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

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.

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-01

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

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.)

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.

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

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

International Nuclear Information System (INIS)

Chang, R.C.

2002-01-01

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

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).

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.

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)

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.

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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.

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)

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)

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.

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.

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.

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.

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.

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.

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

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

International Nuclear Information System (INIS)

Talnagi, J.W.

1989-01-01

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

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

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)

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

Directory of Open Access Journals (Sweden)

Seung-Kon Lee

2016-06-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-15

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

Influence of uncertainties of isotopic composition of the reprocessed uranium on effectiveness of its enrichment in gas centrifuge cascades

Science.gov (United States)

Smirnov, A. Yu; Mustafin, A. R.; Nevinitsa, V. A.; Sulaberidze, G. A.; Dudnikov, A. A.; Gusev, V. E.

2017-01-01

The effect of the uncertainties of the isotopic composition of the reprocessed uranium on its enrichment process in gas centrifuge cascades while diluting it by adding low-enriched uranium (LEU) and waste uranium. It is shown that changing the content of 232U and 236U isotopes in the initial reprocessed uranium within 15% (rel.) can significantly change natural uranium consumption and separative work (up to 2-3%). However, even in case of increase of these parameters is possible to find the ratio of diluents, where the cascade with three feed flows (depleted uranium, LEU and reprocessed uranium) will be more effective than ordinary separation cascade with one feed point for producing LEU from natural uranium.

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)

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)

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

International Nuclear Information System (INIS)

1994-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

1994-01-01

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

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

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

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

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)

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)

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

Management of high enriched uranium for peaceful purposes: Status and trends

International Nuclear Information System (INIS)

2005-06-01

Arms control agreements between some Nuclear Weapon States have led to the dismantling of many of the nuclear weapons in their military stockpiles, which in turn have produced stockpiles of excess weapons-grade high enriched uranium (HEU) from the dismantled weapons. Considering the proliferation potential of HEU, the management, control and disposition of this fissile material has become a primary focus of nuclear non-proliferation efforts worldwide. To lessen the proliferation threat of excess HEU stockpiles, the USA agreed to purchase several tonnes of excess Russian HEU down-blended to low enriched uranium (LEU). Proliferation concerns about HEU have also resulted in a global effort to convert research reactors from HEU to LEU fuel and to minimize civilian use of HEU. This publication addresses HEU management declared excesses, non-proliferation programmes and options for the use of HEU stockpiles, including disposition programmes. Also addressed are the influence of LEU derived from surplus HEU on the global market for uranium, technical issues associated with utilization and the disposition of HEU

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

Conversion and start up of Tehran Research Reactor with LEU fuel

International Nuclear Information System (INIS)

Zaker, M.

2004-01-01

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

Production of annular blanks for Mo-99 using natural uranium, LEU uranium, nickel and structural Al-3003 plates

International Nuclear Information System (INIS)

Lisboa, J.R.; Barrera, M.E.; Marin, J.

2010-01-01

The Tc-99m radioisotope for medical use is the one most used in nuclear medicine worldwide. In Chile the Tc-99m is applied in more than 90% of nuclear medicine studies. In order to supply the whole country with this radioisotope, in 2005-2007 the CCHEN developed its own production of Tc-99m generators from Mo-99 imported from Canada, which are prepared with the activity needed by the Chilean hospitals and clinics. As of 2007 Mo-99 was no longer imported, and since then the Tc-99m is produced only by neutron activation of the Mo. The present challenge is to produce Mo-99 by irradiating blanks that contain enriched uranium foils, with locally produced LEU. The annular blank consists of 2 concentric tubes of A1-3003 structural aluminum that, in an interior annular space, contain a LEU foil, covered on both sides by a nickel foil. This work presents the development of the production technology for annular blanks using natural uranium and U-325 enriched uranium. The structural components are made with A1-3003 aluminum alloy, the foils are 13 grams of uranium measuring 100 x 50 mm and 120-150 μ thick. The blank was assembled using a methodology to control, adapt and assemble the blank's different internal components. A foil of natural uranium and LEU uranium, and a nickel foil are included, used as a barrier for the escape of fission products. During the blank's expansion, for analysis alcohol as lubricant was used, allowing the expander to move smoothly through the inside of the blank. The blank was sealed by TIG welding with a pulsed AC current and a mixture of Ar-5% He gases. Two methods were used for the water tightness test; for high escape levels the temperature was used as a promoter of the ΔP provided by hot water and liquid nitrogen, for low escape levels high vacuum technology was used where the ΔP is provided by a high pressure helium atmosphere. The technology for the production of annular LEU blanks was achieved by applying innovations to technologies

Neutronic performance of a 14 MW TRIGA reactor: LEU vs HEU fuel

International Nuclear Information System (INIS)

Bretscher, M.M.; Snelgrove, J.L.; Cornella, R.J.

1983-01-01

A primary objective of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is to develop means for replacing, wherever possible, currently used highly-enriched uranium (HEU) fuel ( 235 U enrichment > 90%) with low-enriched uranium (LEU) fuel ( 235 U enrichment < 20%) without significantly degrading the performance of research and test reactors. The General Atomic Company has developed a low-enriched but high uranium content Er-U-ZrH/sub 1.6/ fuel to enable the conversion of TRIGA reactors (and others) from HEU to LEU. One possible application is to the water-moderated 14 MW TRIGA Steady State Reactor (SSR) at the Romanian Institute for Nuclear Power Reactors. The work reported here was undertaken for the purpose of comparing the neutronic performance of the SSR for HEU fuel with that for LEU fuel. In order to make these relative comparisons as valid as possible, identical methods and models were used for the neutronic calculations

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

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

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

International Nuclear Information System (INIS)

1995-11-01

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

Production of MO-99 from LEU targets-base-side processing

International Nuclear Information System (INIS)

Vandegrift, George F.; Koma, Yoshikazu; Cols, Hector; Conner, Cliff; Aase, Scott; Peter, Magdalin; Walker, David; Leonard, Ralph A.; Snelgrove, James L.

2000-01-01

Argonne National Laboratory (ANL) is cooperating with the Argentine Comision Nacional de Energia Atomica (CNEA) to convert their 99 Mo production process, which uses high enriched uranium (HEU), to low-enriched uranium (LEU). Progress discussed in this year's paper includes optimization of (1) the digestion of LEU foil by sodium hydroxide solution and (2) the primary recovery of molybdenum by anion exchange. Also discussed are ANL/CNEA plans for demonstrating the irradiation and digestion of LEU-foil targets and recovering 99 Mo in Argentina later this year. Our results show that, up to this point in our study, conversion of the CNEA process to LEU appears viable. (author)

Impact of the use of low or medium enriched uranium on the masses of space nuclear reactor power systems

International Nuclear Information System (INIS)

1994-12-01

The design process for determining the mass increase for the substitution of low-enriched uranium (LEU) for high-enriched uranium (HEU) in space nuclear reactor systems is an optimization process which must simultaneously consider several variables. This process becomes more complex whenever the reactor core operates on an in-core thermionic power conversion, in which the fissioning of the nuclear fuel is used to directly heat thermionic emitters, with the subsequent elimination of external power conversion equipment. The increased complexity of the optimization process for this type of system is reflected in the work reported herein, where considerably more information has been developed for the moderated in-core thermionic reactors

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

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)

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.

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.

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

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)

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)

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

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

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)

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)

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

Comparison of the FRM-II HEU design with an alternative LEU design

International Nuclear Information System (INIS)

Mo, S.C.; Hanan, N.A.; Matos, J.E.

2004-01-01

The FRM-II reactor design of the Technical University of Munich has a compact core that utilizes fuel plates containing highly-enriched uranium (HEU, 93%). This paper presents an alternative core design utilizing low-enriched uranium (LEU, 3 that provides nearly the same neutron flux for experiments as the HEU design, but has a less favourable fuel cycle economy. If an LEU fuel with a uranium density of 6.0 - 6.5 g/cm 3 . were developed, the alternative design would provide the same neutron flux and use the same number of cores per year as the HEU design. The results of this study show that there are attractive possibilities for using LEU fuel instead of HEU fuel in the FRM-II. Further optimization of the LEU design and near-term availability of LEU fuel with a uranium density greater than 4.8 g/cm 3 would enhance the performance of the LEU core. The REKIR Program is ready to exchange information with the Technical University of Munich to resolve any differences that may exist and to identify design modifications that would optimize reactor performance utilizing LEU fuel. (author)

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.

Low enrichment Mo-99 target development program at ANSTO

International Nuclear Information System (INIS)

Donlevy, Therese M.; Anderson, Peter J.; Beattie, David; Braddock, Ben; Fulton, Scott; Godfrey, Robert; Law, Russell; McNiven, Scott; Sirkka, Pertti; Storr, Greg; Wassink, David; Wong, Alan; Yeoh, Guan

2002-01-01

The Australian Nuclear Science and Technology Organisation (ANSTO, formerly AAEC) has been producing fission product Mo-99 in HIFAR, from the irradiation of Low Enrichment Uranium (LEU) UO 2 targets, for nearly thirty years. Over this period, the U-235 enrichment has been increased in stages, from natural to 1.8% to 2.2%. The decision to provide Australia with a replacement research reactor (RRR) for HIFAR has created an ideal opportunity to review and improve the current Mo-99 production process from target design through to chemical processing and waste management options. ANSTO has entered into a collaboration with Argonne National Laboratory (RERTR) to develop a target using uranium metal foil with U-235 enrichment of less than 20% The initial focus has been to demonstrate use of LEU foil targets in HIFAR, using existing irradiation methodology. The current effort focussed on designing a target assembly with optimised thermohydraulic characteristics to accommodate larger LEU foils to meet Mo-99 production needs. The ultimate goal is to produce an LEU target suitable for use in the Replacement Research Reactor when it is commissioned in 2005. This paper reports our activities on: - The regulatory approval processes required in order to undertake irradiation of this new target; -Supporting calculations (neutronics, computational fluid dynamics) for safety submission; - Design challenges and changes to prototype irradiation; - Trial irradiation of LEU foil target in HIFAR; - Future target and rig development program at ANSTO. (author)

ANL progress in developing an LEU target and process for Mo-99 production: Cooperation with CNEA

International Nuclear Information System (INIS)

Gelis, A.V.; Vandegrift, G.F.; Aase, S.B.; Bakel, A.J.; Falkenberg, J.R.; Regalbuto, M.C.; Quigley, K.J.

2003-01-01

The primary mission of the Reduced Enrichment in Research and Test Reactors (RERTR) Program is to facilitate the conversion of research and test-reactor fuel and targets from high-enriched uranium (HEU) to low-enriched uranium (LEU). One of the current goals at Argonne National Laboratory (ANL) is to assist the Argentine Comision Nacional de Energia Atomica (CNEA) in developing an LEU foil target and a process for 99 Mo production. Specifically addressed in this paper is ANL R and D related to this conversion: (1) designing a prototype production vessel for digesting irradiated LEU foils in alkaline solutions and (2) developing a new digestion method to address all issues related to HEU to LEU conversion. (author)

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)

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

International Nuclear Information System (INIS)

Westmacott, Chad; Cameron, Ron

2012-01-01

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

Progress in chemical processing of LEU targets for 99Mo production - 1997

International Nuclear Information System (INIS)

Vandegrift, G.F.; Conner, C.; Sedlet, J.; Wygmans, D.G.; Wu, D.; Iskander, F.; Landsberger, S.

1997-01-01

Presented here are recent experimental results of our continuing development activities associated with converting current processes for producing fission-product 99 Mo from targets using high-enriched uranium (HEU) to low-enriched uranium (LEU). Studies were focused in four areas: (1) measuring the chemical behavior of iodine, rhodium, and silver in the LEU-modified Cintichem process, (2) performing experiments and calculations to assess the suitability of zinc fission barriers for LEU metal foil targets, (3) developing an actinide separations method for measuring alpha contamination of the purified 99 Mo product, and (4) developing a cooperation with Sandia National Laboratories and Los Alamos National Laboratory that will lead to approval by the U.S. Federal Drug Administration for production of 99 Mo from LEU targets. Experimental results continue to show the technical feasibility of converting current HEU processes to LEU. (author)

TRIGA high wt -% LEU fuel development program. Final report

International Nuclear Information System (INIS)

West, G.B.

1980-07-01

The principal purpose of this work was to investigate the characteristics of TRIGA fuel where the contained U-235 was in a relatively high weight percent (wt %) of LEU (low enriched uranium - enrichment of less than 20%) rather than a relatively low weight percent of HEU (high enriched uranium). Fuel with up to 45 wt % U was fabricated and found to be acceptable after metallurgical examinations, fission product retention tests and physical property examinations. Design and safety analysis studies also indicated acceptable prompt negative temperature coefficient and core lifetime characteristics for these fuels

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.

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.

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)

Measurements of the HEU and LEU in-core spectra at the Ford Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Wehe, D K [Oak Ridge National Laboratory, Oak Ridge, TN (United States); King, J S; Lee, J C; Martin, W R [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

1985-07-01

The Ford Nuclear Reactor (FNR) at the University of Michigan has been serving as the test site for a low-enriched uranium (LEU) fuel whole-core demonstration. As part of the experimental program, the differential neutron spectrum has been measured in a high-enriched uranium (HEU) core and an LEU core. The HEU and LEU spectra were determined by unfolding the measured activities of foils that were irradiated in the reactor. When the HEU and LEU spectra are compared from meV to 10 MeV, significant differences between the two spectra are apparent below 10 eV. These are probably caused by the additional {sup 238}U resonance absorption in the LEU fuel. No measurable difference occurs in the shape of the spectra above MeV. (author)

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.

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

Making of fission 99Mo from LEU silicide(s): A radiochemists' view

International Nuclear Information System (INIS)

Kolar, Z.I.; Wolterbeek, H.Th.

2005-01-01

The present-day industrial scale production of 99 Mo is fission based and involves thermal-neutron irradiation in research reactors of highly enriched uranium (HEU, > 20 % 235 U) containing targets, followed by radiochemical processing of the irradiated targets resulting in the final product: a 99 Mo containing chemical compound of molybdenum. In 1978 a program (RERTR) was started to develop a substitute for HEU reactor fuel i.e. a low enriched uranium (LEU, 235 U) one. In the wake of that program studies were undertaken to convert HEU into LEU based 99 Mo production. Both new targets and radiochemical treatments leading to 99 Mo compounds were proposed. One of these targets is based on LEU silicide, U 3 Si 2 . Present paper aims at comparing LEU U 3 Si 2 and LEU U 3 Si with another LEU target i.e. target material and arriving at some preferences pertaining to 99 Mo production. (author)

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)

Development of LEU targets for 99Mo production and their chemical processing status 1989

International Nuclear Information System (INIS)

Vandegrift, G.F.; Kwok, J.D.; Chamberlain, D.B.; Hoh, J.C.; Streets, E.W.; Vogler, S.; Thresh, H.R.; Domagala, R.F.; Wiencek, T.C.; Matos, J.E.

1991-01-01

Most of the world's supply of Tc-99m for medical purposes is currently produced from Mo-99 derived from the fissioning of high enriched uranium (HEU). Substitution of low enriched uranium (LEU) silicide fuel for the HEU alloy and aluminide fuels used in current target designs will allow equivalent Mo-99 yields with no change in target geometries. Substitution of uranium metal will also allow the substitution of LEU for HEU. Efforts performed in 1989 focused on (1) fabrication of a uranium metal target by Hot Isostatic Pressing uranium metal foil to zirconium, (2) experimental investigation of the dissolution step for U 3 Si 2 targets, allowing us to present a conceptual design for the dissolution process and equipment, and (3) investigation of the procedures used to reclaim irradiated uranium from Mo-production targets, allowing us to further analyze the waste and by-product problems associated with the substitution of LEU for HEU. (orig.)

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

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

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

Operational experience in the production of 131Molybdenum and 99Iodine with high and low uranium enrichment

International Nuclear Information System (INIS)

Bravo, C.; Cristini, Pablo R..; Novello, A.; Bronca, M.; Cestau, Daniel; Centurion, R.; Bavaro, R.; Cestau, J.; Gualda, E.; Bronca, P.; Carranza, Eduardo C.

