enrichment services. (J.P.N.)

  10. Process for enriching uranium from seawater

    International Nuclear Information System (INIS)

    Heitkamp, D.; Inden, P.

    1982-01-01

    In selective elutriation of uranium deposited on titanium oxide hydrate by carbonate solution, only uranium should be dissolved from the absorption material forming carbonate compounds, without the deposited ballast ions, above all of magnesium, calcium and sodium being elutriated. The uranium elutriation according to the invention is therefore carried out in the presence of these ballast ions in the same concentrations as those in seawater. The carbonate concentration can only be raised as far as the solubility product of the basic magnesium carbonate permits, so that magnesium remains in the solution, as well as carbonate, in the concentration present in seawater. One must accept the absence of calcium ions in the elutriation solution, as their solubility product with carbonate is considerably less than that for magnesium. (orig./PW) [de

  11. Transformations of highly enriched uranium into metal or oxide

    International Nuclear Information System (INIS)

    Nollet, P.; Sarrat, P.

    1964-01-01

    The enriched uranium workshops in Cadarache have a double purpose on the one hand to convert uranium hexafluoride into metal or oxide, and on the other hand to recover the uranium contained in scrap materials produced in the different metallurgical transformations. The principles that have been adopted for the design and safety of these workshops are reported. The nuclear safety is based on the geometrical limitations of the processing vessels. To establish the processes and the technology of these workshops, many studies have been made since 1960, some of which have led to original achievements. The uranium hexafluoride of high isotopic enrichment is converted either by injection of the gas into ammonia or by an original process of direct hydrogen reduction to uranium tetrafluoride. The uranium contained m uranium-zirconium metal scrap can be recovered by combustion with hydrogen chloride followed treatment of the uranium chloride by fluorine in order to obtain the uranium in the hexafluoride state. Recovery of the uranium contained m various scrap materials is obtained by a conventional refining process combustion of metallic scrap, nitric acid dissolution of the oxide, solvent purification by tributyl phosphate, ammonium diuranate precipitation, calcining, reduction and hydro fluorination into uranium tetrafluoride, bomb reduction by calcium and slag treatment. Two separate workshops operate along these lines one takes care of the uranium with an isotopic enrichment of up to 3 p. 100, the other handles the high enrichments. The handling of each step of this process, bearing in mind the necessity for nuclear safety, has raised some special technological problems and has led to the conception of new apparatus, in particular the roasting furnace for metal turnings, the nitric acid dissolution unit, the continuous precipitator and ever safe filter and dryer for ammonium diuranate, the reduction and hydro fluorination furnace and the slag recovery apparatus These are

  12. Uranium enrichment: heading for a cliff

    International Nuclear Information System (INIS)

    Norman, C.

    1987-01-01

    Thanks to drastic cost cutting in the past 2 years, US enrichment plants now have the lowest cost production in the world, but US prices are still higher than those of overseas competitors because the business is paying for past mistakes. The most serious difficulty is that the Department of Energy (DOE), which owns and operates the US enrichment enterprise, is paying more than $500 million a year to the Tennessee Valley Authority (TVA) for electricity it once thought it would need but no longer requires. Another is that billions of dollars were spent in the 1970s and early 1980s to build new capacity that is now not needed. As a result, the enrichment enterprise has accumulated a debt to the US Treasury that the General Accounting Office (GAO) estimates at $8.8 billion. This paper presents the background and current debate in Congress about the difficulties facing the enrichment industry. In the midst of this debate over the future of the enterprise, the development of the next generation of enrichment technology is being placed in jeopardy. Known as atomic vapor laser isotope separation, or AVLIS, the process was viewed as the key to the long-term competitiveness of US enrichment. As the federal deficit mounted, however, funding for the AVLIS program was cut back and the timetable was stretched out. The US enrichment program has reached the point at which Congress will be forced to make some politically difficult decisions

  13. 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

  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. Principles and techniques of uranium enrichment

    International Nuclear Information System (INIS)

    Frejacques, Claude; Mezin, Michel

    1975-01-01

    The main separation processes already used industrially or likely to be used before the end of century (gas diffusion, ultracentrifugation, laser, the nozzle process, a process developed in South Africa) are presented. Some data on the costs of the enrichment are clarified. The main characteristics of the enrichment market in which the Eurodif plant is called upon, on the expiration of five years, to take a foremost place are reported [fr

  16. An optimal sequence of the reload charge fuel enrichment to a reactor

    International Nuclear Information System (INIS)

    Sato, S.

    1975-01-01

    An optimal sequence of enrichment of the reload charge of a three regions PWR during its life has been determined by dynamic programming. The state of the reactor is specified by the burnup of the fuel in the three regions and their initial enrichments. Constraints were imposed on the power peaking factor, the maximum burnup, the length of each stage between refueling and the total life of the reactor. 'Central-scatter loading' was assumed at each reloading. The two group diffusion equations were solved by the modal method for the static calculations of the reactor. Otimization of enrichment of the reload charge was performed under several hypotheses on the variation of the costs of uranium, costs of enrichment and the plant factor during the reactor life. It was observed that the optimum enrichment of the reload fuel is influenced more by the cost of enrichment rather than plant factor or cost of uranium. (Author) [pt

  17. 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)

  18. Project and supply agreement. The text of the agreement of 15 January 1993 between the International Atomic Energy Agency and the Government of the Republic of Indonesia and the Government of the United States of America concerning the transfer of enriched uranium for materials test reactor fuel development

    International Nuclear Information System (INIS)

    1994-08-01

    The text of the Project and Supply Agreement, which was approved by the Agency's Board of Governors on 4 December 1992 and concluded on 15 January 1993 between the Agency and the Governments of the Republic of Indonesia and the United States of America for the transfer of enriched uranium for materials test reactor fuel development is reproduced herein for the information of all Members. The agreement entered into force on 15 January 1993, pursuant to Article XII.1

  19. 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

  20. Research reactors. Problems of fuel element enrichment reduction. Deliberations and comments

    International Nuclear Information System (INIS)

    1978-10-01

    This paper summarises the main data from the major research reactors in the Federal Republic of Germany utilising highly enriched uranium (HEU) and presently available fuel technology for their fuel elements. The required modification for an adaption of the fabrication to lower enriched fuel are considered as well as the consequences on reactor performance operation and licensing. On the basis of past experience with reactor modifications a rough estimate of 82 months is given for the conversion of a reactor to a modified type of fuel and of 70 months for a fuel test program. The conclusions reflect the own calculations and data from other papers submitted to INFCE-WG 8C

  1. 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

  2. The jet nozzle process for uranium 235 isotopic enrichment

    International Nuclear Information System (INIS)

    Jordan, I.; Umeda, K.; Brown, A.E.P.

    1979-01-01

    A general survey of the isotopic enrichment of Uranium - 235, principally by jet nozzle process, is made. Theoretical treatment of a single stage and cascade of separation stages of the above process with its development in Germany until 1976 is presented [pt

  3. Soviet supplies of enriched uranium to capitalist countries

    International Nuclear Information System (INIS)

    Valvoda, Z.

    1977-01-01

    The Soviet supplies of enriched uranium to the following capitalist countries are surveyed: Belgium, Finland, France, FRG, Austria, Spain, Sweden, and the United Kingdom. The time period, total supplies, average annual supplies, estimated average price per separation work unit, and the date of the conclusion of the contract are reported. (J.B.)

  4. Radiological safety aspects in the fabrication of mixed oxide fuel elements. [Derived working limits in air and water for plutonium, enriched uranium and their mixture

    Energy Technology Data Exchange (ETDEWEB)

    Krishnamurthi, T.N.; Janardhanan, S.; Soman, S.D. (Bhabha Atomic Research Centre, Bombay (India). Health Physics Div.)

