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Sample records for irradiated leu u-mo

  1. 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.

  2. Results of Microstructural Examinations of Irradiated LEU U-Mo Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, ID 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organization (Australia)

    2009-06-15

    Introduction: The RERTR program is responsible for converting research reactors that use high-enriched uranium fuels to ones that use low-enriched uranium fuels [1]. As part of the development of LEU fuels, a variety of irradiation experiments are being conducted using the Advanced Test Reactor. Based on the results of initial fuel plate testing, adjustments have been made to the characteristics of fuel plates to improve the stability of the fuel microstructure. One improvement has been to add Si to the matrix of a dispersion fuel. This material is also being added at the fuel/cladding interface of a monolithic fuel. This paper will discuss the irradiation performance of these fuels, in terms of the stability of their microstructures during irradiation. Results and discussion: The post-irradiation examinations of fuel plates are performed at the Idaho National Laboratory. These examinations consist of visual examinations of fuel plates, gamma scanning, thickness measurements, oxide thickness measurements, and optical metallographic examinations of the fuel plate microstructures. Microstructural analysis is also performed using scanning electron microscopy. Overall, U-7Mo and U-10Mo alloy fuels have displayed the best irradiation performance, particularly, when a Si-containing Al alloy is used as the dispersion fuel matrix. The benefit of using this type of matrix is that the commonly observed fuel/cladding interaction that occurs during irradiation is reduced and the interaction layer that forms exhibit stable behavior during irradiation. Monolithic-type fuels, which consist of a U-Mo foil encased in Al alloy cladding, are also being developed. These types of fuels are also showing promise and will continue to be developed. One challenge with this type of fuel is in trying to maximize the bond strength at the foil/cladding interface. Fuel/cladding interactions can affect the quality of the boding at this interface. Si is being added to improve the characteristics

  3. Manufacturing and investigation of U-Mo LEU fuel granules by hydride-dehydride processing

    International Nuclear Information System (INIS)

    Stetskiy, Y.A.; Trifonov, Y.I.; Mitrofanov, A.V.; Samarin, V.I.

    2002-01-01

    Investigations of hydride-dehydride processing for comminution of U-Mo alloys with Mo content in the range 1.9/9.2% have been performed. Some regularities of the process as a function of Mo content have been determined as well as some parameters elaborated. Hydride-dehydride processing has been shown to provide necessary phase and chemical compositions of U-Mo fuel granules to be used in disperse fuel elements for research reactors. Pin type disperse mini-fuel elements for irradiation tests in the loop of 'MIR' reactor (Dmitrovgrad) have been fabricated using U-Mo LEU fuel granules obtained by hydride-dehydride processing. Irradiation tests of these mini-fuel elements loaded to 4 g U tot /cm 3 are planned to start by the end of this year. (author)

  4. Thermal conductivity of fresh and irradiated U-Mo fuels

    Science.gov (United States)

    Huber, Tanja K.; Breitkreutz, Harald; Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Elgeti, Stefan; Reiter, Christian; Robinson, Adam. B.; Smith, Frances. N.; Wachs, Daniel. M.; Petry, Winfried

    2018-05-01

    The thermal conductivity of fresh and irradiated U-Mo dispersion and monolithic fuel has been investigated experimentally and compared to theoretical models. During in-pile irradiation, thermal conductivity of fresh dispersion fuel at a temperature of 150 °C decreased from 59 W/m·K to 18 W/m·K at a burn-up of 4.9·1021 f/cc and further to 9 W/m·K at a burn-up of 6.1·1021 f/cc. Fresh monolithic fuel has a considerably lower thermal conductivity of 15 W/m·K at a temperature of 150 °C and consequently its decrease during in-pile irradiation is less steep than for dispersion fuel. For a burn-up of 3.5·1021 f/cc of monolithic fuel, a thermal conductivity of 11 W/m·K at a temperature of 150 °C has been measured by Burkes et al. (2015). The difference of decrease for both fuels originates from effects in the matrix that occur during irradiation, like for dispersion fuel the gradual disappearance of the Al matrix with increased burn-up and the subsequent growth of an interaction layer (IDL) between the U-Mo fuel particle and Al matrix and subsequent matrix hardening. The growth of fission gas bubbles and the decomposition of the U-Mo crystal lattice also affect both dispersion and monolithic fuel.

  5. Irradiation performance of U-Mo monolithic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, M. K.; Gan, J.; Jue, J. F.; Keiser, D. D.; Perez, E.; Robinson, A.; Wachs, D. M.; Woolstenhulme, N. [Idaho National Laboratory, Idaho (Korea, Republic of); Kim, Y.S.; Hofman, G. L. [Argonne National Laboratory, Lemont (United States)

    2014-04-15

    High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  6. Irradiation of diffusion couples U-Mo/Al. Thermal calculation

    International Nuclear Information System (INIS)

    Fortis, Ana M.; Mirandou, Monica; Denis, Alicia C.