2009-01-01

In 1992, in an effort to curtail use of Highly Enriched Uranium (HEU), hoping to alleviate nuclear security concerns, United States passed the Schumer amendment to the Energy Policy Act. This legislation conditioned U.S. export of HEU to foreign companies, understanding that these companies would switch as soon as possible to Lowly Enriched Uranium (LEU). This paper describes 99 Mo production flow chart, characteristics of process cells, shielding, systems of manipulation at distance, cell ventilation system and the method for personal dose monitoring. Production evolution for the span of years 1998 to 2007 is given by indicators, keeping in mind enrichment proportion change. Evolution shown on the indicators is directly related to the application of Safety Culture concepts adopted by personnel. (author)

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

ANL progress in developing a target and process for converting CNEA Mo-99 production to LEU

International Nuclear Information System (INIS)

Vandegrift, G.F.; Gelis, A.; Aase, S.; Bakel, A.; Freiberg, E.; Conner, C.

2002-01-01

The primary mission of the Reduced Enrichment in Research and Test Reactors (RERTR) Program is to facilitate the conversion of research and test reactor fuel and targets from high-enriched uranium (HEU) to low-enriched uranium (LEU). One of the current goals at Argonne National Laboratory (ANL) is to convert 99 Mo production at Argentine Commission Nacional de Energia Atomica (CNEA) from HEU to LEU targets. Specifically addressed in this paper is ANL R and D related to this conversion: (1) designing a prototype production vessel for digesting irradiated LEU foils in alkaline solutions, (2) developing means to improve digestion efficiency, and (3) modifying ion-exchange processes used in the CNEA recovery and purification of 99 Mo to deal with the lower volumes generated from LEU-foil digestion. (author)

Uranium silicide activities at Babcock and Wilcox

International Nuclear Information System (INIS)

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

1983-01-01

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

Processing of LEU targets for 99Mo production--testing and modification of the Cintichem process

International Nuclear Information System (INIS)

Wu, D.; Landsberger, S.; Buchholz, B.

1995-09-01

Recent experimental results on testing and modification of the Cintichem process to allow substitution of low enriched uranium (LEU) for high enriched uranium (HEU) targets are presented in this report. The main focus is on 99 Mo recovery and purification by its precipitation with α-benzoin oxime. Parameters that were studied include concentrations of nitric and sulfuric acids, partial neutralization of the acids, molybdenum and uranium concentrations, and the ratio of α-benzoin oxime to molybdenum. Decontamination factors for uranium, neptunium, and various fission products were measured. Experiments with tracer levels of irradiated LEU were conducted for testing the 99 Mo recovery and purification during each step of the Cintichem process. Improving the process with additional processing steps was also attempted. The results indicate that the conversion of molybdenum chemical processing from HEU to LEU targets is possible

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)

The conversion of NRU from HEU to LEU fuel

International Nuclear Information System (INIS)

Sears, D.F.; Atfield, M.D.; Kennedy, I.C.

1990-01-01

The program at Chalk River Nuclear Laboratories (CRNL) to develop and test low-enriched uranium fuel (LEU, 3 Si, USiAl, USi Al and U 3 Si 2 (U-3.96 wt% Si; U-3.5 wt% Si-1.5 wt% AL; U-3.2 wt%; Si-3 wt% Al; U-7.3 wt% Si, respectively). Fuel elements were fabricated with uranium loadings suitable for NRU, 3.15 gU/cm 3 , and for NRX, 4.5 gU/cm 3 , and were irradiated under normal fuel-operating conditions. Eight experimental irradiations involving 100 mini-elements and 84 full-length elements (7X12-element rods) were completed to qualify the LEU fuel and the fabrication technology. Post irradiation examinations confirmed that the performance of the LEU fuel, and that of a medium enrichment uranium (MEU, 45% U-235) alloy fuel tested as a back-up, was comparable to the HEU fuel. The uranium silicide dispersion fuel swelling was approximately linear up to burnups exceeding NRU's design terminal burnup (80 at%). NRU was partially converted to LEU fuel when the first 31 prototype fuel rods manufactured with industrial scale production equipment were installed in the reactor. The rods were loaded in NRU at a fuelling rate of about two rods per week over the period 1988 September to December. This partial LEU core (one third of a full NRU core) has allowed the reactor engineers and physicists to evaluate the bulk effects of the LEU conversion on NRU operations. As expected, the irradiation is proceeding without incident

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)

Determination of Dancoff correction thermal utilization and thermal disadvantage factors of HEU and LEU cores of an MNSR

International Nuclear Information System (INIS)

Ofori, Y. T.

2013-07-01

Ghana Research Reactor-1 (GHARR-1), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (Highly Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of the conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. In this research work, a comparative study has been performed for the determination of the Dancoff, thermal utilization and thermal disadvantage factors of highly enriched uranium (HEU) and potential low enriched uranium (LEU) cores of GHARR-1. A one group transport theory and collision probability based methodologies was used to develop mathematical formulations for thermal utilization factor and thermal disadvantage factor assuming isotropic scattering. This methodology was implemented in a FORTRAN 95 based computer program THERMCALC, which uses Bessell and BesselK as subroutines developed to calculate the modified Bessel functions I n and K n respectively using the polynomial approximation method. Furthermore, a Dancoff correction factor of 0.1519 thermal utilization factor of 0.9767 and a thermal disadvantage factor of 1.894 were obtained for the 90.2% highly enriched Uranium core of GHARR-1. The results compare favorably with literature. Thus THERMCALC can be used as a reliable tool for the calculation of Dancoff, thermal utilization and disadvantage factors of MNSR cores. Other potential LEU cores; UO 2 (with different fuel meat densities and enrichments) and U 3 Si 2 have also been analysed. UO 2 with 12.6% of Uranium-235 was chosen as the most potential LEU core for the GHARR-1. (au)

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)

Safeguards considerations for uranium enrichment facilities, as applied to gas centrifuge and gaseous diffusion facilities

International Nuclear Information System (INIS)

1979-03-01

The goals and objectives of IAEA safeguards as they are understood by the authors based on published documents are reviewed. These goals are then used to derive safeguards concerns, diversion strategies, and potential safeguards measures for four base cases, the production of highly enriched uranium (HEU) at a diffusion plant, the diversion of low enriched uranium (LEU) at a diffusion plant, the diversion of HEU at a gas centrifuge plant, and the diversion of LEU at a gas centrifuge plant. Tables of estimated capabilities are given for each case, under the assumption that the inspector would have access: to the cascade perimeter at or after the start of operations, to the cascade perimeter throughout construction and operation, to the cascade perimeter during operation plus a one-time access to the cascade itself, to the cascade during construction but only its perimeter during operation, or to the cascade itself during construction and operation

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.

Development of Fission Mo-99 Process for LEU Dispersion Target

Energy Technology Data Exchange (ETDEWEB)

Lee, Seung Kon; Lee, Su Seung; Hong, Soon Bog; Jang, Kyung Duk; Park, Ul Jae; Lee, Jun Sig [KAERI, Daejeon (Korea, Republic of)

2016-05-15

KAERI (Korea Atomic Energy Research Institute) is developing LEU-based fission {sup 99}Mo production process which is connected to the new research reactor (Kijang New Research Reactor, KJRR), which is being constructed in Gijang, Busan, Korea. Historically, the most fission {sup 99}Mo producers have been used highly enriched uranium (HEU) targets so far. However, to reduce the use of HEU in private sector for non-proliferation, {sup 99}Mo producers are forced to convert their HEU-based process to use low enriched uranium (LEU) targets. Economic impact of a target conversion from HEU to LEU is significant. Overall cost for the production of the fission {sup 99}Mo increases significantly with the conversion of fission {sup 99}Mo targets from HEU to LEU. It is not only because the yield of LEU is only 50% of HEU, but also because radioactive waste production increases 200%. On the basis, worldwide efforts on the development of {sup 99}Mo production process that is optimized for the LEU target become an important issue. In this study, fission {sup 99}Mo process with non-irradiated LEU targets was presented except separation and purification steps. Pre- and post-irradiation tests of the fission {sup 99}Mo target will be done in 4th quarter of 2016.

Development of Fission Mo-99 Process for LEU Dispersion Target

International Nuclear Information System (INIS)

Lee, Seung Kon; Lee, Su Seung; Hong, Soon Bog; Jang, Kyung Duk; Park, Ul Jae; Lee, Jun Sig

2016-01-01

KAERI (Korea Atomic Energy Research Institute) is developing LEU-based fission 99 Mo production process which is connected to the new research reactor (Kijang New Research Reactor, KJRR), which is being constructed in Gijang, Busan, Korea. Historically, the most fission 99 Mo producers have been used highly enriched uranium (HEU) targets so far. However, to reduce the use of HEU in private sector for non-proliferation, 99 Mo producers are forced to convert their HEU-based process to use low enriched uranium (LEU) targets. Economic impact of a target conversion from HEU to LEU is significant. Overall cost for the production of the fission 99 Mo increases significantly with the conversion of fission 99 Mo targets from HEU to LEU. It is not only because the yield of LEU is only 50% of HEU, but also because radioactive waste production increases 200%. On the basis, worldwide efforts on the development of 99 Mo production process that is optimized for the LEU target become an important issue. In this study, fission 99 Mo process with non-irradiated LEU targets was presented except separation and purification steps. Pre- and post-irradiation tests of the fission 99 Mo target will be done in 4th quarter of 2016

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

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.

Irradiation experiment conceptual design parameters for MURR LEU U-Mo fuel conversion

International Nuclear Information System (INIS)

Stillman, J.; Feldman, E.; Stevens, J.; Wilson, E.

2013-03-01

This report contains the results of reactor design and performance calculations 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 from the U. S. government. 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 MURR. This report presents the nominal steady-state irradiation conditions of a key set of plates containing peak irradiation parameters found in MURR cores fueled with the LEU monolithic U-Mo alloy fuel with 10 wt% Mo.

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)

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)

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

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)

HEU to LEU conversion and blending facility: Metal blending alternative to produce LEU oxide for disposal

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. The nuclear material is converted to a form more proliferation- resistant than the original form. Blending HEU (highly enriched uranium) with less-enriched uranium to form LEU has been proposed as a disposition option. Five technologies are being assessed for blending HEU. This document provides data to be used in environmental impact analysis for the HEU-LEU disposition option that uses metal blending with an oxide waste product. It is divided into: mission and assumptions, conversion and blending facility descriptions, process descriptions and requirements, resource needs, employment needs, waste and emissions from plant, hazards discussion, and intersite transportation.

HEU to LEU conversion and blending facility: Metal blending alternative to produce LEU oxide for disposal

International Nuclear Information System (INIS)

1995-09-01

US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. The nuclear material is converted to a form more proliferation- resistant than the original form. Blending HEU (highly enriched uranium) with less-enriched uranium to form LEU has been proposed as a disposition option. Five technologies are being assessed for blending HEU. This document provides data to be used in environmental impact analysis for the HEU-LEU disposition option that uses metal blending with an oxide waste product. It is divided into: mission and assumptions, conversion and blending facility descriptions, process descriptions and requirements, resource needs, employment needs, waste and emissions from plant, hazards discussion, and intersite transportation

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-02-21

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

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

HEU to LEU conversion experience at the UMass-Lowell research reactor

International Nuclear Information System (INIS)

White, John R.; Bobek, Leo M.

2005-01-01

The UMass-Lowell Research Reactor (UMLRR) operated safely with high-enriched uranium (HEU) fuel for over 25 years. Having reached the end of core lifetime and due to proliferation concerns, the reactor was recently converted to low-enriched uranium silicide (LEU) fuel. The actual process for converting the UMLRR from HEU to LEU fuel covered a period of over 15 years. The conversion effort - from the initial conceptual design studies in the late 1980s to the final offsite shipment of the spent HEU fuel in August 2004 - was a unique experience for the faculty and staff of a small university research reactor. This paper gives a historical view of the process and it highlights several key milestones along the road to successful completion of this project. (author)

Production of MO-99 from LEU targets - Acid-side processing

International Nuclear Information System (INIS)

Conner, C.; Sedlet, J.; Wiencek, T.C.

2000-01-01

During 2000, additional targets of the new annular design containing low enriched uranium (LEU) foils were irradiated in the Indonesian RSG-GAS reactor. This new design significantly decreases the target fabrication cost. This irradiation allowed us to compare the irradiation performance of several batches of LEU foil. We also processed one of the irradiated foils to recover 99 Mo using a slightly modified Cintichem process. Finally, we measured some important physical properties of uranyl nitrate solutions (i.e., density and solubility), which will be useful in future efforts to further increase the amount of uranium that can be processed by the Cintichem process. (author)

Reduced Enrichment for Research and Test Reactors. Proceedings of the XVIII international meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

Almost 50 papers presented were showing the status of the national programs related to conversion of research reactor cores from highly enriched (HEU) to low enriched uranium (LEU) fuel elements. Design of new fuel elements (uranium silicides) and safety related calculations were dealt with taking into account fuel cycle issues, meaning spent fuel storage and transportation. A number of presentations were devoted to Mo-99 production using LEU targets.

Reduced Enrichment for Research and Test Reactors. Proceedings of the XVIII international meeting

International Nuclear Information System (INIS)

2004-01-01

Almost 50 papers presented were showing the status of the national programs related to conversion of research reactor cores from highly enriched (HEU) to low enriched uranium (LEU) fuel elements. Design of new fuel elements (uranium silicides) and safety related calculations were dealt with taking into account fuel cycle issues, meaning spent fuel storage and transportation. A number of presentations were devoted to Mo-99 production using LEU targets

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)

Fission product release from TRIGA-LEU reactor fuels

International Nuclear Information System (INIS)

Baldwin, N.L.; Foushee, F.C.; Greenwood, J.S.

1980-01-01

Due to present international concerns over nuclear proliferation, TRIGA reactor fuels will utilize only low-enriched uranium (LEU) (enrichment <20%). This requires increased total uranium loading per unit volume of fuel in order to maintain the appropriate fissile loading. Tests were conducted to determine the fractional release of gaseous and metallic fission products from typical uranium-zirconium hydride TRIGA fuels containing up to 45 wt-% uranium. These tests, performed in late 1977 and early 1978, were similar to those conducted earlier on TRIGA fuels with 8.5 wt-% U. Fission gas release measurements were made on prototypic specimens from room temperature to 1100 deg. C in the TRIGA King Furnace Facility. The fuel specimens were irradiated in the TRIGA reactor at a low power level. The fractional releases of the gaseous nuclides of krypton and xenon were measured under steady-state operating conditions. Clean helium was used to sweep the fission gases released during irradiation from the furnace into a standard gas collection trap for gamma counting. The results of these tests on TRIGA-LEU fuel agree well with data from the similar, earlier tests on TRIGA fuel. The correlation used to calculate the release of fission products from 8.5 wt-% U TRIGA fuel applies equally well for U contents up to 45 wt-%. (author)

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

Multilateral nonproliferation cooperation: US - Led effort to remove HEU/LEU fresh and spent fuels from the Republic of Georgia to Dounreay, Scotland

International Nuclear Information System (INIS)

Shelton, Thomas A.; Viebrock, James M.; Riedy, Alexander W.; Moses, Stanley D.; Bird, Helen M.

1998-01-01

This paper presents the efforts led by United States for removing HEU/LEU fresh and spent fuel from dhe Republic of Georgia to Dounreay, Scotland. These efforts are resulted from a plan approved by the United States Government, in cooperation with the United Kingdom and Georgia Governments to rapidly retrieve and transport circa 4.3 kilograms of enriched uranium. This material consisted largely of highly enriched uranium (HEU) and a small amount of low enriched uranium (LEU) fresh fuel, as well as about 800 grams of HEU/LEU-based spent fuel from a shutdown IR T-M research reactor on the outskirts of Table's, Georgia. The technical team lead by DOE consisted of HEU handling, packaging and transportation experts from the Oak Ridge Y-12 plant, managed and operated by Lockheed Martin Energy Systems, and fuel handling and transportation experts from Nac International in Norcross, Georgia, United States

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?

Development of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.

1997-01-01

The Reduced Enrichment Research and Test Reactor Program has continued its effort in the past 3 yr to develop use of low-enriched uranium (LEU) to produce the fission product 99 Mo. This work comprises both target and chemical processing development and demonstration. Two major target systems are now being used to produce 99 Mo with highly enriched uranium-one employing research reactor fuel technology (either uranium-aluminum alloy or uranium aluminide-aluminum dispersion) and the other using a thin deposit of UO 2 on the inside of a stainless steel (SST) tube. This paper summarizes progress in irradiation testing of targets based on LEU uranium metal foils. Several targets of this type have been irradiated in the Indonesian RSG-GAS reactor operating at 22.5 MW

The global threat reduction initiative and conversion of isotope production to LEU targets

International Nuclear Information System (INIS)

Kuperman, A. J.