    The problems of radiological safety in the fabrication of (U, Pu)O/sub 2/ fuel assemblies for fast reactors utilising high exposure plutonium are discussed. Derived working limits for plutonium as a function of the burn-up of RAPS (Rajasthan Atomic Power Station) fuel, external gamma and neutron exposures from feed product batches, finished fuel pins and assemblies are presented. Shielding requirements for the various glove box operations are also indicated. In general, high exposure plutonium handling calls for remote fabrication and automation at various stages would play a key role in minimising exposures to personnel in a large production plant.

  5. A confirmatory measurement technique for highly enriched uranium

    International Nuclear Information System (INIS)

    Sprinkle, J.K. Jr.

    1987-07-01

    This report describes a confirmatory measurement technique for measuring uranium items in their shipping containers. The measurement consists of a weight verification and the detection of three gamma rays. The weight can be determined very precisely, thus it severely constrains the options of the diverter who might want to imitate the gamma signal with a bogus item. The 185.7-keV gamma ray originates from 235 U, the 1001 keV originates from a daughter of 238 U, and the 2614 keV originates from a daughter of 232 U. These three gamma rays exhibit widely different attenuation properties, they correlate with enrichment and total uranium mass, and they rigorously discriminate against a likely diversion scenario (low-enriched uranium substitution). These four measured quantities, when combined, provide a signature that is very difficult to counterfeit

  6. 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.

  7. 10 CFR 40.33 - Issuance of a license for a uranium enrichment facility.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Issuance of a license for a uranium enrichment facility... License Applications § 40.33 Issuance of a license for a uranium enrichment facility. (a) The Commission... the licensing of the construction and operation of a uranium enrichment facility. The Commission will...

  8. 10 CFR 70.23a - Hearing required for uranium enrichment facility.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Hearing required for uranium enrichment facility. 70.23a... MATERIAL License Applications § 70.23a Hearing required for uranium enrichment facility. The Commission... license for construction and operation of a uranium enrichment facility. The Commission will publish...

  9. 77 FR 60482 - Regulatory Guide 5.67, Material Control and Accounting for Uranium Enrichment Facilities...

    Science.gov (United States)

    2012-10-03

    ... Accounting for Uranium Enrichment Facilities Authorized To Produce Special Nuclear Material of Low Strategic... Accounting for Uranium Enrichment Facilities Authorized to Produce Special Nuclear Material of Low Strategic... and is applicable to the Paducah GDP and other uranium enrichment facilities that have been licensed...

  10. A new era in U.S. uranium enrichment

    International Nuclear Information System (INIS)

    Longenecker, J.R.

    1984-01-01

    Complex market conditions, including lower than anticipated electrical growth rates, creation of a large spot market of enriched uranium, fluctuations in currency exchange rates, and certain political considerations, have created an unstable market for all primary producers, including the United States. In response to these conditions, the Department of Energy made significant changes to the U.S. program including the issuance of the Utility Services contract on January 18, 1984. Other major changes include redirecting research and development efforts on the advanced gas centrifuge and atomic vapor laser isotope separation processes, rescoping of the Gas Centrifuge Enrichment Plant project, and reevaluation of the operational mode of the three gaseous diffusion plants. Taken together, we believe these actions will retain the U.S. position of leadership in uranium enrichment. In summary, we plan to compete--through introduction of the world's most advanced, lowest cost technology and through responsiveness to our customers' needs

  11. Unallocated Off-Specification Highly Enriched Uranium: Recommendations for Disposition

    Energy Technology Data Exchange (ETDEWEB)

    Bridges, D. N.; Boeke, S. G.; Tousley, D. R.; Bickford, W.; Goergen, C.; Williams, W.; Hassler, M.; Nelson, T.; Keck, R.; Arbital, J.

    2002-02-27

    The U.S. Department of Energy (DOE) has made significant progress with regard to disposition planning for 174 metric tons (MTU) of surplus Highly Enriched Uranium (HEU). Approximately 55 MTU of this 174 MTU are ''offspec'' HEU. (''Off-spec'' signifies that the isotopic or chemical content of the material does not meet the American Society for Testing and Materials standards for commercial nuclear reactor fuel.) Approximately 33 of the 55 MTU have been allocated to off-spec commercial reactor fuel per an Interagency Agreement between DOE and the Tennessee Valley Authority (1). To determine disposition plans for the remaining {approx}22 MTU, the DOE National Nuclear Security Administration (NNSA) Office of Fissile Materials Disposition (OFMD) and the DOE Office of Environmental Management (EM) co-sponsored this technical study. This paper represents a synopsis of the formal technical report (NNSA/NN-0014). The {approx} 22 MTU of off-spec HEU inventory in this study were divided into two main groupings: one grouping with plutonium (Pu) contamination and one grouping without plutonium. This study identified and evaluated 26 potential paths for the disposition of this HEU using proven decision analysis tools. This selection process resulted in recommended and alternative disposition paths for each group of HEU. The evaluation and selection of these paths considered criteria such as technical maturity, programmatic issues, cost, schedule, and environment, safety and health compliance. The primary recommendations from the analysis are comprised of 7 different disposition paths. The study recommendations will serve as a technical basis for subsequent programmatic decisions as disposition of this HEU moves into the implementation phase.

  12. 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

  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. Strategic outlook for the US Dept. of Energy uranium enrichment enterprise

    International Nuclear Information System (INIS)

    Longenecker, J.R.

    1987-01-01

    Mounting problems in the early 1980s resulted in a business crisis for the US uranium enrichment enterprise. To address these problems, the US Dept. of Energy (DOE) developed and announced in 1984 a plan for reestablishing its competitive position in the world marketplace. This recovery plan has been very successful. The US DOE is now implementing the third phase of that strategy, institutional restructuring. Maintaining a healthy, stable, and competitive uranium enrichment capability in the US is a high-priority effort for DOE, whose uranium enrichment enterprise has two major goals: in the near term, DOE's goal is to provide reliable, competitive enrichment services to commercial and defense customers in a manner that assures a reasonable return on the government's investment. In the future, to assure adequate supplies of competitively priced nuclear fuel, DOE must adapt to the changing needs of the marketplace. Thus, we must stabilize and maintain the long-term vitality of the US enrichment enterprise by restructuring it and by providing the means by which the private sector can become involved in the business. Restructuring and private ownership obviously will require the approval of Congress. To achieve these goals, DOE has developed strategic objectives, which are briefly described

  15. Uranium enrichment: an activity for the riches

    International Nuclear Information System (INIS)

    Mundim, S.G.

    1985-01-01

    The United States in the first place, the European countries next, have begun to maintain strict control of the technology transfer of the methods of enrichment, helped largely by the International Atomic Energy Agency which creates additional obstacles with is so-called 'safeguards'. So that it can possess a complete energy cycle, Brazil needs to have access to the techniques of enrichment. The methods are many and still extremely costly. Committed to signing a risk contract with Germany in the scope of its nuclear treaty with the intention of implementing here the jetcentrofuge method, Brazil should, nevertheless, make an effort to cause to emerge from its laboratories a new procedure which would permit it to become independent in this sector. (Author) [pt

  16. 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

  17. Experience with a fuel rod enrichment scanner

    International Nuclear Information System (INIS)

    Kubik, R.N.; Pettus, W.G.

    1975-01-01

    This enrichment scanner views all fuel rods produced at B and W's Commercial Nuclear Fuel Plant. The scanner design is derived from the PAPAS System reported by R. A. Forster, H. D. Menlove, and their associates at Los Alamos. The spatial resolution of the system and smoothing of the data are discussed in detail. The cost-effectiveness of multi-detector versus single detector scanners of this general design is also discussed

  18. Feasibility of nondestructive assay measurements in uranium enrichment plants

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.B.