2005-01-01

The U.S. Global Threat Reduction Initiative (GTRI) has given a decisive impetus to the RERTR program's longstanding goal of converting worldwide production of medical radioisotopes from reliance on bomb-grade, highly enriched uranium (HEU) to low-enriched uranium (LEU) unsuitable for weapons. Although the four major; isotope producers continue to resist calls for conversion, they face mounting pressure from a variety of fronts including: (1) GTRI; (2) a related, multilateral U.S. initiative to forge agreement on conversion among the states that are home to the major producers; (3) an IAEA effort to provide technical assistance that will facilitate large-scale production of medical isotopes using LEU by producers who seek to do so; (4) planned production in the United States of substantial quantities of medical isotopes using LEU; and (5) pending U.S. legislation that would prohibit the export of HEU for production of isotopes as soon as alternative, LEU-produced isotopes are available. Accordingly, it now appears inevitable that worldwide isotope production will be converted from reliance on HEU to LEU. The only remaining question is which producers will be the first to reliably deliver sizeable quantities of LEU-produced isotopes and thereby capture global market share from the others. (author)

Disposition of highly enriched uranium obtained from the Republic of Kazakhstan. Environmental assessment

International Nuclear Information System (INIS)

1995-05-01

This EA assesses the potential environmental impacts associated with DOE's proposal to transport 600 kg of Kazakhstand-origin HEU from Y-12 to a blending site (B ampersand W Lynchburg or NFS Erwin), transport low-enriched UF6 blending stock from a gaseous diffusion plant to GE Wilmington and U oxide blending stock to the blending site, blending the HEU and uranium oxide blending stock to produce LEU in the form of uranyl nitrate, and transport the uranyl nitrate from the blending site to USEC Portsmouth

Development of production of {sup 99}Mo from LEU target

Energy Technology Data Exchange (ETDEWEB)

Adang, H G; Mutalib, A; Lubis, H [Radioisotope Production Centre, National Atomic Energy Agency, Kawasan Puspiptek, Serpong (Indonesia); and others

1998-10-01

{sup 99}TC, the most popular radioisotope in nuclear medicine, is daughter of {sup 99}Mo. {sup 99}Mo is produced in research reactor by irradiating of high enriched uranium (HEU). However, in recent year, strict regulation that has been implemented by USA DOE and NPT has led to the difficulty in getting HEU. Therefore, BATAN has tried to develop the production of {sup 99}Mo by using low enriched uranium (LEU). The research involves the use of LEU in the production of {sup 99}Mo. This research was started in 1994 by joint-research between BATAN and Argonne National Laboratory USA. This program is divided into three research groups. The first group emphasizes its research on fabrication of LEU foil that is going to be irradiated. The second group studies the irradiation`s aspects and physical characteristic of irradiated LEU foils. The third group studies the radiochemical separation process of fission product {sup 99}Mo from solution of irradiated LEU foils. There are five steps that are carried out in studying of radiochemical separation of {sup 99}Mo from irradiated LEU. First is designing a dissolver that is going to be used in dissolving of LEU foil and testing its reliability. Second is dissolving LEU in the new design dissolver. Third is evaluation the modified of Cintichem`s radiochemical separation process of {sup 99}Mo from LEU. Forth is modifying the Cintichem`s radiochemical separation process of {sup 99}Mo from the solution of irradiated LEU. And fifth is using the modified of Cintichem`s radiochemical separation process for separation {sup 99}Mo from solution of irradiated LEU. The first through the forth steps of experiments were already carried out and will be reported in this workshop, whereas the fifth step of experiment is going to be conducted in February 1998. (author)

Finding of no significant impact: Interim storage of enriched uranium above the maximum historical level at the Y-12 Plant Oak Ridge, Tennessee

International Nuclear Information System (INIS)

1995-01-01

The US Department of Energy (DOE) has prepared an Environmental Assessment (EA) for the Proposed Interim Storage of Enriched Uranium Above the Maximum Historical Storage Level at the Y-12 Plant, Oak Ridge, Tennessee (DOE/EA-0929, September, 1994). The EA evaluates the environmental effects of transportation, prestorage processing, and interim storage of bounding quantities of enriched uranium at the Y-12 Plant over a ten-year period. The State of Tennessee and the public participated in public meetings and workshops which were held after a predecisional draft EA was released in February 1994, and after the revised pre-approval EA was issued in September 1994. Comments provided by the State and public have been carefully considered by the Department. As a result of this public process, the Department has determined that the Y-12 Plant-would store no more than 500 metric tons of highly enriched uranium (HEU) and no more than 6 metric tons of low enriched uranium (LEU). The bounding storage quantities analyzed in the pre-approval EA are 500 metric tons of HEU and 7,105.9 metric tons of LEU. Based on-the analyses in the EA, as revised by the attachment to the Finding of No Significant Impact (FONSI), DOE has determined that interim storage of 500 metric tons of HEU and 6 metric tons of LEU at the Y-12 Plant does not constitute a major Federal action significantly affecting the quality of the human environment, within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, an Environmental Impact Statement (EIS) is not required and the Department is issuing this FONSI

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.

Neutronic analysis of the conversion of HEU to LEU fuel for a 5-MW MTR core

International Nuclear Information System (INIS)

Pazirandeh, A.; Bartsch, G.

1987-01-01

In recent years, due to cessation of highly enriched uranium (HEU) fuel supply, practical steps have been taken to substitute HEU fuel in almost all research reactors by medium-enriched uranium or low-enriched uranium (LEU) fuels. In this study, a neutronic calculation of a 5-MW research reactor core fueled with HEU (93% 235 U) is presented. In order to assess the performance of the core with the LEU ( 235 U loadings were examined. The core consists of 22 standard fuel elements (SFEs) and 6 control fuel elements (CFEs). Each fuel elements has 18 curved plates of which two end plates are dummies. Initial 235 U content is 195 g 235 U/SFE and 9.7 g 235 U/CFE or /PFE. In all calculations the permitted changes to the fuel elements are (a) 18 active plates per SFE, (b) fuel plates assumed to be flat, and (c) 8 or 9 active plates per CFE

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

Analysis of the TREAT LEU Conceptual Design

Energy Technology Data Exchange (ETDEWEB)

Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-03-01

Analyses were performed to evaluate the performance of the low enriched uranium (LEU) conceptual design fuel for the conversion of the Transient Reactor Test Facility (TREAT) from its current highly enriched uranium (HEU) fuel. TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. TREAT is currently in non-operational standby, but is being restarted under the U.S. Department of Energy’s Resumption of Transient Testing Program. The conversion of TREAT is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration’s Material Management and Minimization (M3) Reactor Conversion Program. The focus of this study was to demonstrate that the converted LEU core is capable of maintaining the performance of the existing HEU core, while continuing to operate safely. Neutronic and thermal hydraulic simulations have been performed to evaluate the performance of the LEU conceptual-design core under both steady-state and transient conditions, for both normal operation and reactivity insertion accident scenarios. In addition, ancillary safety analyses which were performed for previous LEU design concepts have been reviewed and updated as-needed, in order to evaluate if the converted LEU core will function safely with all existing facility systems. Simulations were also performed to evaluate the detailed behavior of the UO₂-graphite fuel, to support future fuel manufacturing decisions regarding particle size specifications. The results of these analyses will be used in conjunction with work being performed at Idaho National Laboratory and Los Alamos National Laboratory, in order to develop the Conceptual Design Report project deliverable.

Comparison of the parameters of the IR-8 reactor with different fuel assembly designs with LEU fuel

International Nuclear Information System (INIS)

Vatulin, A.; Stetsky, Y.; Dobrikova, I.

1999-01-01

The estimation of neutron-physical, heat and hydraulic parameters of the IR-8 research reactor with low enriched uranium (LEU) fuel was performed. Two fuel assembly (FA) designs were reviewed: IRT-4M with the tubular type fuel elements and IRT-MR with the rod type fuel elements. UO 2 -Al dispersion 19.75% enrichment fuel is used in both cases. The results of the calculations were compared with main parameters of the reactor, using the current IRT-3M FA with 90% high enriched uranium (HEU) fuel. The results of these comparisons showed that during the LEU conversion of the reactor the cycle length, excess reactivity and peak power of the IRT-MR type FA are higher than for the IRT-3M type FA and IRT-4M type FA. (author)

Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA® Reactor

International Nuclear Information System (INIS)

Schickler, R.A.; Marcum, W.R.; Reese, S.R.

2013-01-01

Highlights: • The Oregon State TRIGA ® Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA ® Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least-squares technique. The quantification of

The Y-12 National Security Complex Foreign Research Reactor Uranium Supply Production

Energy Technology Data Exchange (ETDEWEB)

Nelson, T. [Nuclear Technology and Nonproliferation Programs, B and W Y-12, L.L.C., Y-12 National Security Complex, Oak Ridge, Tennessee (United States); Keller, A.P. [Disposition and Supply Programs, B and W Y-12, L.L.C., Y-12 National Security Complex, Oak Ridge, Tennessee (United States)

2011-07-01

The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the National Nuclear Security Administration (NNSA) HEU Disposition, the Reduced Enrichment Research and Test Reactors (RERTR), and the United States (U.S.) FRR Spent Nuclear Fuel (SNF) Acceptance Programs. The FRR Supply Program supports the important U.S. government nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to Low-Enriched Uranium (LEU) fuel under the RERTR Program. The NNSA Y-12 Site Office maintains the prime contracts with foreign government agencies for the supply of LEU for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. In addition to uranium metal feedstock for fuel fabrication, Y-12 can produce LEU in different forms to support new fuel development or target fabrication for medical isotope production. With production improvements and efficient delivery preparations, Y-12 continues to successfully support the global research reactor community. (author)

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

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.

Implementation of the United States-Russian Highly Enriched Uranium Agreement: Current Status and Prospects

International Nuclear Information System (INIS)

R.rutkowski, E; Armantrout, G; Mastal, E; Glaser, J; Benton, J

2004-01-01

The National Nuclear Security Administration's (NNSA) Highly Enriched Uranium (HEU) Transparency Implementation Program (TIP) monitors and provides assurance that Russian weapons-grade HEU is processed into low enriched uranium (LEU) under the transparency provisions of the 1993 United States (U.S.)-Russian HEU Purchase Agreement. Meeting the Agreement's transparency provisions is not just a program requirement; it is a legal requirement. The HEU Purchase Agreement requires transparency measures to be established to provide assurance that the nonproliferation objectives of the Agreement are met. The Transparency concept has evolved into a viable program that consists of complimentary elements that provide necessary assurances. The key elements include: (1) monitoring by technical experts; (2) independent measurements of enrichment and flow; (3) nuclear material accountability documents from Russian plants; and (4) comparison of transparency data with declared processing data. In the interest of protecting sensitive information, the monitoring is neither full time nor invasive. Thus, an element of trust is required regarding declared operations that are not observed. U.S. transparency monitoring data and independent instrument measurements are compared with plant accountability records and other declared processing data to provide assurance that the nonproliferation objectives of the 1993 Agreement are being met. Similarly, Russian monitoring of U. S. storage and fuel fabrication operations provides assurance to the Russians that the derived LEU is being used in accordance with the Agreement. The successful implementation of the Transparency program enables the receipt of Russian origin LEU into the United States. Implementation of the 1993 Agreement is proceeding on schedule, with the permanent elimination of over 8,700 warhead equivalents of HEU. The successful implementation of the Transparency program has taken place over the last 10 years and has provided the

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)

A novel monolithic LEU foil target based on a PVD manufacturing process for 99Mo production via fission.

Science.gov (United States)

Hollmer, Tobias; Petry, Winfried

2016-12-01

99 Mo is the most widely used radioactive isotope in nuclear medicine. Its main production route is the fission of uranium. A major challenge for a reliable supply is the conversion from highly enriched uranium (HEU) to low enriched uranium (LEU). A promising candidate to realize this conversion is the cylindrical LEU irradiation target. The target consists of a uranium foil encapsulated between two coaxial aluminum cladding cylinders. This target allows a separate processing of the irradiated uranium foil and the cladding when recovering the 99 Mo. Thereby, both the costs and the volume of highly radioactive liquid waste are significantly reduced compared to conventional targets. The presented manufacturing process is based on the direct coating of the uranium on the inside of the outer cladding cylinder. This process was realized by a cylindrical magnetron enhanced physical vapor deposition (PVD) technique. The method features a highly automated process, a good quality of the resulting uranium foils and a high material utilization. Copyright © 2016 Elsevier Ltd. All rights reserved.

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

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

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

Performance and Fabrication Status of TREAT LEU Conversion Conceptual Design Concepts

Energy Technology Data Exchange (ETDEWEB)

IJ van Rooyen; SR Morrell; AE Wright; E. P Luther; K Jamison; AL Crawford; HT III Hartman

2014-10-01

Resumption of transient testing at the TREAT facility was approved in February 2014 to meet U.S. Department of Energy (DOE) objectives. The National Nuclear Security Administration’s Global Threat Reduction Initiative Convert Program is evaluating conversion of TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU). This paper describes briefly the initial pre-conceptual designs screening decisions with more detailed discussions on current feasibility, qualification and fabrication approaches. Feasible fabrication will be shown for a LEU fuel element assembly that can meet TREAT design, performance, and safety requirements. The statement of feasibility recognizes that further development, analysis, and testing must be completed to refine the conceptual design. Engineering challenges such as cladding oxidation, high temperature material properties, and fuel block fabrication along with neutronics performance, will be highlighted. Preliminary engineering and supply chain evaluation provided confidence that the conceptual designs can be achieved.

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

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

Neutronics analysis of the proposed 25-MW leu TRIGA Multipurpose Research Reactor

International Nuclear Information System (INIS)

Nurdin, M.; Bretscher, M.M.; Snelgrove, J.L.

1982-01-01

More than two years ago the government of Indonesia announced plans to purchase a research reactor for the Puspiptek Research Center in Serpong Indonesia to be used for isotope production, materials testing, neutron physics measurements, and reactor operator training. Reactors using low-enriched uranium (LEU) plate-type and rod-type fuel elements were considered. This paper deals with the neutronic evaluation of the rod-type 25-MW LEU TRIGA Multipurpose Research Reactor (MPRR) proposed by the General Atomic Company of the United States of America

Waste Treatment of Acidic Solutions from the Dissolution of Irradiated LEU Targets for 99-Mo Production

Energy Technology Data Exchange (ETDEWEB)

Bakel, Allen J. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Conner, Cliff [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Quigley, Kevin [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Vandegrift, George F. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

2016-10-01

One of the missions of the Reduced Enrichment for Research and Test Reactors (RERTR) program (and now the National Nuclear Security Administrations Material Management and Minimization program) is to facilitate the use of low enriched uranium (LEU) targets for ⁹⁹Mo production. The conversion from highly enriched uranium (HEU) to LEU targets will require five to six times more uranium to produce an equivalent amount of ⁹⁹Mo. The work discussed here addresses the technical challenges encountered in the treatment of uranyl nitrate hexahydrate (UNH)/nitric acid solutions remaining after the dissolution of LEU targets. Specifically, the focus of this work is the calcination of the uranium waste from ⁹⁹Mo production using LEU foil targets and the Modified Cintichem Process. Work with our calciner system showed that high furnace temperature, a large vent tube, and a mechanical shield are beneficial for calciner operation. One- and two-step direct calcination processes were evaluated. The high-temperature one-step process led to contamination of the calciner system. The two-step direct calcination process operated stably and resulted in a relatively large amount of material in the calciner cup. Chemically assisted calcination using peroxide was rejected for further work due to the difficulty in handling the products. Chemically assisted calcination using formic acid was rejected due to unstable operation. Chemically assisted calcination using oxalic acid was recommended, although a better understanding of its chemistry is needed. Overall, this work showed that the two-step direct calcination and the in-cup oxalic acid processes are the best approaches for the treatment of the UNH/nitric acid waste solutions remaining from dissolution of LEU targets for ⁹⁹Mo production.

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)

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)

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

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)

Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA{sup ®} Reactor

Energy Technology Data Exchange (ETDEWEB)

Schickler, R.A., E-mail: robert.schickler@oregonstate.edu; Marcum, W.R., E-mail: wade.marcum@oregonstate.edu; Reese, S.R.

2013-09-15

Highlights: • The Oregon State TRIGA{sup ®} Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA{sup ®} Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least

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)

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

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.

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

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

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

Selection of LEU/Th reference fuel for the HTGR-SC/C lead plant

International Nuclear Information System (INIS)

Turner, R.F.; Neylan, A.J.; Baxter, A.M.; McEachern, D.W.; Stansfield, O.M.

1983-05-01

This paper describes the reference fuel materials for the high-temperature gas-cooled reactor (HTGR) plant for steam cycle/cogeneration (SC/C). A development and testing program carried out in 1978 through 1982 led to the selection of coated fuel particles of uranium-oxycarbide (UCO) for fissile materials and thorium oxide (ThO 2 ) for fertiel materials. Low-enriched uranium (LEU) is the enrichment basis for the HTGR-SC/C application. While UC 2 and UO 2 would also meet the essential criteria for fissile fuel, the UCO, alternative was selected on the basis of improved performance, economics, and process conditions

White Paper – Use of LEU for a Space Reactor

Energy Technology Data Exchange (ETDEWEB)

Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-08-11

Historically space reactors flown or designed for the U.S. and Russia used Highly Enriched Uranium (HEU) for fuel. HEU almost always produces a small and lighter reactor. Since mass increases launch costs or decreases science payloads, HEU was the natural choice. However in today’s environment, the proliferation of HEU has become a major concern for the U.S. government and hence a policy issue. In addition, launch costs are being reduced as the space community moves toward commercial launch vehicles. HEU also carries a heavy security cost to process, test, transport and launch. Together these issues have called for a re-investigation into space reactors the use Low Enriched Uranium (LEU) fuel.

Foreign research reactor uranium supply program: The Y-12 national security complex process

International Nuclear Information System (INIS)

Nelson, T.; Eddy, B.G.

2010-01-01

The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the HEU Disposition Program, the Reduced Enrichment Research and Test Reactors (RERTR) Program, and the United States FRR Spent Nuclear Fuel (SNF) Acceptance Program. The Y-12 National Nuclear Security Administration (NNSA) Y-12 Site Office maintains the prime contracts with foreign governments for the supply of Low-Enriched Uranium (LEU) for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. This program supports the important U.S. government and nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to LEU fuel under the guidance of the NNSA RERTR Program. In conjunction with the FRR SNF Acceptance Program which supports the global nonproliferation efforts to disposition U.S.-origin HEU, the Y-12 FRR Uranium Supply Program can provide the LEU for the replacement fuel fabrication. In addition to feedstock for fuel fabrication, Y-12 supplies LEU for target fabrication for medical isotope production. The Y-12 process uses supply forecasting tools, production improvements and efficient delivery preparations to successfully support the global research reactor community

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

Uranium management activities

International Nuclear Information System (INIS)

Jackson, D.; Marshall, E.; Sideris, T.; Vasa-Sideris, S.

2001-01-01

One of the missions of the Department of Energy's (DOE) Oak Ridge Office (ORO) has been the management of the Department's uranium materials. This mission has been accomplished through successful integration of ORO's uranium activities with the rest of the DOE complex. Beginning in the 1980's, several of the facilities in that complex have been shut down and are in the decommissioning process. With the end of the Cold War, the shutdown of many other facilities is planned. As a result, inventories of uranium need to be removed from the Department facilities. These inventories include highly enriched uranium (HEU), low enriched uranium (LEU), normal uranium (NU), and depleted uranium (DU). The uranium materials exist in different chemical forms, including metals, oxides, solutions, and gases. Much of the uranium in these inventories is not needed to support national priorities and programs. (author)

Fuel Management Strategies for a Possible Future LEU Core of a TRIGA Mark II Vienna

Energy Technology Data Exchange (ETDEWEB)

Khan, R.; Villa, M.; Steinhauser, G.; Boeck, H. [Vienna University of Technology-Atominstitut (Austria)

2011-07-01

The Vienna University of Technology/Atominstitut (VUT/ATI) operates a TRIGA Mark II research reactor. It is operated with a completely mixed core of three different types of fuel. Due to the US fuel return program, the ATI have to return its High Enriched Uranium (HEU) fuel latest by 2019. As an alternate, the Low Enrich Uranium (LEU) fuel is under consideration. The detailed results of the core conversion study are presented at the RRFM 2011 conference. This paper describes the burn up calculations of the new fuel to predict the future burn up behavior and core life time. It also develops an effective and optimized fuel management strategy for a possible future operation of the TRIGA Mark II with a LEU core. This work is performed by the combination of MCNP5 and diffusion based neutronics code TRIGLAV. (author)

Russia ends pact to curb uranium use

Science.gov (United States)

Allen, Michael

2016-11-01

The Russian government has terminated an agreement between the country's nuclear body, Rosatom, and the US Department of Energy (DOE) into the feasibility of converting research reactors in Russia to low-enriched uranium (LEU).

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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.)

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.)

TRANSPARENCY: Tracking Uranium under the U.S./Russian HEU Purchase Agreement

International Nuclear Information System (INIS)

Benton, J B; Decman, D J; Leich, D A

2005-01-01

By the end of August, 2005, the Russia Federation delivered to the United States (U.S.) more than 7,000 metric tons (MT) of low enriched uranium (LEU) containing approximately 46 million SWU and 75,000 MT of natural uranium. This uranium was blended down from weapons-grade (nominally enriched to 90% 235 U) highly enriched uranium (HEU) under the 1993 HEU Purchase Agreement that provides for the blend down of 500 MT HEU into LEU for use as fuel in commercial nuclear reactors. The HEU Transparency Program, under the National Nuclear Security Administration (NNSA), monitored the conversion and blending of the more than 250 MT HEU used to produce this LEU. The HEU represents more than half of the 500 MT HEU scheduled to be blended down through the year 2013 and is equivalent to the elimination of more than 10,000 nuclear devices. The HEU Transparency Program has made considerable progress in its mission to develop and implement transparency measures necessary to assure that Russian HEU extracted from dismantled Russian nuclear weapons is blended down into LEU for delivery to the United States. U.S. monitor observations include the inventory of inprocess containers, observation of plant operations, nondestructive assay measurements to determine 235 U enrichment, as well as the examination of Material Control and Accountability (MC and A) documents. During 2005, HEU Transparency Program personnel will conduct 24 Special Monitoring Visits (SMVs) to four Russian uranium processing plants, in addition to staffing a Transparency Monitoring Office (TMO) at one Russian site

Heat-transfer analysis of the existing HEU and proposed LEU cores of Pakistan research reactor

International Nuclear Information System (INIS)

Khan, L.A.; Nabbi, R.

1987-02-01

In connection with conversion of Pakistan Research Reactor (PARR) from the use of Highly Enriched Uranium (HEU) fuel to the use of Low Enriched Uranium (LEU) fuel, steady-state thermal hydraulic analysis of both existing HEU and proposed LEU cores has been carried out. Keeping in mind the possibility of power upgrading, the performance of proposed LEU core, under 10 MW operating conditions, has also been evaluated. Computer code HEATHYD has been used for this purpose. In order to verify the reliability of the code, IAEA benchmark 2 MW reactor was analyzed. The cooling parameters evaluated include: coolant velocity, critical velocity, pressure drop, temperature distribution in the core, heat fluxes at onset of nucleate boiling, flow instability and burnout and corresponding safety margins. From the results of the study it can be concluded that the conversion of the core to LEU fuel will result in higher safety margins, as compared to existing HEU core, mainly because the increased number of fuel plates in the proposed design will reduce the average heat flux significantly. Anyhow upgrading of the reactor power to 10 MW will need the flow rate to be adjusted between 850 to 900 m 3 /hr, to achieve reasonable safety margins, at least, comparable with the existing HEU core. (orig.)

A conversion development program to LEU targets for medical isotope production in the MAPLE Facilities

International Nuclear Information System (INIS)

Malkoske, G.R.

2000-01-01

Historically, the production of molybdenum-99 in the NRU research reactors at Chalk River, Canada has been extracted from reactor targets employing highly enriched uranium (HEU). The molybdenum extraction process from the HEU targets provided predictable, consistent yields for our high-volume molybdenum production process. A reliable supply of HEU for the NRU research reactor targets has enabled MDS Nordion to develop a secure chain of medical isotope supply for the international nuclear medicine community. Each link of the isotope supply chain, from isotope production to patient application, has been established on a proven method of HEU target irradiation and processing. To ensure a continued reliable and timely supply of medical isotopes, the design of the MAPLE facilities was based on our established process - extraction of isotopes from HEU target material. However, in concert with the global trend to utilize low enriched uranium (LEU) in research reactors, MDS Nordion has launched a program to convert the MAPLE facilities to LEU targets. An initial feasibility study was initiated to identify the technical issues to convert the MAPLE targets from HEU to LEU. This paper will present the results of the feasibility study. It will also describe future challenges and opportunities in converting the MAPLE facilities to LEU targets for large scale, commercial medical isotope production. (author)

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

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.

Conceptual designs parameters for MURR LEU U-Mo fuel conversion design demonstration experiment. Revision 1

International Nuclear Information System (INIS)

Stillman, J.; Feldman, E.; Stevens, J.

2013-01-01

The design parameters for the conceptual design of a fuel assembly containing U-10Mo fuel foils with low-enriched uranium (LEU) for the University of Missouri Research Reactor (MURR) are described. The Design Demonstration Experiment (MURR-DDE) will use a prototypic MURR-LEU element manufactured according to the parameters specified here. Also provided are calculated performance parameters for the LEU element in the MURR, and a set of goals for the MURR-DDE related to those parameters. The conversion objectives are to develop a fuel element design that will ensure safe reactor operations, as well as maintaining existing performance. The element was designed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. A set of manufacturing assumptions were provided by the Fuel Development (FD) and Fuel Fabrication Capability (FFC) pillars of the GTRI Reduced Enrichment for Research and Test Reactors (RERTR) program to reliably manufacture the fuel plates. The proposed LEU fuel element has an overall design and exterior dimensions that are similar to those of the current highly-enriched uranium (HEU) fuel elements. There are 23 fuel plates in the LEU design. The overall thickness of each plate is 44 mil, except for the exterior plate that is furthest from the center flux trap (plate 23), which is 49 mil thick. The proposed LEU fuel plates have U-10Mo monolithic fuel foils with a 235U enrichment of 19.75% varying from 9 mil to 20 mil thick, and clad with Al-6061 aluminum. A thin layer of zirconium exists between the fuel foils and the aluminum as a diffusion barrier. The thinnest nominal combined zirconium and aluminum clad thickness on each side of the fuel plates is 12 mil. The LEU U-10Mo monolithic fuel is not yet qualified as driver fuel in research reactors, but is under intense development under the auspices of the GTRI FD and FFC programs.

Criticality of moderated and undermoderated low-enriched uranium oxide systems

International Nuclear Information System (INIS)

Goebel, G.R.

1980-06-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Reduced enrichment for research and test reactors: Proceedings

Energy Technology Data Exchange (ETDEWEB)

1988-05-01

The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

Reduced enrichment for research and test reactors: Proceedings

International Nuclear Information System (INIS)

1988-05-01

The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm 3 was by then in routine use, illustrated how far work has progressed

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.)

A neutronic feasibility study for LEU conversion of the SAFARI-1 reactor

International Nuclear Information System (INIS)

Pond, R.B.; Hanan, N.A.; Matos, J.E.; Ball, G.

2000-01-01

A neutronic feasibility study to convert the SAFARI-1 reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with NECSA. Comparisons were made of the reactor performance with the current 90% enriched HEU fuel type (UAl) and two 19.75% enriched LEU fuel types (U 3 Si 2 and U7Mo). The thermal fluxes with the LEU fuels were 3 - 9% lower than with the current HEU fuel. For the same fuel assembly design, a uranium density of approximately 4.5 g/cm 3 was required with U 3 Si 2 -Al fuel and a uranium density of about 4.6 g/cm 3 was required with U7Mo-Al fuel to match the 24.6-day cycle of the UAl-alloy fuel with 0.92 gU/cm 3 . The selection of a suitable LEU fuel and the decision to convert SAFARI-1 will be an economic matter that depends upon the fuel type, fuel assembly design, experiment performance and fuel cycle costs. (author)

HEU benchmark calculations and LEU preliminary calculations for IRR-1

International Nuclear Information System (INIS)

Caner, M.; Shapira, M.; Bettan, M.; Nagler, A.; Gilat, J.

2004-01-01

We performed neutronics calculations for the Soreq Research Reactor, IRR-1. The calculations were done for the purpose of upgrading and benchmarking our codes and methods. The codes used were mainly WIMS-D/4 for cell calculations and the three dimensional diffusion code CITATION for full core calculations. The experimental flux was obtained by gold wire activation methods and compared with our calculated flux profile. The IRR-1 is loaded with highly enriched uranium fuel assemblies, of the plate type. In the framework of preparation for conversion to low enrichment fuel, additional calculations were done assuming the presence of LEU fresh fuel. In these preliminary calculations we investigated the effect on the criticality and flux distributions of the increase of U-238 loading, and the corresponding uranium density.(author)

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

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

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

The development of uranium foil farication technology utilizing twin roll method for Mo-99 irradiation target

CERN Document Server

Kim, C K; Park, H D

2002-01-01

MDS Nordion in Canada, occupying about 75% of global supply of Mo-99 isotope, has provided the irradiation target of Mo-99 using the rod-type UAl sub x alloys with HEU(High Enrichment Uranium). ANL (Argonne National Laboratory) through co-operation with BATAN in Indonesia, leading RERTR (Reduced Enrichment for Research and Test Reactors) program substantially for nuclear non-proliferation, has designed and fabricated the annular cylinder of uranium targets, and successfully performed irradiation test, in order to develop the fabrication technology of fission Mo-99 using LEU(Low Enrichment Uranium). As the uranium foils could be fabricated in laboratory scale, not in commercialized scale by hot rolling method due to significant problems in foil quality, productivity and economic efficiency, attention has shifted to the development of new technology. Under these circumstances, the invention of uranium foil fabrication technology utilizing twin-roll casting method in KAERI is found to be able to fabricate LEU or...

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)

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

Irradiation tests of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.; Suripto, A.; Nasution, H.; Lufti-Amin, D.; Gogo, A.

1996-01-01

Most of the world's supply of 99 mTc for medical purposes is currently produced from the decay of 99 Mo derived from the fissioning of high-enriched uranium (HEU). Substitution of low-enriched uranium (LEU) metal foils for the HEU UO 2 used in current target designs will allow equivalent 99 Mo yields with little change in target geometries. Substitution of uranium metal for uranium alloy and aluminide in other target designs will also allow the conversion of HEU to LEU. Several uranium-metal-foil targets have been fabricated at ANL and irradiated to prototypic burnup in the Indonesian RSG-GAS reactor. Postirradiation examination of the initial test indicated that design modifications were required to allow the irradiated foil to be removed for chemical processing. The latest test has shown good irradiation behavior, satisfactory dismantling and foil removal when the U-foil is separated from its containment by metallic, fission-recoil absorbing barriers. (author)

Irradiation tests of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

Hofman, G.L.; Wiencek, T.C.; Wood, E.L.; Snelgrove, J.L.; Suripto, A.; Nasution, H.; Lufti-Amin, D.; Gogo, A.

1996-01-01

Most of the world's supply of 99m Tc for medical purposes is currently produced form the decay of 99 Mo derived from the fissioning of high-enriched uranium (HEU). Substitution of low-enriched uranium (LEU) metal foils for the HEU UO 2 used in current target designs will allow equivalent 99 Mo yields with little change in target geometries. Substitution of uranium metal for uranium alloy and aluminide in other target designs will also allow the conversion of HEU to LEU. Several uranium-metal-foil targets have been fabricated at ANL and irradiated to prototypic burnup in the Indonesian RSG-GAS reactor. Postirradiation examination of the initial test indicated that design modifications were required to allow the irradiated foil to be removed for chemical processing. The latest test has shown good irradiation behavior, satisfactory dismantling and foil removal when the U-foil is separated from its containment by metallic, fission-recoil absorbing barriers

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

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)

The LEU target development and conversion program for the MAPLE reactors and new processing facility

International Nuclear Information System (INIS)

Malkoske, G.R.

2002-01-01

Historically, the production of molybdenum-99 in the NRU research reactors at Chalk River, Canada has been extracted from reactor targets employing highly enriched uranium (HEU). A reliable supply of HEU metal from the United States used in the manufacture of targets for the NRU research reactor has been a key factor to enable MDS Nordion to develop a secure supply of medical isotopes for the international nuclear medicine community. The molybdenum extraction process from HEU targets provides predictable, consistent yields for our high-volume molybdenum production process. Each link of the isotope supply chain, from isotope production to ultimate use by the physician, has been established using this proven and established method of HEU target irradiation and processing to extract molybdenum-99. To ensure a continued reliable and timely supply of medical isotopes, MDS Nordion is completing the construction of two MAPLE reactors and a New Processing Facility. The design of the MAPLE facilities was based on an established process developed by Atomic Energy of Canada Ltd. (AECL) - extraction of isotopes from HEU target material. However, in concert with the global trend to utilize low enriched uranium (LEU) in research reactors, MDS Nordion has launched a three phase LEU Target Development and Conversion Program for the MAPLE facilities. Phase 1, the Initial Feasibility Study, which identified the technical issues to convert the MAPLE reactor targets from HEU to LEU for large scale commercial production was reported on at the RERTR- 2000 conference. The second phase of the LEU Target Development and Conversion Program was developed with extensive consultation and involvement of experts knowledgeable in target development, process system design, enriched uranium conversion chemistry and commercial scale reactor operations and molybdenum production. This paper will provide an overview of the Phase 2 Conversion Development Program, report on progress to date, and further

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

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)

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

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

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

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.)