    1978-04-01

    Applications of nondestructive assay methods to measurement problems in uranium enrichment facilities are reviewed. The results of a number of test and evaluation projects that were performed over the last decade at ORGDP and Portsmouth are presented. Measurements of the residual holdup in the top enrichment portion of the shut-down K-25 cascade were made with portable neutron and gamma-ray detectors, and inventory estimates based on these data were in good agreement with ORGDP estimates. In the operating cascade, the tests showed that portable NaI detectors are effective for monitoring NaF and alumina media for gaseous effluent traps and that gas phase enrichments and inventories, as well as large deposits of uranium, can be detected with portable neutron and gamma-ray instrumentation. A wide variety of scrap and waste materials, including barrier and compressor blades, incinerator ash and trapping media, and miscellaneous waste, were measured using passive gamma-ray and neutron methods and 14-MeV neutron interrogation. Methods developed for rapid verification of UF/sub 6/ in shipping containers with portable neutron and gamma-ray instruments are now used routinely by safeguards inspectors. Passive assay methods can also be used to measure continuously the enrichments of /sup 235/U and /sup 234/U in the UF/sub 6/ product and tails withdrawals of a gaseous diffusion plant. A system that was developed and installed in the extended-range product withdrawal station of the Portsmouth facility measures enrichment with a relative accuracy of 0.5%. A stand-alone neutron detector has also been successfully evaluated for the measurement of the isotopic abundance of /sup 234/U in UF/sub 6/ in sample cylinders, an application of potential importance to Minor Isotope Safeguards Technology. Recommendations are made on the role of NDA measurements for enrichment plant safeguards, including additional tests and evaluations that may be needed, particularly for advanced uranium

  19. 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.

  20. Automated assay of uranium solution concentration and enrichment

    International Nuclear Information System (INIS)

    Horley, E.C.; Gainer, K.; Hansen, W.J.; Kelley, T.A.; Parker, J.L.; Sampson, T.E.; Walton, G.; Jones, S.A.

    1992-01-01

    For the first time, the concentration and enrichment of uranium solutions can be measured in one step. We have developed a new instrument to automatically measure the concentration and enrichment of uranium solutions through the adaptation of a commercial robot. Two identical solution enrichment systems are being installed in the Portsmouth Gaseous Diffusion Plant. These automated systems will reduce radiation exposure to personnel and increase the reliability and repeatability of the measurements. Each robotic system can process up to 40 batch and 8 priority samples in an unattended mode. Both passive gamma-ray and x-ray fluorescence (XRF) analyses are performed to determine total uranium concentration and 235 U enrichment. Coded samples are read by a bar-code reader to determine measurement requirements, then assayed by either or both of the gamma-ray and XRF instruments. The robot moves the sample containers and operates all shield doors and shutters, reducing hardware complexity. If the robots is out of service, an operator can manually perform all operations

  1. Inventory control through gamma spectrometry at the enriched uranium laboratory

    International Nuclear Information System (INIS)

    Vicens, H.E.; Korob, R.O.; Goldschmidt, A.E.

    1987-01-01

    The enriched uranium laboratory processes alternatively uranium 90% and 20% enriched in U-235. The control of the isotopic composition of lots is made through mass spectrometry. In the laboratory operation wastes of both enrichments are generated and the recovery is performed with a time delay. To strengthen the administrative controls, avoid errors related to personnel replacement and/or deferred operations, it seemed suitable to adjust the gamma spectrometry as a fast, simple and available method to determine the enrichment. The laboratory work includes a wet and a dry process. The waste recovery necessarily involves the handling of liquid samples. For this reason, it was decided to determine the calibration curve for uranyl nitrate samples of fixed concentration and geometry. The samples were prepared from material purified through double precipitation of uranium peroxide and subsequent ignition to U 3 O 8 in platinum crucible, in tubular oven during 8 hours at 720 deg C. The preparation of samples, the measurement description, the discussion of results and the analysis of errors due to the presence of insoluble material and concentration changes are included. (Author)

  2. Benchmark criticality experiments for fast fission configuration with high enriched nuclear fuel

    International Nuclear Information System (INIS)

    Sikorin, S.N.; Mandzik, S.G.; Polazau, S.A.; Hryharovich, T.K.; Damarad, Y.V.; Palahina, Y.A.

    2014-01-01

    Benchmark criticality experiments of fast heterogeneous configuration with high enriched uranium (HEU) nuclear fuel were performed using the 'Giacint' critical assembly of the Joint Institute for Power and Nuclear Research - Sosny (JIPNR-Sosny) of the National Academy of Sciences of Belarus. The critical assembly core comprised fuel assemblies without a casing for the 34.8 mm wrench. Fuel assemblies contain 19 fuel rods of two types. The first type is metal uranium fuel rods with 90% enrichment by U-235; the second one is dioxide uranium fuel rods with 36% enrichment by U-235. The total fuel rods length is 620 mm, and the active fuel length is 500 mm. The outer fuel rods diameter is 7 mm, the wall is 0.2 mm thick, and the fuel material diameter is 6.4 mm. The clad material is stainless steel. The side radial reflector: the inner layer of beryllium, and the outer layer of stainless steel. The top and bottom axial reflectors are of stainless steel. The analysis of the experimental results obtained from these benchmark experiments by developing detailed calculation models and performing simulations for the different experiments is presented. The sensitivity of the obtained results for the material specifications and the modeling details were examined. The analyses used the MCNP and MCU computer programs. This paper presents the experimental and analytical results. (authors)

  3. Gamma-ray measurements for uranium enrichment standards

    International Nuclear Information System (INIS)

    Reilly, T.D.

    1979-01-01

    The gamma-ray spectroscopic measurement of uranium enrichment is one of the most widely used nondestructive analysis techniques. A study has been started of the precision and accuracy achievable with this technique and the physical parameters which affect it. The study was prompted by questions raised during the ongoing ESARDA-NBS experiment to produce uranium oxide reference counting materials for the technique. Results reported using a high-quality Ge(Li) spectrometer system show reproducibility comparable to that attainable with mass spectrometry

  4. 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

  5. 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.

  6. Consequences of the new ICRP recommendations on uranium enrichment and uranium chemistry

    International Nuclear Information System (INIS)

    Bonnefoy-Claudet, J.

    1991-01-01

    From the first available information on the draft of new recommendations of the International Commission Radiological Protection, consequences should be very different depending upon industry type and handled products. That is to say: negligible for uranium enrichment by gaseous diffusion and important for future laser isotope separation techniques and for uranium chemistry especially for oxide treatment. This is enhanced when the products are coming from reprocessing [fr

  7. 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

  8. Application of a method to measure uranium enrichment without use of standards

    International Nuclear Information System (INIS)

    Saule, F.A.; Righetti, M.A.

    1998-01-01

    Full text: The determination of uranium enrichment in the many different stages present at a gaseous diffusion enrichment plant (diffusers, cisterns, deposits in pipes, drums with rests of process), or materials of deposit (plates of fuel elements not irradiated and recipients with uranium oxide), that have several geometries and physics properties of the containers, is very important for safeguards inspections. In this work is tested a non destructive analysis technique to determine the value of uranium enrichment of different samples with uranium materials without use of standards, to apply in safeguards inspections. It was used a hyper pure germanium detector with efficiency of 20% to obtain the gamma spectrum of the samples. In each spectrum, were used the net area values corresponding to four lines of U-235 (at 143, 163, 186 and 205 keV) and three lines of U-238 (258, 766 and 1001 keV); these values were analysed with two different methods. The comparison of the calculated and declared values showed a discrepancy of about 10%. (author) [es

  9. Contribution of CERCA to the US DOE conference on the use of 20% and 45% enriched uranium as fuel for research reactions. Part 2

    International Nuclear Information System (INIS)

    Doumerc, J.