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

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)

Progress in chemical treatment of LEU targets by the modified Cintichem process

International Nuclear Information System (INIS)

Wu, D.; Landsberger, S.; Vandegrift, G.F.

1996-01-01

Presented here are recent experimental results on tests of a modified Cintichem process for producing 99 Mo from low enriched uranium (LEU). Studies were focused in three areas: (1) testing the effects on 99 Mo recovery and purity of dissolving LEU foil in nitric acid alone, rather than in the sulfuric/nitric acid mixture currently used, (2) measuring decontamination factors for radionuclide impurities in each purification step, and (3) testing the effects on processing of adding barrier materials to the LEU metal-foil target. The experimental results show that switching from dissolving the target in the sulfuric/nitric mixture to using nitric acid alone should cause no significant difference in 99 Mo product yield or purity. Further, the results show that overall decontamination factors for gamma emitters in the LEU-target processing are high enough to meet the purity requirements for the 99 Mo product. The results also show that the selected barrier materials, Cu, Fe, and Ni, do not interfere with 99 Mo recovery and can be removed during chemical processing of the LEU target

The University of Missouri Research Reactor HEU to LEU conversion project status

Energy Technology Data Exchange (ETDEWEB)

McKibben, James C; Kutikkad, Kiratadas; Foyto, Leslie P; Peters, Nickie J; Solbrekken, Gary L; Kennedy, John [University of Missouri Research Reactor, Missouri (United States); Stillman, John A; Feldman, Earl E; Tzanos, Constantine P; Stevens, John G [Argonne National Laboratory, Argonne, Illinois (United States)

2012-03-15

The University of Missouri Research Reactor (MURR) is one of five U.S. high performance research and test reactors that are actively collaborating with the U.S. Department of Energy (DOE) to find a suitable low-enriched uranium (LEU) fuel replacement for the currently required highly-enriched uranium (HEU) fuel. A conversion feasibility study based on U-10Mo monolithic LEU fuel was completed in 2009. It was concluded that the proposed LEU fuel assembly design, in conjunction with an increase in power level from 10 to 12 MWth, will (1) maintain safety margins during operation, (2) allow operating fuel cycle lengths to be maintained for efficient and effective use of the facility, and (3) preserve an acceptable level and spectrum of key neutron fluxes to meet the scientific mission of the facility. The MURR and Argonne National Laboratory (ANL) team is continuing to work toward realization of the conversion. The 'Preliminary Safety Analysis Report Methodologies and Scenarios for LEU Conversion of MURR' was completed in June 2011. This report documents design parameter values critical to the Fuel Development (FD), Fuel Fabrication Capability (FFC) and Hydromechanical Fuel Test Facility (HMFTF) projects. The report also provides a preliminary evaluation of safety analysis techniques and data that will be needed to complete the fuel conversion Safety Analysis Report (SAR), especially those related to the U-10Mo monolithic LEU fuel. Specific studies are underway to validate the proposed path to an LEU fuel conversion. Coupled fluid-structure simulations and experiments are being conducted to understand the hydrodynamic plate deformation risk for 0.965 mm (38 mil) thick fuel plates. Methodologies that were recently developed to answer the U.S. Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI) regarding the MURR 2006 relicensing submittal will be used in the LEU conversion effort. Transition LEU fuel elements that will have a minimal impact on

Neutronic calculations of PARR-1 cores using LEU silicide fuel

International Nuclear Information System (INIS)

Arshad, M.; Bakhtyar, S.; Hayat, T.; Salahuddin, A.

1991-08-01

Detailed neutronic calculations have been carried out for different PARR-1 cores utilizing low enriched uranium (LEU) silicide fuel and operating at an upgraded power of 9 MW. The calculations include the search for critical loadings in open and stall ends of the pool, neutronic analysis of the first full equilibrium core and calculations cores. The burnup study of inventory have also been carried out. Further, the reactivity coefficients of the first full power operation core are evaluated for use in the accident analysis. 14 figs. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1992-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1992-12-01

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

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.)

Preliminary design study for a carbide LEU-nuclear thermal rocket

International Nuclear Information System (INIS)

Venneri, P.F.; Kim, Y.

2014-01-01

Nuclear space propulsion is a requirement for the successful exploration of the solar system. It offers the possibility of having both a high specific impulse and a relatively high thrust, allowing rapid transit times with a minimum usage of fuel. This paper proposes a nuclear thermal rocket design based on heritage NERVA rockets that makes use of Low Enriched Uranium (LEU) fuel. The Carbide LEU Nuclear Thermal Rocket (C-LEU-NTR) is designed to fulfill the rocket requirements as set forth in the NASA 2009 Mars Mission Design Reference Architecture 5.0, that is provide 25,000 lbf of thrust, operate at full power condition for at least two hours, and have a specific impulse close to 900 s. The neutronics analysis was done using MCNP5 with the ENDF/B-VII.1 neutron library. The thermal hydraulic calculations and size optimization were completed with a finite difference code being developed at the Center for Space Nuclear Research. (authors)

Assessment of the effectiveness of the LEU Reform Rule and its implementation

International Nuclear Information System (INIS)

Moran, B.W.; Nations, J.O.; Hammond, G.A.

1993-11-01

The US Nuclear Regulatory Commission (NRC) amended its material control and accounting (MC ampersand A) requirements in 1985 for licensees possessing and using special nuclear material (SNM) of low strategic significance in quantities larger than one effective kilogram (kg). The goal of the Low-Enriched Uranium (LEU) Reform Rule (i.e., 10CFR 74.31) was to establish MC ampersand A requirements for the LEU licensees at a level consistent with the safeguards risk associated with the relatively low strategic importance of such material. The amended requirements were written in a performance-oriented manner, rather than a prescriptive one, in an effort to allow the licensees the opportunity to choose the most cost-effective means of satisfying the requirements. The LEU Reform Rule was implemented in January 1988 and the fuel cycle facilities have had sufficient experience in implementing the rule to allow a meaningful review of its effectiveness. This document provides technical analysis and recommendations to assist the NRC in making a determination if the rule is achieving its intended purpose, and if not, to make the necessary changes to accomplish this

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

International Nuclear Information System (INIS)

Tuck, G.; Oh, I.

1979-08-01

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

The whole-core LEU silicide fuel demonstration in the JMTR

Energy Technology Data Exchange (ETDEWEB)

Aso, Tomokazu; Akashi, Kazutomo; Nagao, Yoshiharu [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)] [and others

1997-08-01

The JMTR was fully converted to LEU silicide (U{sub 3}Si{sub 2}) fuel with cadmium wires as burnable absorber in January, 1994. The reduced enrichment program for the JMTR was initiated in 1979, and the conversion to MEU (enrichment ; 45%) aluminide fuel was carried out in 1986 as the first step of the program. The final goal of the program was terminated by the present LEU conversion. This paper describes the results of core physics measurement through the conversion phase from MEU fuel core to LEU fuel core. Measured excess reactivities of the LEU fuel cores are mostly in good agreement with predicted values. Reactivity effect and burnup of cadmium wires, therefore, were proved to be well predicted. Control rod worth in the LEU fuel core is mostly less than that in the MEU fuel core. Shutdown margin was verified to be within the safety limit. There is no significant difference in temperature coefficient of reactivity between the MEU and LEU fuel cores. These results verified that the JMTR was successfully and safely converted to LEU fuel. Extension of the operating cycle period was achieved and reduction of spend fuel elements is expected by using the fuel with high uranium density.

An update on the LEU target development and conversion program for the MAPLE reactors and new processing facility

International Nuclear Information System (INIS)

Malkoske, G.R.; Eng, B.Sc; Eng, P.

2002-01-01

Historically, the production of molybdenum-99 in the NRU research reactors at Chalk River, Canada, has been extracted from reactor targets employing highly enriched uranium (HEU). A reliable supply of HEU metal from the United States used in the manufacture of targets for the NRU research reactor has been a key factor to enable MDS Nordion to develop a secure supply of medical isotopes for the international nuclear medicine community. The molybdenum extraction process from HEU targets provides predictable, consistent yields for our high-volume molybdenum production process. Each link of the isotope supply chain, from isotope production to ultimate use by the physician, has been established using this proven and established method of HEU target irradiation and processing to extract molybdenum-99. To ensure a continued reliable and timely supply of medical isotopes, MDS Nordion is completing the construction of two MAPLE reactors and a New Processing Facility. The design of the MAPLE facilities was based on an established process developed by Atomic Energy of Canada Ltd. (AECL)-extraction of isotopes from HEU target material. However, in concert with the global trend to utilize low enriched uranium (LEU) in research reactors, MDS Nordion has launched a three phase LEU Target Development and Conversion Program for the MAPLE facilities. Phase 1, the Initial Feasibility Study, which identified the technical issues to convert the MAPLE reactor targets from HEU to LEU for large scale commercial production was reported on at the RERTR-2000 conference. The second phase of the LEU Target Development and Conversion Program was developed with extensive consultation and involvement of experts knowledgeable in target development, process system design, enriched uranium conversion chemistry and commercial scale reactor operations and molybdenum production. This paper will provide an overview of the Phase 2 Conversion Development Program, report on progress to date, and further

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.)

HEU to LEU conversion and blending facility: UNH blending alternative to produce LEU oxide for disposal

International Nuclear Information System (INIS)

1995-09-01

The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than is the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This report provides data to be used in the environmental impact analysis for the uranyl nitrate hexahydrate blending option to produce oxide for disposal. This the Conversion and Blending Facility (CBF) alternative will have two missions (1) convert HEU materials into HEU uranyl nitrate (UNH) and (2) blend the HEU uranyl nitrate with depleted and natural assay uranyl nitrate to produce an oxide that can be stored until an acceptable disposal approach is available. 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

HEU to LEU conversion and blending facility: UNH blending alternative to produce LEU oxide for disposal

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than is the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This report provides data to be used in the environmental impact analysis for the uranyl nitrate hexahydrate blending option to produce oxide for disposal. This the Conversion and Blending Facility (CBF) alternative will have two missions (1) convert HEU materials into HEU uranyl nitrate (UNH) and (2) blend the HEU uranyl nitrate with depleted and natural assay uranyl nitrate to produce an oxide that can be stored until an acceptable disposal approach is available. 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.

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

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

Science.gov (United States)

2013-11-07

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

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...

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

Main results and status of the development of LEU fuel for Russian research reactors

International Nuclear Information System (INIS)

Vatulin, A.; Morozov, A.; Suprun, V.; Dobrikova, I.

2005-01-01

VNIINM develops low enrichment uranium (LEU) fuel on base U-Mo alloys and a novel design of pin-type fuel elements. The development is carried out both for existing reactors, and for new advanced designs of reactors. The work is carried on the following main directions: - irradiate LEU U-Mo dispersion fuel (the uranium density up to 6,0 g/cm 3 ) in two Russian research reactors: MIR (RIAR, Dimitrovgrad) as pin type fuel mini-elements and in WWR-M (PINP, Gatchina) within full-scaled fuel assembly (FA) with pin type fuel elements; - finalize development of design and fabrication process of IRT type FA with pin type fuel elements; - develop methods of reducing of U-Mo fuel --Al matrix interaction under irradiation; - develop fabricating methods of fuel elements on base of monolithic U-Mo fuel. The paper generally reviews the results of calculation, design and technology investigations accomplished by now. (author)

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.)

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...

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)

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

Neutronic analysis for conversion of the Ghana Research Reactor-1 facility using Monte Carlo methods and UO{sub 2} LEU fuel

Energy Technology Data Exchange (ETDEWEB)

Anim-Sampong, S.; Akaho, E.H.K.; Maakuu, B.T.; Gbadago, J.K. [Ghana Research Reactor-1 Centre, Dept. of Nuclear Engineering and Materials Science, National Nuclear Research Institute, Ghana Atomic Energy Commission, Legon, Accra (Ghana); Andam, A. [Kwame Nkrumah Univ. of Science and Technology, Dept. of Physics (Ghana); Liaw, J.J.R.; Matos, J.E. [Argonne National Lab., RERTR Programme, Div. of Nuclear Engineering (United States)

2007-07-01

Monte Carlo particle transport methods and software (MCNP) have been applied to the modelling, simulation and neutronic analysis for the conversion of the HEU-fuelled (high enrichment uranium) core of the Ghana Research Reactor-1 (GHARR-1) facility. The results show that the MCNP model of the GHARR-1 facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR) is good as the simulated neutronic and other reactor physics parameters agree with very well with experimental and zero power results. Three UO{sub 2} LEU (low enrichment uranium) fuels with different enrichments (12.6% and 19.75%), core configurations, core loadings were utilized in the conversion studies. The nuclear criticality and kinetic parameters obtained from the Monte Carlo simulation and neutronic analysis using three UO{sub 2} LEU fuels are in close agreement with results obtained for the reference 90.2% U-Al HEU core. The neutron flux variation in the core, fission chamber and irradiation channels for the LEU UO{sub 2} fuels show the same trend as the HEU core as presented in the paper. The Monte Carlo model confirms a reduction (8% max) in the peak neutron fluxes simulated in the irradiation channels which are utilized for experimental and commercial activities. However, the reductions or 'losses' in the flux levels neither affects the criticality safety, reactor operations and safety nor utilization of the reactor. Employing careful core loading optimization techniques and fuel loadings and enrichment, it is possible to eliminate the apparent reductions or 'losses' in the neutron fluxes as suggested in this paper. Concerning neutronics, it can be concluded that all the 3 LEU fuels qualify as LEU candidates for core conversion of the GHARR-1 facility.

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.)

Innovative nuclear thermal rocket concept utilizing LEU fuel for space application

International Nuclear Information System (INIS)

Nam, Seung Hyun; Venneri, Paolo; Choi, Jae Young; Jeong, Yong Hoon; Chang, Soon Heung

2015-01-01

Space is one of the best places for humanity to turn to keep learning and exploiting. A Nuclear Thermal Rocket (NTR) is a viable and more efficient option for human space exploration than the existing Chemical Rockets (CRs) which are highly inefficient for long-term manned missions such as to Mars and its satellites. NERVA derived NTR engines have been studied for the human missions as a mainstream in the United States of America (USA). Actually, the NERVA technology has already been developed and successfully tested since 1950s. The state-of-the-art technology is based on a Hydrogen gas (H_2) cooled high temperature reactor with solid core utilizing High-Enriched Uranium (HEU) fuel to reduce heavy metal mass and to use fast or epithermal neutron spectrums enabling simple core designs. However, even though the NTR designs utilizing HEU is the best option in terms of rocket performance, they inevitably provoke nuclear proliferation obstacles on all Research and Development (R and D) activities by civilians and non-nuclear weapon states, and its eventual commercialization. To surmount the security issue to use HEU fuel for a NTR, a concept of the innovative NTR engine, Korea Advanced NUclear Thermal Engine Rocket utilizing Low-Enriched Uranium fuel (KANUTER-LEU) is presented in this paper. The design goal of KANUTER-LEU is to make use of a LEU fuel for its compact reactor, but does not sacrifice the rocket performance relative to the traditional NTRs utilizing HEU. KANUTER-LEU mainly consists of a fission reactor utilizing H_2 propellant, a propulsion system and an optional Electricity Generation System as a bimodal engine. To implement LEU fuel for the reactor, the innovative engine adopts W-UO_2 CERMET fuel to drastically increase uranium density and thermal neutron spectrum to improve neutron economy in the core. The moderator and structural material selections also consider neutronic and thermo-physical characteristics to reduce non-fission neutron loss and

Criticality Calculations of Fresh LEU and MOX Assemblies for Transport and Storage at the Balakovo Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Goluoglu, S.