    1993-01-01

    As far as prices are concerned, the author would like to state first that for the time being we have to compare the prices of a steady state situation, which is represented by a well known process that has been in use for many years, with a transitory situation which is the achievement of the same expertise in new, extrapolated fuels. That is why the comparison has to be corrected for the results within the next 2 to 3 years. Obviously, it seems that there are other factors which contribute to the price increases. I think there is a very significant example for this

  10. The development of lower enrichment fuels for Canadian research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A; Belanger, L; Grolway, C M [AECL, Atomic Energy of Canada Limited, Chalk River, ON (Canada); Foo, M T [CRNL, Combustion Engineering Superheater Ltd., Moncton, NB (Canada)

    1983-08-01

    As part of the world wide move to proliferation resistant fuels, new fuels which use reduced enrichment uranium are being developed for use in the NRX and NRU reactors. A fuel consisting of particles of a USiAl alloy dispersed in an Al matrix has been selected for development along with Al-37 wt% U alloy and Al-U{sub 3}O{sub 8} cermet as backup fuels. This report outlines the progress made in the development of the Al-USiAl and Al-37 wt% U. Results show that good quality extruded rods containing either fuel can be made with techniques similar to those used to fabricate the current NRX and NRU fuels. However, the new fuels will be more expensive to make. Although the oxidation behaviour of the Al-USiAl is not as good as that of the Al-U alloys, its corrosion behaviour in high temperature water does not seem much worse. The oxidation and aqueous corrosion of A-37 wt% U are not much different from those of the Al-U alloys currently used. (author)

  11. Disposal of wastes from uranium conversion and enrichment processes

    International Nuclear Information System (INIS)

    Costello, J.M.

    1981-11-01

    This paper reviews the general principles and objectives in radioactive waste management, and shows how these are applied in options for management and disposal of wastes from uranium upgrading operations. Some estimates of radiological dose commitments and health effects from LWR nuclear power and its fuel cycle have been made for US conditions

  12. 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)

  13. Important matter by confirmation of administrative office regarding repair of enriched uranium dissolution tanks in reprocessing plant of Power Reactor and Nuclear Fuel Development Corp

    International Nuclear Information System (INIS)

    1985-01-01

    The Nuclear Safety Commission acknowledged the policy of handling this matter by Science and Technology Agency after having received a report from the Committee on Examination of Nuclear Fuel Safety on April 11, 1985, and carried out the deliberation. The investigation and deliberation of this matter were instructed by the NSC to the Committee on January 24, 1985. It was confirmed that the repair welding applied to the place of leak of the dissolution tanks would not hinder the expected test dissolution, and if the leak occurs, the measures to detect it properly have been taken. In order to confirm the soundness of the repair welding, the Power Reactor and Nuclear Fuel Development Corp. is to carry out the test dissolution for about 400 hours per one tank dividing into three runs, and the observation of appearance is to be made after every run. The time of test dissolution, the items and contents of inspection were confirmed to be adequate. Moreover, the immersion corrosion test of test pieces and the long term corrosion test in a laboratory are to be carried out. (Kako, I.)

  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. 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)

  16. Enriched uranium workshops towards decommissioning CEA Cadarache

    International Nuclear Information System (INIS)

    Lisbonne, P.; Seisson, D.

    2008-01-01

    The author presents the post operational clean-out following the shut down of the ATUE nuclear facility implemented in the Cadarache research Center (France) and preceding the dismantling operations. ATUE was constructed in 1965, in the support of the fuel cycle industry, to develop the dry process of UF 6 to UO 2 conversion and the wet process with step of nitric acid solution, solvent purification by exchange columns, ammonium precipitation and hydrogen reduction. The dismantling organization, methods and technology are also presented and discussed. In a last part the author highlights some important points of the project. (A.L.B.)

  17. Strategies and planning for enriched uranium

    International Nuclear Information System (INIS)

    Villard, C.

    1991-01-01

    The author describes the main elements of Florida Power and Light (FPL) nuclear fuels procurement strategy: costs, security of supply, and quality. FPL's procurement strategy calls for the evaluation of all alternatives as follows: purchase materials and services separately or in various combinations; short or long term contracting; spot procurement; inventory purchasing for economic reasons, if market and long term contract prices are projected to rise much above the current market price; and, a combination of some or all of the above. At the same time, FPL would want to preserve enough flexibility and options to participate in the spot market and minimize its costs, should the market remain depressed

  18. Adaptive control theory of concentration in the uranium enrichment plant

    International Nuclear Information System (INIS)

    Sugitsue, Noritake; Miyagawa, Hiroshi; Yokoyama, Kaoru; Nakakura, Hiroyuki

    1999-01-01

    This paper presents the new adaptive control of concentration in the uranium enrichment plant. The purpose of this control system is average concentration control in production tram. As a result the accuracy and practical use of this control system have already been confirmed by the operation of the uranium enrichment demonstration plant. Three elements of technology are required to this method. The first is the measurement of the concentration using product flow quantity change. This technology shall be called 'Qp difference to Xp transform method'. The second is the relationship between temperature change and flow quantity using G.M.D.H. (Groupe Method of Data Handling) and the third is the estimation of temperature change using AR (Auto-regressive) model. (author)

  19. 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

  20. Historical review of CEA researches on uranium enrichment

    International Nuclear Information System (INIS)

    Camarcat, N.

    1997-01-01

    The various uranium enrichment processes that have been studied at the CEA since 1953 are briefly reviewed: gaseous diffusion (which led to the construction of EURODIF plant), chemical treatments (which were abandoned in 1988 for cost reasons), gaseous ultracentrifugation, electromagnetic processes, laser techniques (since 1980) and especially the SILVA technique (atomic vapour laser isotopic separation) which could take the place of the gaseous diffusion technique when the EURODIF plant will need to be renewed before 2010

  1. Radioactive effluents, Portsmouth Uranium Enrichment Complex, calendar year 1983

    International Nuclear Information System (INIS)

    Acox, T.A.; Klein, L.S.

    1984-03-01

    Radioactive discharges from the Portsmouth Uranium Enrichment Complex are discussed and tabulated. Tables indicate both the location of the discharge and the nuclides discharged. Routine discharges for 1983 are well below the Radioactive Concentration Guide limits specified in DOE Order 5480.1, Chapter XI. There was, however, an unplanned release in December from the X-326 Building Side Purge which exceeded the limits. 1 figure

  2. An automated solution enrichment system for uranium analysis

    International Nuclear Information System (INIS)

    Jones, S.A.; Sparks, R.; Sampson, T.; Parker, J.; Horley, E.; Kelly, T.

    1993-01-01

    An automated Solution Enrichment system (SES) for analysis of Uranium and U-235 isotopes in process samples has been developed through a joint effort between Los Alamos National Laboratory and Martin Marietta Energy systems, Portsmouth Gaseous Diffusion Plant. This device features an advanced robotics system which in conjuction with stabilized passive gamma-ray and X-ray fluorescence detectors provides for rapid, non-destructive analyses of process samples for improved special nuclear material accountability and process control

  3. Prompt neutron decay constant for the Oak Ridge Research Reactor with 20 wt % 235U enriched fuel

    International Nuclear Information System (INIS)

    Ragan, G.E.; Mihalczo, J.T.

    1986-01-01

    This paper describes measurements of the prompt neutron decay constant at delayed criticality for the Oak Ridge Research Reactor (ORR) using 20 wt % 235 U enriched fuel and compares these measurements with similar measurements using 93.2 wt % 235 U enriched fuel. This reactor parameter is of interest because it affects the transient behavior of the reactor in prompt criticality accident situations. This experiment is part of a program to investigate the differences in the performance of research reactors fueled with highly enriched and low enriched uranium. The prompt neutron decay constants were obtained using noise analysis measurement techniques for a core with newly fabricated, unirradiated fuel elements

  4. 77 FR 33253 - Regulatory Guide 8.24, Revision 2, Health Physics Surveys During Enriched Uranium-235 Processing...

    Science.gov (United States)

    2012-06-05

    ... NUCLEAR REGULATORY COMMISSION [NRC-2010-0115] Regulatory Guide 8.24, Revision 2, Health Physics..., ``Health Physics Surveys During Enriched Uranium-235 Processing and Fuel Fabrication'' was issued with a... specifically with the following aspects of an acceptable occupational health physics program that are closely...

  5. Use of Savannah River Site facilities for blend down of highly enriched uranium

    International Nuclear Information System (INIS)

    Bickford, W.E.; McKibben, J.M.

    1994-02-01

    Westinghouse Savannah River Company was asked to assess the use of existing Savannah River Site (SRS) facilities for the conversion of highly enriched uranium (HEU) to low enriched uranium (LEU). The purpose was to eliminate the weapons potential for such material. Blending HEU with existing supplies of depleted uranium (DU) would produce material with less than 5% U-235 content for use in commercial nuclear reactors. The request indicated that as much as 500 to 1,000 MT of HEU would be available for conversion over a 20-year period. Existing facilities at the SRS are capable of producing LEU in the form of uranium trioxide (UO 3 ) powder, uranyl nitrate [UO 2 (NO 3 ) 2 ] solution, or metal. Additional processing, and additional facilities, would be required to convert the LEU to uranium dioxide (UO 2 ) or uranium hexafluoride (UF 3 ), the normal inputs for commercial fuel fabrication. This study's scope does not include the cost for new conversion facilities. However, the low estimated cost per kilogram of blending HEU to LEU in SRS facilities indicates that even with fees for any additional conversion to UO 2 or UF 6 , blend-down would still provide a product significantly below the spot market price for LEU from traditional enrichment services. The body of the report develops a number of possible facility/process combinations for SRS. The primary conclusion of this study is that SRS has facilities available that are capable of satisfying the goals of a national program to blend HEU to below 5% U-235. This preliminary assessment concludes that several facility/process options appear cost-effective. Finally, SRS is a secure DOE site with all requisite security and safeguard programs, personnel skills, nuclear criticality safety controls, accountability programs, and supporting infrastructure to handle large quantities of special nuclear materials (SNM)

  6. 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

  7. 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

  8. Uranium enrichment by the gaseous diffusion process

    International Nuclear Information System (INIS)

    Petit, J.F.

    1977-01-01

    After a brief description of the process and technology (principle, stage constitution, cascade constitution, and description of a plant), the author gives the history of gaseous diffusion and describes the existing facilities. Among the different enrichment processes contemplated in the USA during and after the last world war, gaseous diffusion has been the only one to be developed industrially on a wide scale in the USA, then in the UK, in the USSR and in France. The large existing capacities in the USA provided the country with a good starting base for the development of a light-water nuclear power plant programme, the success of which led to a shortfall in production means. France and the USA, possessing the necessary know-how, have been able to undertake the realization of two industrial programmes: the CIP-CUP programme in the USA and the Eurodif project in France. Current plans still call for the construction of several plants by 1990. Can the gaseous diffusion process meet this challenge. Technically, there is no doubt about it. Economically, this process is mainly characterized by large energy consumption and the necessity to build large plants. This leads to a large investment, at least for the first plant. Other processes have been developed with a view to reducing both energy and capital needs. However, in spite of continuous studies and technological progress, no process has yet proved competitive. Large increments in capacities are still expected to come from gaseous diffusion, and several projects taking into account the improvements in flexibility, automatization, reliability and reduced investment, are analysed in the paper. Combining new facilities with existing plants has already proved to be of great interest. This situation explains why gaseous diffusion is being further investigated and new processes are being studied. (author)

  9. MUICYCL and MUIFAP: models tracking minor uranium isotopes in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Blum, S.R.; McLaren, R.A.

    1979-10-01

    Two computer programs have been written to provide information on the buildup of minor uranium isotopes in the nuclear fuel cycle. The Minor Uranium Isotope Cycle Program, MUICYCL, tracks fuel through a multiyear campaign cycle of enrichment, reactor burnup, reprocessing, enrichment, etc. MUICYCL facilities include preproduction stockpiles, U 235 escalation, and calculation of losses. The Minor Uranium Isotope Flowsheet Analyzer Program, MUIFAP, analyzes one minor isotope in one year of an enrichment operation. The formulation of the enrichment cascade, reactors, and reprocessing facility is presented. Input and output descriptions and sample cases are presented. The programs themselves are documented by short descriptions of each routine, flowcharts, definitions of common blocks and variables, and internal documentation. The programs are written in FORTRAN for use in batch mode

  10. 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

  11. 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).

  12. Past and present supply of enriched uranium for research reactors in the European Union

    International Nuclear Information System (INIS)

    Mueller, H.

    2002-01-01

    In the last decade research reactor operators have focused mainly on the issues of disposal of spent research reactor fuel and the development of high density fuels. The safe supply of fresh uranium did not receive as much attention. This is surprising since the United States - who was the main supplier for LEU and HEU since the late 1950's - stopped supplying non-US research reactors with enriched uranium a decade ago. The reason for this stop of supply is described in this paper. This paper explains how research reactors in the E U continued to operate during the last decade, in spite of the fact that their primary supply source had not provided LEU and HEU over the same period. (author)

  13. 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)

  14. 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

  15. 78 FR 16303 - Request To Amend a License To Export; High-Enriched Uranium

    Science.gov (United States)

    2013-03-14

    ... NUCLEAR REGULATORY COMMISSION Request To Amend a License To Export; High-Enriched Uranium Pursuant to 10 CFR 110.70 (b) ``Public Notice of Receipt of an Application,'' please take notice that the... Application No. Docket No. U.S. Department of Energy, High-Enriched Uranium 10 kilograms uranium To...

  16. 78 FR 17942 - Request To Amend a License To Export High-Enriched Uranium

    Science.gov (United States)

    2013-03-25

    ... NUCLEAR REGULATORY COMMISSION Request To Amend a License To Export High-Enriched Uranium Pursuant... Administration. Enriched Uranium contained in 99.7 Reactor in the be processed for March 6, 2013 (93.35%)) kilograms Czech Republic to medical isotope March 11, 2013 uranium) the list of production at the XSNM3622...

  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. 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

  19. The uranium enrichment market and long-term technological options

    International Nuclear Information System (INIS)

    Schneider-Maunoury, A.

    1992-01-01

    The world enrichment market situation is clearly delineated up to the year 2000. Including the East European countries, worldwide enriched uranium requirements should reach 40 million separative work units (SWUs) a year and production capacity should reach 44 millions SWUs. Two-thirds of this capacity will be supplied by the gaseous diffusion process and one-third by the centrifuge process. The enrichment processes currently considered are: (i) the gaseous diffusion process, (ii) the centrifuge process, (iii) the chemical treatment process and (iv) the laser processes, long-term assessment of the enrichment market up to the year 2015. Two scenarios may be envisioned for the (i) Public opinion will continue to block the development of nuclear power, and requirements will level off at 40 million SWUs. (ii) Changing attitudes will favor a reasonable approach enabling a revival of nuclear power expansion around 1995. Requirements should then increase starting in 2005 and would readily attain 60 million SWUs a year by 2015. Depending on market conditions, enrichment process options will be influenced either entirely by cost considerations, without allowance for the time factor, or by need to meet demand. Demonstrations of the industrial validity of laser processes are expected by 1992 - 1995 and, if interest in nuclear power makes a comeback, decisions should be made between 1995 and 2000 to build new large-capacity enrichment plants. The gaseous diffusion process may still be used for a long time if nuclear power is judiciously employed. The centrifuge process will be fully mature by the year 2000. The uranium vapor laser processes offer the most promise and should ultimately prevail. the chemical processes, though outsiders, deserve watching. (author)

  20. A Non-Proliferating Fuel Cycle: No Enrichment, Reprocessing or Accessible Spent Fuel - 12375

    Energy Technology Data Exchange (ETDEWEB)

    Parker, Frank L. [Vanderbilt University (United States)

    2012-07-01

    Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storage sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded

  1. 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

  2. 78 FR 60928 - Request To Amend a License To Export High-Enriched Uranium

    Science.gov (United States)

    2013-10-02

    ... NUCLEAR REGULATORY COMMISSION Request To Amend a License To Export High-Enriched Uranium Pursuant... manufacture HEU The Netherlands. National Nuclear Security Uranium uranium (17.1 targets in France... export from 9.4 kg of U-235 contained in 10.1 kg uranium to a new cumulative total of 17.1 kg of U-235...