2001-01-11

Transportation of low-enriched uranium (LEU) and mixed-oxide (MOX) assemblies to and within the VVER-1000-type Balakovo Nuclear Power Plant is investigated. Effective multiplication factors for fresh fuel assemblies on the railroad platform, fresh fuel assemblies in the fuel transportation vehicle, and fresh fuel assemblies in the spent fuel storage pool are calculated. If there is no absorber between the units, the configurations with all MOX assemblies result in higher effective multiplication factors than the configurations with all LEU assemblies when the system is dry. When the system is flooded, the configurations with all LEU assemblies result in higher effective multiplication factors. For normal operating conditions, effective multiplication factors for all configurations are below the presumed upper subcritical limit of 0.95. For an accident condition of a fully loaded fuel transportation vehicle that is flooded with low-density water (possibly from a fire suppression system), the presumed upper subcritical limit is exceeded by configurations containing LEU assemblies.

Status of the RERTR [Reduced Enrichment Research and Test Reactor] program in Argentina

International Nuclear Information System (INIS)

Giorsetti, D.R.

1987-01-01

The Argentine Atomic Energy Commission started in 1978 the Reduced Enrichment Research and Test Reactors in the field of reactor engineering; engineering, development and manufacturing of fuel elements and research reactors operators. This program was initiated with the conviction that it would contribute to the international efforts to reduce risks of nuclear weapons proliferation owing to an uncontrolled use of highly enriched uranium. It was intended to convert RA-3 reactor to make possible its operation with low enriched fuel (LEU), instead of high enriched fuel (HEU) and to develop manufacturing techniques for said LEU. Afterwards, this program was adapted to assist other countries in reactors conversion, development of the corresponding fuel elements and supply of fuel elements to other countries. (Author)

The NNSA global threat reduction initiative's efforts to minimize the use of highly enriched uranium for medical isotope production

International Nuclear Information System (INIS)

Staples, Parrish

2010-01-01

The mission of the National Nuclear Security Administration's (NNSA) Office of Global Threat Reduction (GTRI) is to reduce and protect vulnerable nuclear and radiological materials located at civilian sites worldwide. GTRI is a key organization for supporting domestic and global efforts to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications. GTRI implements the following activities in order to achieve its threat reduction and HEU minimization objectives: Converting domestic and international civilian research reactors and isotope production facilities from the use of HEU to low enriched uranium (LEU); Demonstrating the viability of medical isotope production technologies that do not use HEU; Removing or disposing excess nuclear and radiological materials from civilian sites worldwide; and Protecting high-priority nuclear and radiological materials worldwide from theft and sabotage. This paper provides a brief overview on the recent developments and priorities for GTRI program activities in 2010, with a particular focus on GTRI's efforts to demonstrate the viability of non-HEU based medical isotope production technologies. (author)

Production, inventories and HEU in the world uranium market: Production's vital role

International Nuclear Information System (INIS)

Underhill, D.H.

1997-01-01

This paper analyses recent uranium supply and demand relationship and projects supply through 2010. The extremely depressed record low market prices have led to the ongoing annual inventory drawdown of over 25,000 t U resulting from the current 45% world production shortfall. The policy of the European Union and anti-dumping related activities in the USA are restricting imports of uranium from CIS producers to a majority of the world's nuclear utilities. These factors are reducing low priced uranium supply and forcing buyers to again obtain more of their requirements from producers. It discusses how the sale of Low Enriched Uranium (LEU) produced from of 550 t High Enriched Uranium (HEU) from Russia and Ukraine could potentially supply about 15% of world requirements through 2010. However, legislation currently being developed by the US Congress may ration the sale of this material, extending the LEU supply well into the next century. Nuclear generation capacity and its uranium requirements are projected to grow at about 1.5% through 2010. Demand for new uranium purchases is however, increasing at the much higher rate of 25-30% over the next 10-15 years. This increasing demand in the face of decreasing supply is resulting in a market recovery in which the spot price for non-CIS produced uranium has risen over 25% since October 1994. Prices will continue to increase as the market equilibrium shifts from a balance with alternative excess low priced supply to an equilibrium between production and demand. 19 refs, 14 figs, 2 tabs

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)

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)

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)

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.

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

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

Energy Technology Data Exchange (ETDEWEB)

Larsen, H.

1974-03-15

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

Preliminary Results of Ancillary Safety Analyses Supporting TREAT LEU Conversion Activities

Energy Technology Data Exchange (ETDEWEB)

Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-10-01

The Transient Reactor Test Facility (TREAT), located at Idaho National Laboratory (INL), is a test facility designed to evaluate the performance of reactor fuels and materials under transient accident conditions. The facility, an air-cooled, graphite-moderated reactor designed to utilize fuel containing high-enriched uranium (HEU), has been in non-operational standby status since 1994. Currently, in support of the missions of the Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program, a new core design is being developed for TREAT that will utilize low-enriched uranium (LEU). The primary objective of this conversion effort is to design an LEU core that is capable of meeting the performance characteristics of the existing HEU core. Minimal, if any, changes are anticipated for the supporting systems (e.g. reactor trip system, filtration/cooling system, etc.); therefore, the LEU core must also be able to function with the existing supporting systems, and must also satisfy acceptable safety limits. In support of the LEU conversion effort, a range of ancillary safety analyses are required to evaluate the LEU core operation relative to that of the existing facility. These analyses cover neutronics, shielding, and thermal hydraulic topics that have been identified as having the potential to have reduced safety margins due to conversion to LEU fuel, or are required to support the required safety analyses documentation. The majority of these ancillary tasks have been identified in [1] and [2]. The purpose of this report is to document the ancillary safety analyses that have been performed at Argonne National Laboratory during the early stages of the LEU design effort, and to describe ongoing and anticipated analyses. For all analyses presented in this report, methodologies are utilized that are consistent with, or improved from, those used in analyses for the HEU Final Safety Analysis

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

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

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.)

Light-water reactors: preliminary safety and environmental information document. Volume I

International Nuclear Information System (INIS)

1980-01-01

Information is presented concerning the reference PWR reactor system; once-through, low-enrichment uranium-235 fuel, 30 MWD per kilogram (PWR LEU(5)-OT); once-through, low-enrichment, high-burnup uranium fuel (PWR LEU(5)-Mod OT); self-generated plutonium spiked recycle (PWR LEU(5)-Pu-Spiked Recycle); denatured uranium-233/thorium cycle (PWR DU(3)-Th Recycle DU(3)); and plutonium/thorium cycle

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

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)

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

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.)

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.

Gas generation during waste treatment of acidic solutions from the dissolution of irradiated LEU targets for ⁹⁹Mo production

Energy Technology Data Exchange (ETDEWEB)

Bakel, Allen J. [Argonne National Lab. (ANL), Argonne, IL (United States); Conner, Cliff [Argonne National Lab. (ANL), Argonne, IL (United States); Quigley, Kevin [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, George F. [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-01-01

The goal of the Reduced Enrichment for Research and Test Reactors Program is to limit the use of high-enriched uranium (HEU) in research and test reactors by substituting low-enriched uranium (LEU) wherever possible. The work reported here documents our work to develop the calcining technologies and processes that will be needed for ⁹⁹Mo production using LEU foil targets and the Modified Cintichem Process. The primary concern with the conversion to LEU from HEU targets is that it would result in a five- to six-fold increase in the total uranium. This increase results in more liquid waste from the process. We have been working to minimize the increase in liquid waste and to minimize the impact of any change in liquid waste. Direct calcination of uranium-rich nitric acid solutions generates NO₂ gas and UO₃ solid. We have proposed two processes for treating the liquid waste from a Modified Cintichem Process with a LEU foil. One is an optimized direct calcination process that is similar to the process currently in use. The other is a uranyl oxalate precipitation process. The specific goal of the work reported here was to characterize and compare the chemical reactions that occur during these two processes. In particular, the amounts and compositions of the gaseous and solid products were of interest. A series of experiments was carried out to show the effects of temperature and the redox potential of the reaction atmosphere. The primary products of the direct calcination process were mixtures of U₃O₈ and UO₃ solids and NO₂ gas. The primary products of the oxalate precipitation process were mixtures of U₃O₈ and UO₂ solid and CO₂ gas. Higher temperature and a reducing atmosphere tended to favor quadrivalent over hexavalent uranium in the solid product. These data will help producers to decide between the two processes. In addition, the data can be used

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

A Very High Uranium Density Fission Mo Target Suitable for LEU Using atomization Technology

Energy Technology Data Exchange (ETDEWEB)

Kim, C. K.; Kim, K. H.; Lee, Y. S.; Ryu, H. J.; Woo, Y. M.; Jang, S. J.; Park, J. M.; Choi, S. J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-05-15

Currently HEU minimization efforts in fission Mo production are underway in connection with the global threat reduction policy. In order to convert HEU to LEU for the fission Mo target, higher uranium density material could be applied. The uranium aluminide targets used world widely for commercial {sup 99}Mo production are limited to 3.0 g-U/cc in uranium density of the target meat. A consideration of high uranium density using the uranium metal particles dispersion plate target is taken into account. The irradiation burnup of the fission Mo target are as low as 8 at.% and the irradiation period is shorter than 7 days. Pure uranium material has higher thermal conductivity than uranium compounds or alloys. It is considered that the degradation by irradiation would be almost negligible. In this study, using the computer code of the PLATE developed by ANL the irradiation behavior was estimated. Some considerations were taken into account to improve the irradiation performance further. It has been known that some alloying elements of Si, Cr, Fe, and Mo are beneficial for reducing the swelling by grain refinement. In the RERTR program recently the interaction problem could be solved by adding a small amount of Si to the aluminum matrix phase. The fabrication process and the separation process for the proposed atomized uranium particles dispersion target were reviewed

Effect of molybdenum addition on metastability of cubic γ-uranium

International Nuclear Information System (INIS)

Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Dey, G.K.; Kamath, H.S.

2010-01-01

Over the years U 3 Si 2 compound dispersed in aluminium matrix has been used successfully as the potential low enriched uranium (LEU 235 ) base dispersion fuel for use in new research and test reactors and also for converting high enriched uranium (HEU > 85%U 235 ) cores to LEU for most of the existing research and test reactors world over, though maximum 4.8 g U cm -3 density is achievable with U 3 Si 2 -Al dispersion fuel. To achieve a uranium density of 8.0-9.0 g U cm -3 in dispersion fuel with aluminium as matrix material, it is required to use γ-stabilized uranium metal powders. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop these high density uranium base alloys. This paper describes the alloying behaviour of uranium with varying amount of molybdenum. The U-Mo alloys with different molybdenum content have been prepared by using an induction melting furnace with uranium and molybdenum metal pellets as starting materials. U-Mo alloys with different molybdenum content were characterized by X-ray diffraction (XRD) for phase identification and lattice parameter measurements. The optical microstructure of different U-Mo alloy composition has also been discussed in this paper. Quantitative image analysis was also carried out to determine the amount of various phases in each composition.

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

Neutronic and thermo-hydraulic design of LEU core for Japan Research Reactor 4

International Nuclear Information System (INIS)

Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

1988-04-01

As a part of the Reduced Enrichment Research and Test Reactor (RERTR) program in JAERI, the enrichment reduction for Japan Research Reactor 4 (JRR-4) is in progress. A fuel element using a 19.75 % enriched UAlx-Al dispersion type with a uranium density of 2.2 g/cm 3 was designed as the LEU fuel and the neutronic and thermo-hydraulic performances of the LEU core were compared with those of the current HEU core. The results of the neutronic design are as follows: (1) the excess reactivity of the LEU core becomes about 1 % Δk/k less, (2) the thermal neutron flux in the fuel region decreases about 25 % on the average, (3) the thermal neutron fluxes in the irradiation pipes are almost the same and (4) the core burnup lifetime becomes about 20 % longer. The thermo-hydraulic design also shows that: (1) the fuel plate surface temperature decreases about 10 deg C due to the increase of the number of fuel plates and (2) the temperature margin with respect to the ONB temperature increases. Therefore, it is confirmed that the same utilization performance as the HEU core is attainable with the LEU core. (author)

Characterization of Uranium-Bearing Material by Passive Non-Destructive Gamma Spectrometry

International Nuclear Information System (INIS)

Lakosi, L.; Zsigrai, J.; Nguyen, C.T.

2009-01-01

Characterization of nuclear materials is equally important in nuclear safeguards (inventory verification) and in nuclear security (revealing illicit trafficking). Analysis of materials is a key issue in both fields. Natural (NU), depleted (DU), low-enriched (LEU), and high-enriched uranium (HEU) samples were analysed by high resolution gamma spectrometry (HRGS). Isotopic composition and total U-content of reactor fuel pellets and powder were determined. A unique HRGS method was developed for the first time for determining the production date of the material of unknown origin. Identifying reprocessed uranium proved to be possible by HRGS as well.

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)

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.

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

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

Energy Technology Data Exchange (ETDEWEB)

Neef, Hans Joachim

1974-03-15

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

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

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)

Effects of enriched uranium on developing brain damage of neonatal rats

International Nuclear Information System (INIS)

Gu Guixiong; Zhu Shoupeng; Wang Liuyi; Yang Shuqin; Zhu Lingli

2001-01-01

The model of irradiation-induced brain damage in vivo was settled first of all. The micro-auto-radiographic tracing showed that when the rat's brain at postnatal day after lateral ventricle injection with enriched uranium 235 U the radionuclides were mainly accumulated in the nucleus. At the same time autoradiographic tracks appeared in the cytoplasm and interval between cells. The effects of cerebrum exposure to alpha irradiation by enriched uranium on somatic growth and neuro-behavior development of neonatal rats were examined by determination of multiple parameters. In the growth and development of the neonatal rat's cerebrum exposure to enriched uranium, the somatic growth such as body weight and brain weight increase was lower significantly. The data indicated that the neonatal wistar rats having cerebrum exposure to alpha irradiation by enriched uranium showed delayed growth and abnormal neuro-behavior. The changes of neuron specific enolase (NSE), interleukin-1 β (IL- β), superoxide dismutase (SOD), and endothelin (ET) in cerebellum, cerebral cortex, hippocampus, diencephalons of the rat brain after expose to alpha irradiation by enriched uranium were examined with radioimmunoassay. The results showed that SOD and ET can be elevated by the low dose irradiation of enriched uranium, and can be distinctly inhibited by the high dose. The data in view of biochemistry indicated firstly that alpha irradiation from enriched uranium on the developing brain damage of neonatal rats were of sensibility, fragility and compensation in nervous cells

Effects of enriched uranium on developing brain damage of neonatal rats

Energy Technology Data Exchange (ETDEWEB)

Guixiong, Gu; Shoupeng, Zhu; Liuyi, Wang; Shuqin, Yang; Lingli, Zhu [Suzhou Medical College, Suzhou (China)

2001-04-01

The model of irradiation-induced brain damage in vivo was settled first of all. The micro-auto-radiographic tracing showed that when the rat's brain at postnatal day after lateral ventricle injection with enriched uranium {sup 235}U the radionuclides were mainly accumulated in the nucleus. At the same time autoradiographic tracks appeared in the cytoplasm and interval between cells. The effects of cerebrum exposure to alpha irradiation by enriched uranium on somatic growth and neuro-behavior development of neonatal rats were examined by determination of multiple parameters. In the growth and development of the neonatal rat's cerebrum exposure to enriched uranium, the somatic growth such as body weight and brain weight increase was lower significantly. The data indicated that the neonatal wistar rats having cerebrum exposure to alpha irradiation by enriched uranium showed delayed growth and abnormal neuro-behavior. The changes of neuron specific enolase (NSE), interleukin-1 {beta} (IL- {beta}), superoxide dismutase (SOD), and endothelin (ET) in cerebellum, cerebral cortex, hippocampus, diencephalons of the rat brain after expose to alpha irradiation by enriched uranium were examined with radioimmunoassay. The results showed that SOD and ET can be elevated by the low dose irradiation of enriched uranium, and can be distinctly inhibited by the high dose. The data in view of biochemistry indicated firstly that alpha irradiation from enriched uranium on the developing brain damage of neonatal rats were of sensibility, fragility and compensation in nervous cells.

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)

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

Gas centrifuge enrichment plants inspection frequency and remote monitoring issues for advanced safeguards implementation

International Nuclear Information System (INIS)

Boyer, Brian David; Erpenbeck, Heather H.; Miller, Karen A.; Ianakiev, Kiril D.; Reimold, Benjamin A.; Ward, Steven L.; Howell, John

2010-01-01

Current safeguards approaches used by the IAEA at gas centrifuge enrichment plants (GCEPs) need enhancement in order to verify declared low enriched uranium (LEU) production, detect undeclared LEU production and detect high enriched uranium (BEU) production with adequate probability using non destructive assay (NDA) techniques. At present inspectors use attended systems, systems needing the presence of an inspector for operation, during inspections to verify the mass and 235 U enrichment of declared cylinders of uranium hexafluoride that are used in the process of enrichment at GCEPs. This paper contains an analysis of how possible improvements in unattended and attended NDA systems including process monitoring and possible on-site destructive analysis (DA) of samples could reduce the uncertainty of the inspector's measurements providing more effective and efficient IAEA GCEPs safeguards. We have also studied a few advanced safeguards systems that could be assembled for unattended operation and the level of performance needed from these systems to provide more effective safeguards. The analysis also considers how short notice random inspections, unannounced inspections (UIs), and the concept of information-driven inspections can affect probability of detection of the diversion of nuclear material when coupled to new GCEPs safeguards regimes augmented with unattended systems. We also explore the effects of system failures and operator tampering on meeting safeguards goals for quantity and timeliness and the measures needed to recover from such failures and anomalies.

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...

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

Development of uranium metal targets for 99Mo production

International Nuclear Information System (INIS)

Wiencek, T.C.; Hofman, G.L.

1993-10-01

A substantial amount of high enriched uranium (HEU) is used for the production of medical-grade 99 Mo. Promising methods of producing irradiation targets are being developed and may lead to the reduction or elimination of this HEU use. To substitute low enriched uranium (LEU) for HEU in the production of 99 Mo, the target material may be changed to uranium metal foil. Methods of fabrication are being developed to simplify assembly and disassembly of the targets. Removal of the uranium foil after irradiation without dissolution of the cladding is a primary goal in order to reduce the amount of liquid radioactive waste material produced in the process. Proof-of-concept targets have been fabricated. Destructive testing indicates that acceptable contact between the uranium foil and the cladding can be achieved. Thermal annealing tests, which simulate the cladding/uranium diffusion conditions during irradiation, are underway. Plans are being made to irradiate test targets

Detection of illicit HEU production in gaseous centrifuge enrichment plants using neutron counting techniques on product cylinders

International Nuclear Information System (INIS)

Freeman, Corey R.; Geist, William H.

2010-01-01

Innovative and novel safeguards approaches are needed for nuclear energy to meet global energy needs without the threat of nuclear weapons proliferation. Part of these efforts will include creating verification techniques that can monitor uranium enrichment facilities for illicit production of highly-enriched uranium (HEU). Passive nondestructive assay (NDA) techniques will be critical in preventing illicit HEU production because NDA offers the possibility of continuous and unattended monitoring capabilities with limited impact on facility operations. Gaseous centrifuge enrichment plants (GCEP) are commonly used to produce low-enriched uranium (LEU) for reactor fuel. In a GCEP, gaseous UF 6 spins at high velocities in centrifuges to separate the molecules containing 238 U from those containing the lighter 235 U. Unfortunately, the process for creating LEU is inherently the same as HEU, creating a proliferation concern. Insuring that GCEPs are producing declared enrichments poses many difficult challenges. In a GCEP, large cascade halls operating thousands of centrifuges work together to enrich the uranium which makes effective monitoring of the cascade hall economically prohibitive and invasive to plant operations. However, the enriched uranium exiting the cascade hall fills product cylinders where the UF 6 gas sublimes and condenses for easier storage and transportation. These product cylinders hold large quantities of enriched uranium, offering a strong signal for NDA measurement. Neutrons have a large penetrability through materials making their use advantageous compared to gamma techniques where the signal is easily attenuated. One proposed technique for detecting HEU production in a GCEP is using neutron coincidence counting at the product cylinder take off stations. This paper discusses findings from Monte Carlo N-Particle eXtended (MCNPX) code simulations that examine the feasibility of such a detector.

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

International Nuclear Information System (INIS)

2015-01-01

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

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.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

Long-term outlook for global natural uranium and uranium enrichment supply and demand situations after the impact of Fukushima Daiichi Nuclear Power Plant accident

International Nuclear Information System (INIS)

Matsuo, Yuhji; Murakami, Tomoko

2012-01-01

In this paper, the authors propose long-term projections of global nuclear power generation, uranium production, and uranium enrichment capacities by region, and estimate the trade flows of natural uranium and uranium enrichment activities in 2020 and 2035. In spite of the rapid nuclear power generation capacity growth expected especially in Asia, the natural uranium and uranium enrichment trade will not be tightened by 2020 due to the projected increase in both natural uranium production and uranium enrichment capacities, which may cause a drop in natural uranium and uranium enrichment prices. Thus, there is a great possibility that the current projects for capacity expansion will be delayed considerably. However, in the 'high-demand scenario', where nuclear expansion will be accelerated due to growing concerns about global warming and energy security issues, additional investments in uranium production and enrichment facilities will be needed by 2035. In Asia, the self-sufficiency ratio for both natural uranium supply and uranium enrichment activities will remain relatively low until 2035. However, the Herfindahl-Hirschman (HH) index of natural uranium and uranium enrichment activity trade to Asia will be lowered considerably up to 2035, indicating that nuclear capacity expansion can contribute to enhancing energy security in Asia. (author)

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.)

Enriched-uranium feed costs for the High-Temperature Gas-Cooled reactor: trends and comparison with other reactor concepts

International Nuclear Information System (INIS)

Thomas, W.E.

1976-04-01

This report discusses each of the components that affect the unit cost for enriched uranium; that is, ore costs, U 3 O 8 to UF 6 conversion cost, costs for enriching services, and changes in transaction tails assay. Historical trends and announced changes are included. Unit costs for highly enriched uranium (93.15 percent 235 U) and for low-enrichment uranium (3.0, 3.2, and 3.5 percent 235 U) are displayed as a function of changes in the above components and compared. It is demonstrated that the trends in these cost components will probably result in significantly less cost increase for highly enriched uranium than for low-enrichment uranium--hence favoring the High-Temperature Gas-Cooled Reactor

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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

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

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

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)

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.

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

Valence-associated uranium isotope fractionation of uranium enriched phosphate in a shallow aquifer, Lee County, Florida

International Nuclear Information System (INIS)

Weinberg, J.M.; Levine, B.R.; Cowart, J.B.

1993-01-01

The source of anomalously high concentrations of uranium, characterized by U-234/U-238 activity ratios significantly less than unity, in shallow groundwaters of Lee County, Florida, was investigated. Uranium in cores samples was separated into U(IV) and U(VI) oxidation state fractions, and uranium analyses were conducted by alpha spectrometry. Uranium mobility was also studied in selected leaching experiments. Results indicate that mobilization of unusually soluble uranium, present in uranium enriched phosphate of the Pliocene age Tamiami Formation at determined concentrations of up to 729 ppm, is the source for high uranium concentrations in groundwater. In leaching experiments, approximately one-third of the uranium present in the uranium enriched phosphate was mobilized into the aqueous phase. Results of previous investigations suggest that U-234, produced in rock by U-238 decay, is selectively oxidized to U(VI). The uranium enriched phosphate studied in this investigation is characterized by selective reduction of U-234, with a pattern of increasing isotopic fractionation with core depth. As a consequence, U-234/U-238 activity ratios greater than 1.0 in the U(IV) fraction, and less than 1.0 in the U(VI) fraction have developed in the rock phase. In leaching experiments, the U(VI) fraction from the rock was preferentially mobilized into the aqueous phase, suggesting that U-234/U-238 activity ratios of leaching groundwaters are strongly influenced by the isotopic characteristics of the U(VI) fraction of rock. It is suggested that preferential leaching of U(VI), present in selectivity reduced uranium enriched phosphate, is the source for low activity ratio groundwaters in Lee County

A disposition strategy for highly enriched, aluminum-based fuel from research and test reactors

International Nuclear Information System (INIS)

McKibben, J.M.; Gould, T.H.; McDonell, W.R.; Bickford, W.E.

1994-01-01

The strategy proposed in this paper offers the Department of Energy an approach for disposing of aluminum-based, highly enriched uranium (HEU) spent fuels from foreign and domestic research reactors. The proposal is technically, socially, and economically sound. If implemented, it would advance US non-proliferation goals while also disposing of the spent fuel's waste by timely and proven methods using existing technologies and facilities at SRS without prolonged and controversial storage of the spent fuel. The fuel would be processed through 221-H. The radioactive fission products (waste) would be treated along with existing SRS high level waste by vitrifying it as borosilicate glass in the Defense Waste Processing Facility (DWPF) for disposal in the national geological repository. The HEU would be isotopically diluted, during processing, to low-enriched uranium (LEU) which can not be used to make weapons, thus eliminating proliferation concerns. The LEU can be sold to fabricators of either research reactor fuel or commercial power fuel. This proposed processing-LEU recycle approach has several important advantages over other alternatives, including: Lowest capital investment; lowest net total cost; quickest route to acceptable waste form and final geologic disposal; and likely lowest safety, health, and environmental impacts

A neutronic feasibility study for LEU conversion of the IR-8 research reactor

International Nuclear Information System (INIS)

Deen, J.R.; Hanan, N.A.; Matos, J.E.; Egorenkov, P.M.; Nasonov, V.A.

1998-01-01

Equilibrium fuel cycle comparisons for the IR-8 research reactor were made for HEU (90%), HEU (36%), and LEU (19.75%) fuel assembly (FA) designs using three dimensional multi-group diffusion theory models benchmarked to detailed Monte Carlo models of the reactor. Comparisons were made of changes in reactivity, cycle length, average 235 U discharge burnup, thermal neutron flux, and control rod worths for the 90% and 36% enriched IRT-3M fuel assembly and the 19.75% enriched IRT-4M fuel assembly with the same fuel management strategy. The results of these comparisons showed that a uranium density of 3.5 g/cm 3 in the fuel meat would be required in the LEU IRT-4M fuel assembly to match the cycle length of the HEU (90%) IRT-3M FA and an LEU density of 3.7 g/cm 3 is needed to match the cycle length of the HEU (36%) IRT-3M FA. (author)

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

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

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

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

LEU-HTR critical experiment program for the PROTEUS facility in Switzerland

International Nuclear Information System (INIS)

Brogli, R.; Bucher, K.H.; Chawla, R.; Foskolos, K.; Luchsinger, H.; Mathews, D.; Sarlos, G.; Seiler, R.

1990-01-01

New critical experiments in the framework of an IAEA Coordinated Research Program on 'Validation of Safety Related Reactor Physics Calculations for Low Enriched HTRs' are planned at the PSI PROTEUS facility. The experiments are designed to supplement the experimental data base and reduce the design and licensing uncertainties for small- and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel. The main objectives of the new experiments are to provide first-of-a-kind high quality experimental data on: 1) The criticality of simple, easy to interpret, single core region LEU HTR systems for several moderator-to-fuel ratios and several lattice geometries; 2) the changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor, and 3) the effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. Work on the design and licensing of the modified PROTEUS critical facility is now in progress with the HTR experiments scheduled to begin early in 1991. Several international partners will be involved in the planning, execution and analysis of these experiments in order to insure that they are relevant and cost effective with respect to the various gas cooled reactor national programs. (author)

LEU-HTR critical experiment program for the PROTEUS facility in Switzerland

Energy Technology Data Exchange (ETDEWEB)

Brogli, R; Bucher, K H; Chawla, R; Foskolos, K; Luchsinger, H; Mathews, D; Sarlos, G; Seiler, R [Paul Scherrer Institute, Laboratory for Reactor Physics and System Technology Wuerenlingen and Villigen, Villigen PSI (Switzerland)

1990-07-01

New critical experiments in the framework of an IAEA Coordinated Research Program on 'Validation of Safety Related Reactor Physics Calculations for Low Enriched HTRs' are planned at the PSI PROTEUS facility. The experiments are designed to supplement the experimental data base and reduce the design and licensing uncertainties for small- and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel. The main objectives of the new experiments are to provide first-of-a-kind high quality experimental data on: 1) The criticality of simple, easy to interpret, single core region LEU HTR systems for several moderator-to-fuel ratios and several lattice geometries; 2) the changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor, and 3) the effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. Work on the design and licensing of the modified PROTEUS critical facility is now in progress with the HTR experiments scheduled to begin early in 1991. Several international partners will be involved in the planning, execution and analysis of these experiments in order to insure that they are relevant and cost effective with respect to the various gas cooled reactor national programs. (author)

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.)

LEU fuel fabrication in Argentina

International Nuclear Information System (INIS)

Giorsetti, D.R.; Gomez, J.O.; Marajofsky, A.; Kohut, C.

1985-01-01

As an Institution, aiming to meet with its own needs, CNEA has been intensively developing reduced enriched fuel to use in its own research and test reactors. Development of the fabrication technology as well as the design, installation and operation of the manufacturing plant, have been carried out with its own funds. Irradiation and post-irradiation of test miniplates have been taking place within the framework of the RERTR program. During the last years, CNEA has developed three LEU fuel types. In the previous RERTR meetings, we presented the technological results obtained with these fuel types. This paper focuses on CNEA LEU fuel element manufacturing status and the trained personnel we can offer in design and manufacture fuel capability. CNEA has its own fuel manufacturing technology; the necessary facilities to start the fuel fabrication; qualified technicians and professionals for: fuel design and behaviour analysis; fuel manufacturing and QA; international recognition of its fuel development and manufacturing capability through its ORR miniplate irradiation; its own natural uranium and the future possibility to enrich up to 20% U 235 ; the probability to offer a competitive fuel manufacturing cost in the international market; the disposition to cooperate with all countries that wish to take part and aim to reach an self-sufficiency in their own fuel supply needs

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

Critical review of uranium resources and production capability to 2020

International Nuclear Information System (INIS)

1998-08-01

This report was prepared to assess the changing uranium supply and demand situation as well as the adequacy of uranium resources and the production capability to supply uranium concentrate to meet reactor demand through 2020. Uranium production has been meeting only 50 to 60 percent of the world requirements with the balance met from sale of excess inventory offered on the market at low prices. It is generally agreed by most specialists that the end of the excess inventory is approaching. With inventory no longer able to meet the production shortfall it is necessary to significantly expand uranium production to fill an increasing share of demand. Non-production supplies of uranium, such as the blending of highly enriched uranium (HEU) warheads to produce low enriched reactor fuel and reprocessing of spent fuel, are also expected to grow in importance as a fuel source. This analysis addresses three major concerns as follows: adequacy of resources to meet projected demand; adequacy of production capability to produce the uranium; and market prices to sustain production to fill demand. This analysis indicates uranium mine production to be the primary supply providing about 76 to 78 percent of cumulative needs through 2020. Alternative sources supplying the balance, in order of relative importance are: (1) low enriched uranium (LEU) blended from 500 tonnes of highly enriched uranium (HEU) Russian weapons, plus initial US Department of Energy (US DOE) stockpile sales (11 to 13%); (2) reprocessing of spent nuclear fuel (6%) and; (3) utility and Russian stockpiles. Further this report gives uranium production profiles by countries: CIS producers (Kazakhstan, Russian Federation, Ukraine, Uzbekistan) and other producers (Australia, Canada, China, Gabon, Mongolia, Namibia, Niger, South Africa, United States of America)

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

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

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

International Nuclear Information System (INIS)

Muhammad, Farhan

2011-01-01

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

Operating experience, measurements, and analysis of the LEU whole core demonstration at the FNR

International Nuclear Information System (INIS)

Weha, D.K.; Drumm, C.R.; King, J.S.; Martin, W.R.; Lee, J.C.

1984-01-01

The 2-MW Ford Nuclear Reactor at the University of Michigan is serving as the demonstration reactor for the MTR-type low enrichment (LEU) fuel for the Reduced Enrichment for Research and Test Reactor program. Operational experience gained through six months of LEU core operation and seven months of mixed HEU-LEU core operation is presented. Subcadmium flux measurements performed with rhodium self-powered neutron detectors and iron wire activations are compared with calculations. Measured reactivity parameters are compared for HEU and LEU cores. Finally, the benchmark calculations for several HEU, LEU, and mixed HEU-LEU FNR cores and the International Atomic Energy Agency (IAEA) benchmark problem are presented. (author)

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)

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...

Status of the reduced enrichment for research reactors program in Argentina

International Nuclear Information System (INIS)

Perez, E.; Kohut, C.