  3. 77 FR 73056 - Application for a License To Export High-Enriched Uranium

    Science.gov (United States)

    2012-12-07

    ... NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant... Complex. Uranium (93.2%). uranium-235 at CERCA AREVA Romans October 10, 2012 contained in 6.2 in France and to October 12, 2012 kilograms irradiate targets at XSNM3729 uranium. the BR-2 Research 11006053...

  4. 77 FR 73055 - Application for a License To Export High-Enriched Uranium

    Science.gov (United States)

    2012-12-07

    ... NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant.... Security Complex. Uranium uranium-235 at CERCA AREVA October 10, 2012 (93.35%). contained in Romans in France October 12, 2012 10.1 kilograms and to irradiate XSNM3730 uranium. targets at the HFR 11006054...

  5. 78 FR 33448 - Application for a License To Export High-Enriched Uranium

    Science.gov (United States)

    2013-06-04

    ... NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant.... Security Complex, May 13, Uranium (93.35%). uranium-235 at the National 2013, May 21, 2013, XSNM3745, contained in 7.5 Research Universal 11006098. kilograms reactor in Canada for uranium. ultimate use in...

  6. 78 FR 72123 - Request To Amend a License to Export High-Enriched Uranium

    Science.gov (United States)

    2013-12-02

    ... NUCLEAR REGULATORY COMMISSION Request To Amend a License to Export High-Enriched Uranium Pursuant... manufacture HEU targets in Belgium. National Nuclear Security Uranium (HEU) uranium France for irradiation in... 5.8 kg of U- 235 contained in 6.2 kg uranium to a new cumulative total of 12.615 kg of U-235...

  7. 10 CFR 140.13b - Amount of liability insurance required for uranium enrichment facilities.

    Science.gov (United States)

    2010-01-01

    ... enrichment facilities. 140.13b Section 140.13b Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) FINANCIAL... required for uranium enrichment facilities. Each holder of a license issued under Parts 40 or 70 of this chapter for a uranium enrichment facility that involves the use of source material or special nuclear...

  8. 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

  9. 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.

  10. 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)

  11. 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

  12. 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

  13. ZPR-3 Assembly 11: A cylindrical sssembly of highly enriched uranium and depleted uranium with an average 235U enrichment of 12 atom % and a depleted uranium reflector

    International Nuclear Information System (INIS)

    Lell, R.M.; McKnight, R.D.; Tsiboulia, A.; Rozhikhin, Y.

    2010-01-01

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was 235 U or 239 Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core 235 U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation Working Group (CSEWG) Benchmark

  14. 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)

  15. 77 FR 13367 - General Electric-Hitachi Global Laser Enrichment, LLC, Proposed Laser-Based Uranium Enrichment...

    Science.gov (United States)

    2012-03-06

    ... NUCLEAR REGULATORY COMMISSION [NRC-2009-0157] General Electric-Hitachi Global Laser Enrichment, LLC, Proposed Laser-Based Uranium Enrichment Facility, Wilmington, NC AGENCY: Nuclear Regulatory... Impact Statement (EIS) for the proposed General Electric- Hitachi Global Laser Enrichment, LLC (GLE...

  16. Progress on the IPNS Enriched Uranium Booster Target

    International Nuclear Information System (INIS)

    Knox, A.E.; Carpenter, J.M.; Bailey, J.L.

    1986-09-01

    We describe the Enriched Uranium Booster Target designed for use in Argonne's Intense Pulsed Neutron Source. This report contains a general description of the system, and descriptions of the thermal-hydraulic and loss-of-coolant accident analyses, of the neutronic, criticality and power density calculations, of the assessment of radiation and thermal cycling growth, and of the disk fabrication methods. We also describe the calculations of radionuclide buildup and the related hazards analysis and our calculations of the temperature and stress profiles in the disks, and briefly allude to considerations of security and safeguards

  17. Projected uranium measurement uncertainties for the Gas Centrifuge Enrichment Plant

    International Nuclear Information System (INIS)

    Younkin, J.M.

    1979-02-01

    An analysis was made of the uncertainties associated with the measurements of the declared uranium streams in the Portsmouth Gas Centrifuge Enrichment Plant (GCEP). The total uncertainty for the GCEP is projected to be from 54 to 108 kg 235 U/year out of a measured total of 200,000 kg 235 U/year. The systematic component of uncertainty of the UF 6 streams is the largest and the dominant contributor to the total uncertainty. A possible scheme for reducing the total uncertainty is given

  18. Video monitoring system for enriched uranium casting furnaces

    International Nuclear Information System (INIS)

    Turner, P.C.

    1978-03-01

    A closed-circuit television (CCTV) system was developed to upgrade the remote-viewing capability on two oralloy (highly enriched uranium) casting furnaces in the Y-12 Plant. A silicon vidicon CCTV camera with a remotely controlled lens and infrared filtering was provided to yield a good-quality video presentation of the furnace crucible as the oralloy material is heated from 25 to 1300 0 C. Existing tube-type CCTV monochrome monitors were replaced with solid-state monitors to increase the system reliability

  19. Supply of the Federal Republic of Germany with enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    Oboussier, F [Commission of the European Communities, Brussels (Belgium). Supply Agency

    1976-08-01

    After giving a short historical background of past policies concerning uranium supply (policies which were mainly in the hands of the USAEC), the essential characteristics of the enrichment agreements with ERDA, URENCO, EURODIF, and the USSR are explained. A comparison is made difficult by completely different conditions. New side issues (the latest estimates on the development of electricity demand, delays in the licensing of nuclear power plants, etc.) lead to the assumption that demand will exceed supply by 1985/86, which means that the construction of new separation plants will be necessary.

  20. Highly Enriched Uranium Metal Cylinders Surrounded by Various Reflector Materials

    International Nuclear Information System (INIS)

    Bernard Jones; J. Blair Briggs; Leland Monteirth

    2007-01-01

    A series of experiments was performed at Los Alamos Scientific Laboratory in 1958 to determine critical masses of cylinders of Oralloy (Oy) reflected by a number of materials. The experiments were all performed on the Comet Universal Critical Assembly Machine, and consisted of discs of highly enriched uranium (93.3 wt.% 235U) reflected by half-inch and one-inch-thick cylindrical shells of various reflector materials. The experiments were performed by members of Group N-2, particularly K. W. Gallup, G. E. Hansen, H. C. Paxton, and R. H. White. This experiment was intended to ascertain critical masses for criticality safety purposes, as well as to compare neutron transport cross sections to those obtained from danger coefficient measurements with the Topsy Oralloy-Tuballoy reflected and Godiva unreflected critical assemblies. The reflector materials examined in this series of experiments are as follows: magnesium, titanium, aluminum, graphite, mild steel, nickel, copper, cobalt, molybdenum, natural uranium, tungsten, beryllium, aluminum oxide, molybdenum carbide, and polythene (polyethylene). Also included are two special configurations of composite beryllium and iron reflectors. Analyses were performed in which uncertainty associated with six different parameters was evaluated; namely, extrapolation to the uranium critical mass, uranium density, 235U enrichment, reflector density, reflector thickness, and reflector impurities. In addition to the idealizations made by the experimenters (removal of the platen and diaphragm), two simplifications were also made to the benchmark models that resulted in a small bias and additional uncertainty. First of all, since impurities in core and reflector materials are only estimated, they are not included in the benchmark models. Secondly, the room, support structure, and other possible surrounding equipment were not included in the model. Bias values that result from these two simplifications were determined and associated

  1. 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

  2. Nonproliferation analysis of the reduction of excess separated plutonium and high-enriched uranium

    International Nuclear Information System (INIS)

    Persiani, P.J.