2004-01-01

In the area of Research and Test Reactors' fuel elements, the different stages of development carried out by the Atomic Energy Commission of Argentina (CNEA) until now, and the future plans are presented in this paper. Own and foreign programs, for reducing the risk of proliferation due the use of high enriched uranium fuel elements in these types of reactors, is mentioned. A brief description of different work performed is presented: At first the experience with the use of highly enriched uranium, and then the activities related with the development done in order to achieve a good knowledge in low-enriched (LEU) fuels, particularly in the area of U308-Al fuels. This experience has permitted us, supported by the excellent results obtained, to be in a position to satisfy our own requirements and also to supply to other countries, not only fuels but also technology transferences and facilities of the development appropriate for this purpose. The main modifications brought in the design and fabrication of these types of fuel elements is also described. Finally, and with the main objective to complete the development and to qualify the LEU fuels based on silicides and to improve the actual MO-99 blanket fabrication technology two new C.N.E.A. projects, are outlined.(author)

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

International Nuclear Information System (INIS)

Opperman, E; Kenneth Yates, K

2007-01-01

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

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)

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)

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

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

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)

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)

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)

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)

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

International safeguards at the feed and withdrawal area of a gas centrifuge uranium enrichment plant

International Nuclear Information System (INIS)

Gordon, D.M.; Sanborn, J.B.

1980-01-01

This paper discusses the application of International Atomic Energy Agency (IAEA) safeguards at a model gas centrifuge uranium enrichment plant designed for the production of low-enriched uranium; particular emphasis is placed upon the verification by the IAEA of the facility material balance accounting. 13 refs

Neutronic calculations for the conversion of the University of Florida Training Reactor from HEU to LEU fuel

International Nuclear Information System (INIS)

Dugan, E.T.; Diaz, N.J.; Kniedler, G.S.

1983-01-01

The University of Florida Training Reactor (UFTR) is located on the University of Florida campus in Gainesville, Florida. The reactor is the Argonaut type, heterogeneous in design and currently fueled with 93% enriched, uranium-aluminum alloy MTR plate-type fuel. Investigations are being performed to examine te feasibility of replacing the highly-enriched fuel of the current UFTR with 4.8% enriched, cylindrical pin SPERT fuel. The SPERT fuel is stainless steel clad and contains uranium dioxide (UO 2 ) pellets. On a broad spectrum, training reactor conversion from high enrichment uranium (HEU) to low enrichment uranium (LEU) fueled facilities has been a continuing concern in the International Atomic Energy Agency (IAEA) and significant work has been done in this area by the Argonne RERTR Program. The International Atomic Energy Agency cites three reasons for reactor conversion to low-enriched uranium. The main reason is the desire to reduce the proliferation potential of research reactor fuels. The second is to increase the assurance of continued fuel availability in the face of probable restrictions on the supply of highly-enriched uranium. The third reason is the possible reduction in requirements for physical security measures during fabrication, transportation, storage and use. This same IAEA report points out that the three reasons stated for the conversion of the fuel of research reactors are interrelated and cannot be considered individually. The concerns of the Nuclear Engineering Sciences Department at the University of Florida relating to the HEU fuel of the UFTR coincide with those of the International Atomic Energy Agency. The primary reason for going to low-enriched pin-type fuel is the concern with proliferation provoked by the highly-enriched plate fuel which has led to tighter security of nuclear facilities such as the UFTR. A second reason for changing to the pin-type fuel is because of difficulties that are being encountered in the supply of the

Neutronic calculations for the conversion of the University of Florida Training Reactor from HEU to LEU fuel

Energy Technology Data Exchange (ETDEWEB)

Dugan, E T; Diaz, N J [Department of Nuclear Engineering Sciences, University of Florida, Gainesville, FL (United States); Kniedler, G S [Reactor Analysis Group, TVA, Chattanooga, TN (United States)

1983-09-01

The University of Florida Training Reactor (UFTR) is located on the University of Florida campus in Gainesville, Florida. The reactor is the Argonaut type, heterogeneous in design and currently fueled with 93% enriched, uranium-aluminum alloy MTR plate-type fuel. Investigations are being performed to examine te feasibility of replacing the highly-enriched fuel of the current UFTR with 4.8% enriched, cylindrical pin SPERT fuel. The SPERT fuel is stainless steel clad and contains uranium dioxide (UO{sub 2}) pellets. On a broad spectrum, training reactor conversion from high enrichment uranium (HEU) to low enrichment uranium (LEU) fueled facilities has been a continuing concern in the International Atomic Energy Agency (IAEA) and significant work has been done in this area by the Argonne RERTR Program. The International Atomic Energy Agency cites three reasons for reactor conversion to low-enriched uranium. The main reason is the desire to reduce the proliferation potential of research reactor fuels. The second is to increase the assurance of continued fuel availability in the face of probable restrictions on the supply of highly-enriched uranium. The third reason is the possible reduction in requirements for physical security measures during fabrication, transportation, storage and use. This same IAEA report points out that the three reasons stated for the conversion of the fuel of research reactors are interrelated and cannot be considered individually. The concerns of the Nuclear Engineering Sciences Department at the University of Florida relating to the HEU fuel of the UFTR coincide with those of the International Atomic Energy Agency. The primary reason for going to low-enriched pin-type fuel is the concern with proliferation provoked by the highly-enriched plate fuel which has led to tighter security of nuclear facilities such as the UFTR. A second reason for changing to the pin-type fuel is because of difficulties that are being encountered in the supply of

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.)

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)

LEU-fueled SLOWPOKE-2 modelling with MCNP4A

International Nuclear Information System (INIS)

Pierre, J.R.M.; Bonin, H.W.J.

1996-01-01

Following the commissioning of the Low Enrichment Uranium (LEU) Fueled SLOWPOKE-2 research reactor at Royal Military College,excess reactivity measurements were conducted over a range of temperature and power. Given the advance in computer technology, the use of Monte Carlo N-Particle Transport Code System MCNP 4A appeared possible for the simulation of the LEU-fueled SLOWPOKE-2 reactor core, and this work demonstrates that this is indeed the case. MCNP 4A is a full three dimensional program allowing the user to enter a large amount of complexity. The limit on the geometry complexity is the computing time required to achieve a reasonable standard deviation. To this point several models of the SLOWPOKE-2 have been developed giving some insight on the sensitivity of the code. MCNP4A can use various cross section libraries. The aim of this work is to calculate accurately the reactivity of the core and reproduce The temperature trend of the reactivity. The model preserved as much as possible the details of the core and facility in order to allow further study in the flux mapping

A neutronic feasibility study for LEU conversion of the WWR-M reactor at Gatchina

International Nuclear Information System (INIS)

Petrov, Yu. V.; Erykalov, A.N.; Onegin, M.S.

2000-01-01

In this report we present the results of computations of the full scale reactor core with HEU (90%), MEU (36%) and LEU (19.75%) fuel. The reactor computer model for the MCU RFFI Monte Carlo code includes all peculiarities of the core. Calculations show that a uranium density of 3.3gU/cm 3 of MEU (36%) fuel and 8/25gU/cm 3 of LEU (19.75%) in WWR-M5 fuel assembly (FA) geometry is required to match the fuel cycle length of the HEU (90%) case with the same end of cycle (EOEC) excess reactivity. For the equilibrium fuel cycle the fuel burnup and poisoning, the fast and thermal neutron fluxes, the reactivity worth of control rods were calculated for the reference case with HEU (90%) FA and for the MEU and LEU FA. The relative accuracy of this neutronic feasibility study of fuel enrichment reduction of the WWR-M reactor in Gatchina is sufficient to start the fabrication feasibility study of MEU (36%) WWR-M5 fuel assemblies. At the present stage of technology it seems hardly possible to manufacture LEU (19.75%) fuel elements in WWR-M5 geometry due to too high uranium density. Only a future R and D can solve the problem. (author)

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

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)

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

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

International Nuclear Information System (INIS)

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

1991-07-01

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

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)

Acidic aqueous uranium electrodeposition for target fabrication

International Nuclear Information System (INIS)

Saliba-Silva, A.M.; Oliveira, E.T.; Garcia, R.H.L.; Durazzo, M.

2013-01-01

Direct irradiation of targets inside nuclear research or multiple purpose reactors is a common route to produce 99 Mo- 99m Tc radioisotopes. The electroplating of low enriched uranium over nickel substrate might be a potential alternative to produce targets of 235 U. The electrochemistry of uranium at low temperature might be beneficial for an alternative route to produce 99 Mo irradiation LEU targets. Electrodeposition of uranium can be made using ionic and aqueous solutions producing uranium oxide deposits. The performance of uranium electrodeposition is relatively low because a big competition with H 2 evolution happens inside the window of electrochemical reduction potential. This work explores possibilities of electroplating uranium as UO 2 2+ (Uranium-VI) in order to achieve electroplating uranium in a sufficient amount to be commercially irradiated in the future Brazilian RMB reactor. Electroplated nickel substrate was followed by cathodic current electrodeposition from aqueous UO 2 (NO 3 ) 2 solution. EIS tests and modeling showed that a film formed differently in the three tested cathodic potentials. At the lower level, (-1.8V) there was an indication of a double film formation, one overlaying the other with ionic mass diffusion impaired at the interface with nickel substrate as showed by the relatively lower admittance of Warburg component. (author)

Verification of maximum radial power peaking factor due to insertion of FPM-LEU target in the core of RSG-GAS reactor

Energy Technology Data Exchange (ETDEWEB)

Setyawan, Daddy, E-mail: d.setyawan@bapeten.go.id [Center for Assessment of Regulatory System and Technology for Nuclear Installations and Materials, Indonesian Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada No. 8 Jakarta 10120 (Indonesia); Rohman, Budi [Licensing Directorate for Nuclear Installations and Materials, Indonesian Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada No. 8 Jakarta 10120 (Indonesia)

2014-09-30

Verification of Maximum Radial Power Peaking Factor due to insertion of FPM-LEU target in the core of RSG-GAS Reactor. Radial Power Peaking Factor in RSG-GAS Reactor is a very important parameter for the safety of RSG-GAS reactor during operation. Data of radial power peaking factor due to the insertion of Fission Product Molybdenum with Low Enriched Uranium (FPM-LEU) was reported by PRSG to BAPETEN through the Safety Analysis Report RSG-GAS for FPM-LEU target irradiation. In order to support the evaluation of the Safety Analysis Report incorporated in the submission, the assessment unit of BAPETEN is carrying out independent assessment in order to verify safety related parameters in the SAR including neutronic aspect. The work includes verification to the maximum radial power peaking factor change due to the insertion of FPM-LEU target in RSG-GAS Reactor by computational method using MCNP5and ORIGEN2. From the results of calculations, the new maximum value of the radial power peaking factor due to the insertion of FPM-LEU target is 1.27. The results of calculations in this study showed a smaller value than 1.4 the limit allowed in the SAR.

The Proliferation Security Initiative: A Means to an End for the Operational Commander

Science.gov (United States)

2009-05-04

The Reduced Enrichment for Research and Test Reactors ( RERTR ) Program develops technology necessary to enable the conversion of civilian...facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets. The RERTR Program was initiated by the U.S. Department of...processes have been developed for producing radioisotopes with LEU targets. The RERTR Program is managed by the Office of Nuclear Material Threat

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)

Development and industrial application of gas centrifuges to uranium enrichment in the USSR

International Nuclear Information System (INIS)

Abbakumov, E.I.; Bazhenov, V.A.; Verbin, Yu.V.

1989-01-01

Review of state and studies in the field of gaseous diffusion technology and centrifugal method of uranium enrichment in the USSR is given. Domestic industrial gas centrifuges, forming to-day the main part of separation capacities in the USSR, are noted for low specific energy consumption and high reliability. Centrifugal technology in the USSR is applied both to uranium enrichment (including one for export) and to separation of isotopes of other chemical elements

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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

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.

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

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%.

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

Measurement of the enrichment of uranium-hexafluoride gas in product pipes in the centrifuge enrichment plant at Almelo

International Nuclear Information System (INIS)

Packer, T.W.; Lees, E.W.; Aaldijk, J.K.; Harry, R.J.S.

1987-09-01

One of the objectives of safeguarding centrifuge enrichment plants is to apply non-destructive measurements inside the cascade area to confirm that the enrichment level is in the low enriched uranium range. Research in the UK and USA has developed a NDA instrument which can confirm the presence of low enriched uranium on a rapid go/no go basis in cascade header pipework of their centrifuge enrichment plants. The instrument is based on a gamma spectroscopic measurement coupled with an X-ray fluorescence analysis. This report gives the results of measurements carried out at Almelo by the UKAEA Harwell, ECN Petten and KFA Juelich to determine if these techniques could be employed at Almelo and Gronau. The energy dispersive X-ray fluorescence analysis has been applied to determine the total mass of uranium in the gas phase, and the deposit correction technique and the two geometry technique have been applied at Almelo to correct the measured gamma intensities for those emitted by the deposit. After an executive summary the report discusses the principles of the two correction methods. A short description of the equipment precedes the presentation of the results of the measurements and the discussion. After the conclusions the report contains two appendices which contain the derivation of the formulae for the deposit correction technique and a discussion of the systematic errors of this technique. 8 figs.; 11 refs.; 6 tables

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)

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...

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)

Analysis on possible North Korea's uranium stock by using open source information

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Hyun [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

2012-10-15

The gas centrifuge plant in Yongbyon is the only revealed uranium enrichment plant in North Korea. Yongbyon enrichment plant is believed producing only LEU below 5 percent uranium 235 for use as fuel in the LWR, as they said. But the plant could be easily converted to producing HEU for nuclear weapons. In this paper, we estimated the enrichment capability of the Yongbyon plant based on the known characteristics of its centrifuges and cascades. And then we developed the four possible uranium enrichment scenarios, to examine the future options that North Korea may use to enhance its enrichment capability. Finally, we suggested several key measures to stop North Korea from pursing its nuclear ambitions.

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

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)

Low enrichment fuel conversion for Iowa State University. Final report

International Nuclear Information System (INIS)

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

1996-01-01

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

An Effort to Improve Uranium Foil Target Fabrication Technology by Single Roll Method

Energy Technology Data Exchange (ETDEWEB)

Sim, Moon Soo; Lee, Jong Hyeon [Chungnam National University, Daejeon (Korea, Republic of); Kim, Chang Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-05-15

Technetium-99({sup 99m}Tc) is the most commonly used radioisotope in nuclear medicine for diagnostic procedures. It is produced from the decay of its parent Mo-99, which is sent to the hospital or clinic in the form of a generator. Recently, all of the major providers of Mo-99 have used high-enrichment uranium (HEU) as a target material in a research and test reactor. As a part of a nonproliferation effort, the RERTR program has investigated the production of the fission isotope Mo-99 using low-enrichment uranium(LEU) instead of HEU since 1993, a parent nuclide of {sup 99m}Tc , which is a major isotope for a medical diagnosis. As uranium foils have been produced by the conventional method on a laboratory scale by a repetitive hot-rolling method with significant problems in foil quality, productivity and economic efficiency, attention has shifted to the planar flow casting(PFC) method. In KAERI, many experiments are performed using depleted uranium(DU).

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)

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

Status of fuel element technology for plate type dispersion fuels with high uranium density

International Nuclear Information System (INIS)

Hrovat, M.; Huschka, H.; Koch, K.H.; Nazare, S.; Ondracek, G.

1983-01-01

A number of about 20 Material Test and Research Reactors in Germany and abroad is supplied with fuel elements by the company NUKEM. The power of these reactors differs widely ranging from up to about 100 MW. Consequently, the uranium density of the fuel elements in the meat varies considerably depending on the reactor type and is usually within the range from 0.4 to 1.3 g U/cm 3 if HEU is used. In order to convert these reactors to lower uranium enrichment (19.75% 235-U) extensive work is carried out at NUKEM since about two years with the goal to develop fuel elements with high U-density. This work is sponsored by the German Ministry for Research and Technology in the frame of the AF-program. This paper reports on the present state of development for fuel elements with high U-density fuels at NUKEM is reported. The development works were so far concentrated on UAl x , U 3 O 8 and UO 2 fuels which will be described in more detail. In addition fuel plates with new fuels like e.g. U-Si or U-Fe compounds are developed in collaboration with KfK. The required uranium densities for some typical reactors with low, medium, and high power are listed allowing a comparison of HEU and LEU uranium density requirements. The 235-U-content in the case of LEU is raised by 18%. Two different meat thicknesses are considered: Standard thickness of 0.5 mm; and increased thickness of 0.76 mm. From this data compilation the objective follows: in the case of conversion to LEU (19.75% 235-U-enrichment), uranium densities have to be made available up to 24 gU/cm 3 meat for low power level reactors, up to 33 gU/cm 3 meat for medium power level reactors, and between 5.75 and 7.03 g/cm 3 meat for high power level reactors according to this consideration

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

Developing Techniques for Small Scale Indigenous Molybdenum-99 Production Using LEU Fission at Tajoura Research Center-Libya [Country report: Libya