    1995-01-01

    The purpose of this preliminary investigation is to explore alternatives and strategies aimed at the gradual reduction of the excess inventories of separated plutonium and high-enriched uranium (HEU) in the civilian nuclear power industry. The study attempts to establish a technical and economic basis to assist in the formation of alternative approaches consistent with nonproliferation and safeguards concerns. The analysis addresses several options in reducing the excess separated plutonium and HEU, and the consequences on nonproliferation and safeguards policy assessments resulting from the interacting synergistic effects between fuel cycle processes and isotopic signatures of nuclear materials

  3. Analysis of civilian processing programs in reduction of excess separated plutonium and high-enriched uranium

    International Nuclear Information System (INIS)

    Persiani, P.J.

    1995-01-01

    The purpose of this preliminary investigation is to explore alternatives and strategies aimed at the gradual reduction of the excess inventories of separated plutonium and high-enriched uranium (HEU) in the civilian nuclear power industry. The study attempts to establish a technical and economic basis to assist in the formation of alternative approaches consistent with nonproliferation and safeguards concerns. The analysis addresses several options in reducing the excess separated plutonium and HEU, and the consequences on nonproliferation and safeguards policy assessments resulting from the interacting synergistic effects between fuel cycle processes and isotopic signatures of nuclear materials

  4. Uranium enrichment services activity. Financial statements for fiscal year ended June 30, 1975

    International Nuclear Information System (INIS)

    1975-01-01

    Financial statements for the Uranium Enrichment Services Activity, covering both ERDA and its contractors, are presented to provide information concerning the operations and financial position of this operation

  5. Comments on proposed legislation to restructure DOE's uranium enrichment program

    International Nuclear Information System (INIS)

    1991-04-01

    This book focuses on H.R.145, H.R.788, and S.210. Each of the proposed bills would restructure DOE's enrichment program as a government corporation with private financing and would encourage the eventual sale of the corporation to the private sector. In doing so, the bills would, among other things, allow the corporation to set prices to maximize long-term returns; establish a fund to meet the costs of decontamination, decommissioning, and other environmental cleanup costs associated with uranium enrichment activities; transfer interest in DOE's new atomic vapor laser isotope separation (AVLIS) process to the new corporation; and, except for H.R. 145, require the government to pay its share of the costs to clean up mill tailings (mining wastes) generated under government contracts

  6. Natural uranium/conversion services/enrichment services

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    This article is the 1993 uranium market summary. During this reporting period, there were 50 deals in the concentrates market, 26 deals in the UF6 market, and 14 deals for enrichment services. In the concentrates market, the restricted value closed $0.15 higher at $9.85, and the unrestricted value closed down $0.65 at $7.00. In the UF6 market, restricted prices fluctuated and closed higher at $31.00, and unrestricted prices closed at their initial value of $24.75. The restricted transaction value closed at $10.25 and the unrestricted value closed at $7.15. In the enrichment services market, the restricted value moved steadily higher to close at $84.00 per SWU, and the unrestricted value closed at its initial value of $68.00 per SWU

  7. Isotopic analysis of uranium hexafluoride highly enriched in U-235

    International Nuclear Information System (INIS)

    Chaussy, L.; Boyer, R.

    1968-01-01

    Isotopic analysis of uranium in the form of the hexafluoride by mass-spectrometry gives gross results which are not very accurate. Using a linear interpolation method applied to two standards it is possible to correct for this inaccuracy as long as the isotopic concentrations are less than about 10 per cent in U-235. Above this level, the interpolations formula overestimates the results, especially if the enrichment of the analyzed samples is higher than 1.3 with respect to the standards. A formula is proposed for correcting the interpolation equation and for the extending its field of application to high values of the enrichment (≅2) and of the concentration. It is shown that by using this correction the results obtained have an accuracy which depends practically only on that of the standards, taking into account the dispersion in the measurements. (authors) [fr

  8. 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

  9. Airborne effluent control at fuel enrichment, conversion, and fabrication plants

    International Nuclear Information System (INIS)

    Mitchell, M.E.

    1976-01-01

    Uranium conversion, enrichment, and fuel fabrication facilities generate gaseous wastes that must be treated prior to being discharged to the atmosphere. Since all three process and/or handle similar compounds, they also encounter similar gaseous waste disposal problems, the majority of which are treated in a similar manner. Ventilation exhausts from personnel areas and equipment off-gases that do not contain corrosive gases (such as HF) are usually passed through roughening and/or HEPA filters prior to release. Ventilation exhausts that contain larger quantities of particles, such as the conversion facilities' U 3 O 8 sampling operation, are passed through bag filters or cyclone separators, while process off-gases containing corrosive materials are normally treated by sintered metal filters or scrubbers. The effectiveness of particle removal varies from about 90 percent for a scrubber alone to more than 99.9 percent for HEPA filters or a combination of the various filters and scrubbers. The removal of nitrogen compounds (N 2 , HNO 3 , NO/sub x/, and NH 3 ) is accomplished by scrubbers in the enrichment and fuel fabrication facilities. The conversion facility utilizes a nitric acid recovery facility for both pollution control and economic recovery of raw materials. Hydrogen removal from gaseous waste streams is generally achieved with burners. Three different systems are currently utilized by the conversion, enrichment, and fuel fabrication plants to remove gaseous fluorides from airborne effluents. The HF-rich streams, such as those emanating from the hydrofluorination and fluorine production operations of the conversion plant, are passed through condensers to recover aqueous hydrofluoric acid

  10. 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

  11. 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

  12. Mixed core management: Use of 93% and 72% enriched uranium in the BR2 reactor

    International Nuclear Information System (INIS)

    Ponsard, B.

    2000-01-01

    The BR2 reactor, put into operation in 1963 and refurbished from July 1995 till April 1997, is a 100 MW high-flux Materials Testing Reactor, using 93% 235 U enriched uranium as standard fuel, light water as coolant and beryllium as moderator. The present operating regime consists of five irradiation cycles per year at an operating power between 50 and 70 MW; each cycle is characterized by 21 days operation. In the framework of a 'qualification programme', six 72% 235 U fuel elements fabricated with uranium recovered from the reprocessing of BR2 spent fuel at UKAEA-Dounreay have been successfully irradiated in the period 1994-1995 reaching a maximum mean burnup of 48% without the release of fission products. Since 1998, this type of fuel element is irradiated routinely together with standard 93% 235 U fuel elements in order to optimize the utilization of the available HEU inventory. The purpose of this paper is to present the strategy developed in order to optimize the mixed core management of the BR2 reactor. (author)

  13. National acquisition of sufficient know-how to determine procurement of natural and enriched uranium

    International Nuclear Information System (INIS)

    Peix, J.; Rougeau, J.-P.

    1977-01-01

    Today, the guarantees of a natural uranium supply and of its enrichment, seem to be the necessary conditions for assuring the development of nuclear electric program. However, decisions cannot be made before further consideration of the consequences it will have on the whole service of the nuclear fuel cycle. Whether the supplies are immediate, or medium term, or long term, the lead-times of the fuel cycle and the present situation of the world energy supplies are such that a country committed to develop nuclear electric energy must equip itself quickly with the indispensable human potential to define its own supply policy and then, to assure a strict achievement up to its interests. Thus, without forecasting any sort of policy, the first matter is to study the problems arising from nuclear fuel cycle and to propose a minimal human and technical potential, indicating the different stages of its constitution. (author)

  14. Irradiation program of slightly enriched fuel elements at the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Casario, J.A.; Cesario, R.H.; Perez, R.A.; Sidelnik, J.I.

    1987-01-01

    An irradiation program of fuel elements with slightly enriched uranium is implemented, tending to the homogenization of core at Atucha I nuclear power plant. The main benefits of the enrichment program are: a) to extend the average discharge burnup of fuel elements, reducing the number of elements used to generate the same amount of energy. This implies a smaller annual consumption of elements and consequently the reduction of transport and replacement operations and of the storage pool systems as well as that of radioactive wastes; b) the saving of uranium and structural materials (Zircaloy and others). In the initial stage of program an homogeneous core enrichment of 0.85% by weight of U-235 is anticipated. The average discharge burnup of fuel elements, as estimated by previous studies, is approximately 11.6 MW d/kg U. The annual consumption of fuel elements is reduced from 396 of natural uranium to 205, with a load factor of 0.85. It is intended to reach the next equilibrium steps with an enrichment of 1.00 and 1.20% in U-235. (Author)

  15. Impact of UO{sub 2} Enrichment of Fuel Zoning Rods in Long Cycle Operation of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Ho Cheol; Lee, Deokjung [KHNP CRI, Daejeon (Korea, Republic of); Jeong, Eun; Choe, Jiwon [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    Extending the cycle length can not only increase the energy production, but also bring down outage costs by reducing the number of refueling outages during the lifetime of a nuclear power plant. It is reasonable that more fresh fuels are loaded for long cycle operation. However, minimizing the number of fresh fuels is essential in aspect of fuel economics. This can cause high power peaking near the water holes, due to increased thermalization of neutrons in those regions. To prevent this, special fuel zoning rods are used and surround the water holes. These rods use lower-enriched uranium (they have an enrichment rate lower than the other fuel rods). If we adjust the enrichment rate of fuel zoning rods, we can reduce power peaking and moreover increase cycle length. In this paper, we designed a core suitable for long cycle operation and we conducted sensitivity tests of fuel cycle length on UO2 enrichment rate in fuel zoning region in order to extend the cycle length while using the same number of fresh fuels. The correlations between the fuel zoning enrichment and cycle length, peaking factor, CBC and shutdown margin were analyzed. The more the enrichment rate in fuel zoning region increases, the more the fuel cycle length increases. At the same time, CBC, Fq and shutdown margin do not change significantly. Increasing the fuel zoning enrichment rate presents the right property of increasing the fuel cycle length without causing a large change to CBC, Fq and shutdown margin. In conclusion, by increasing the uranium enrichment rate in fuel zoning region, fuel cycle length can be increased and the safety margins can be maintained for long cycle operation of cores.

  16. Use of minor uranium isotope measurements as an aid in safeguarding a uranium enrichment cascade

    International Nuclear Information System (INIS)

    Levin, S.A.; Blumkin, S.; Von Halle, E.

    1979-01-01

    Surveillance and containment, which are indispensable supporting measures for material accountability, do not provide those charged with safeguarding an installation with the assurance beyond the shadow of a doubt that all the input and output uranium will in fact be measured. Those who are concerned with developing non-intrusive techniques for safeguarding uranium enrichment plants under the Nuclear Non-Proliferation Treaty have perceived the possibility that data on the minor uranium isotope concentrations in an enrichment cascade withdrawal and feed streams may provide a means either to corroborate or to contradict the material accountability results. A basic theoretical study has been conducted to determine whether complete isotopic measurements on enrichment cascade streams may be useful for safeguards purposes. The results of the calculations made to determine the behaviour of the minor uranium isotopes ( 234 U and 236 U) in separation cascades, and the results of three plant tests made to substantiate the validity of the calculations, are reviewed briefly. Based on the fact that the 234 U and 236 U concentrations relative to that of 235 U in cascade withdrawal streams reflect the cascade flow-sheet, the authors conclude that the use of the minor isotope concentration measurements (MIST) in cascade withdrawal streams is a potentially valuable adjunct to material accounting for safeguarding a 235 U enrichment cascade. A characteristic of MIST, which qualifies it particularly for safeguards application under the NPT, is the fact that its use is entirely non-intrusive with regard to process technology and proprietary information. The usefulness of MIST and how it may be applied are discussed briefly. (author)

  17. Preliminary uranium enrichment analysis results using cadmium zinc telluride detectors

    International Nuclear Information System (INIS)

    Lavietes, A.D.; McQuaid, J.H.; Paulus, T.J.

    1995-01-01

    Lawrence Livermore National Laboratory (LLNL) and EG ampersand G ORTEC have jointly developed a portable ambient-temperature detection system that can be used in a number of application scenarios. The detection system uses a planar cadmium zinc telluride (CZT) detector with custom-designed detector support electronics developed at LLNL and is based on the recently released MicroNOMAD multichannel analyzer (MCA) produced by ORTEC. Spectral analysis is performed using software developed at LLNL that was originally designed for use with high-purity germanium (HPGe) detector systems. In one application, the CZT detection system determines uranium enrichments ranging from less than 3% to over 75% to within accuracies of 20%. The analysis was performed using sample sizes of 200 g or larger and acquisition times of 30 min. The authors have demonstrated the capabilities of this system by analyzing the spectra gathered by the CZT detection system from uranium sources of several enrichments. These experiments demonstrate that current CZT detectors can, in some cases, approach performance criteria that were previously the exclusive domain of larger HPGe detector systems

  18. Permit issued for expansion of the Gronau uranium enrichment plant

    International Nuclear Information System (INIS)

    2005-01-01

    Over the next few years, the Gronau Uranium Enrichment Plant (UAG) will be able to raise its capacity in steps from 1 800 tonnes to 4 500 tonnes SWU per annum. This development in steps depends on the extent to which demand on the world market can be turned into contracts for Urenco. After a procedure taking more than six years to complete, Urenco Deutschland GmbH was issued the required permit by the North Rhine-Westphalian State Ministry for Transport, Energy, and Planning on February 14 this year. The permit was preceded by the drafting and examination of two safety reports of several hundred pages each, and of over 1 000 detailed documents. The application documents were examined by a large number of expert consultants and public authorities from all over Germany acting on behalf of the nuclear licensing authority. Urenco Ltd. sells enrichment services worldwide, at present holding a share of approx. 19% of the world market for uranium separative work. Although this world market is not going to grow for a foreseeable period of time, Urenco is expanding its capacities. This is based on the latest centrifuge technology characterized by extremely low power consumption and high availability. It replaces diffusion technology, which currently holds a share of approx. 40% of the world market. (orig.)

  19. 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

  20. Quality assurance in the enriched uranium operations NDA facility

    Energy Technology Data Exchange (ETDEWEB)

    May, P.K.; Ceo, R.N. [Oak Ridge Y-12 Plant, TN (United States)

    1997-11-01

    The Nondestructive Analysis (NDA) Facility at the Oak Ridge Y-12 Plant has characterized process wastes for Enriched Uranium Operations since 1978. Since that time, over 50,000 items have been analyzed. Analysis results are used to determine whether or not recovery of uranium from process wastes is economically feasible. Our instrument complement includes one large segmented gamma scanner (SGS), two smaller SGS, two solution assay systems (SAS), and Active Well Coincidence Counter (AWCC). The large SGS is used for analyzing High Efficiency Particulate Air (HEPA) filters ant 208-L drums filled with combustible contaminated waste. The smaller SGS are used to analyze 4-L containers of ash and leached residues. The SAS are used to analyze 125 ml bottles of aqueous or organic waste solutions that may contain uranium. The gamma-based NDA techniques are used to identify which process wastes can be discarded, and which must be recycled. The AWCC is used to analyze high-density materials which are not amenable to gamma-ray analysis. 1 ref., 4 figs.