    2004-01-01

    The development of new low enrichment fuel elements for research reactors has lead to obtaining a number of compounds and alloys where the decrease in the enrichment is compensated by a higher uranium density in the fuel material. This has been achieved in particular with the uranium silicides dispersed in an aluminum matrix, where uranium densities about 4.8 g/cm 3 have been reached. Among the diverse candidate alloys, those of U-Mo with molybdenum content in the range 6 to 10 w % can yield, upon dispersion, to uranium densities of about 8 g/cm 3 . The first irradiation experiments employing these alloys in fuel plates, either dispersed in Al or monolithic revealed certain phenomena which are worthy of further studies. Failures have been detected apparently due to the formation of reaction products between the fissile material and the aluminum matrix, which exhibit a poor irradiation behavior. An experiment was designed which final purpose is to irradiate diffusion couples U-Mo/Al in the RA-3 reactor and to analyze the interaction zone at the working temperatures of the fuel elements. A simple device was built consisting of two Al 6063 blocks which press the U-Mo sample in between, located in an Al capsule. The ensemble is placed in a tube, which can be filled with different gases and introduced in the reactor. For safety reasons temperature predictions are necessary before performing the experiment. To this end, the COSMOS code was used. As a preliminary step and in order to test to exactness of the numerical estimations, two irradiations were performed in the RA-1 reactor with He and N 2 as transference gases. The agreement between the measured and calculated temperatures was good, particularly in the case of He and, along with the numerical predictions for the RA-3 reactor, provides a reliable basis to proceed with the following steps. (author)

  7. Irradiation performance of U-Mo-Ti and U-Mo-Zr dispersion fuels in Al-Si matrixes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Hofman, G.L. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B.; Wachs, D.M. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Ryu, H.J.; Park, J.M.; Yang, J.H. [Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2012-08-15

    Performance of U-7 wt.%Mo with 1 wt.%Ti, 1 wt.%Zr or 2 wt.%Zr, dispersed in an Al-5 wt.%Si alloy matrix, was investigated through irradiation tests in the ATR at INL and HANARO at KAERI. Post-irradiation metallographic features show that the addition of Ti or Zr suppresses interaction layer growth between the U-Mo and the Al-5 wt.%Si matrix. However, higher fission gas swelling was observed in the fuel with Zr addition, while no discernable effect was found in the fuel with Ti addition as compared to U-Mo without the addition. Known to have a destabilizing effect on the {gamma}-phase U-Mo, Zr, either as alloy addition or fission product, is ascribed for the disadvantageous result. Considering its benign effect on fuel swelling, with slight disadvantage from neutron economy point of view, Ti may be a better choice for this purpose.

  8. IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

    Directory of Open Access Journals (Sweden)

    M.K. MEYER

    2014-04-01

    Full Text Available High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  9. Performance evaluation of large U-Mo particle dispersed fuel irradiated in HANARO

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Oh, Seok Jin; Jang, Se Jung; Yu, Byung Ok; Lee, Choong Seong; Seo, Chul Gyo; Chae, Hee Taek; Kim, Chang Kyu

    2008-01-01

    U-Mo/Al dispersion fuel is being developed as advanced fuel for research reactors. Irradiation behavior of U-Mo/Al dispersion fuel has been studied to evaluate its fuel performance. One of the performance limiting factors is a chemical interaction between the U-Mo particle and the Al matrix because the thermal conductivity of fuel meat is decreased with the interaction layer growth. In order to overcome the interaction problem, large-sized U-Mo particles were fabricated by controlling the centrifugal atomization conditions. The fuel performance behavior of U-Mo/Al dispersion fuel was estimated by using empirical models formulated based on the microstructural analyses of the post-irradiation examination (PIE) on U-Mo/Al dispersion fuel irradiated in HANARO reactor. Temperature histories of U-Mo/Al dispersion fuel during irradiation tests were estimated by considering the effect of an interaction layer growth on the thermal conductivity of the fuel meat. When the fuel performances of the dispersion fuel rods containing U-Mo particles with various sizes were compared, fuel temperature was decreased as the average U-Mo particle size was increases. It was found that the dispersion of a larger U-Mo particle was effective for mitigating the thermal degradation which is associated with an interaction layer growth. (author)

  10. Characterization of the Microstructure of Irradiated U-Mo Dispersion Fuel with a Matrix that Contains Si

    International Nuclear Information System (INIS)

    Keiser, Jr. D.D.; Robinson, A.B.; Jue, J.F.; Medvedev, P.; Finlay, M.R.

    2009-01-01

    RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. Of particular interest is the irradiation performance of U-Mo dispersion fuels with Si added to the Al matrix. Si is added to improve the performance of U-Mo dispersion fuels. Microstructural examinations have been performed on fuel plates with Al-2Si matrix after irradiation to around 50% LEU burnup. Si-rich layers were observed in many areas around the various U-7Mo fuel particles. In one local area of one of the samples, where the Si-rich layer had developed into a layer devoid of Si, relatively large fission gas bubbles were observed in the interaction phase. There may be a connection between the growth of these bubbles and the amount of Si present in the interaction layer. Overall, it was found that having Si-rich layers around the fuel particles after fuel plate fabrication positively impacted the overall performance of the fuel plate

  11. Fuel Performance Modeling of U-Mo Dispersion Fuel: The thermal conductivity of the interaction layers of the irradiated U-Mo dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mistarhi, Qusai M.; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    U-Mo/Al dispersion fuel performed well at a low burn-up. However, higher burn-up and higher fission rate irradiation testing showed enhanced fuel meat swelling which was caused by high interaction layer growth and pore formation. The performance of the dispersion type fuel in the irradiation and un-irradiation environment is very important. During the fabrication of the dispersion type fuel an Interaction Layer (IL) is formed due to the inter-diffusion between the U-Mo fuel particles and the Al matrix which is an intermetallic compound (U,Mo)Alx. During irradiation, the IL becomes amorphous causing a further decrease in the thermal conductivity and an increase in the centerline temperature of the fuel meat. Several analytical models and numerical methods were developed to study the performance of the unirradiated U-Mo/Al dispersion fuel. Two analytical models were developed to study the performance of the irradiated U-Mo/Al dispersion fuel. In these models, the thermal conductivity of the IL was assumed to be constant. The properties of the irradiated U-Mo dispersion fuel have been investigated recently by Huber et al. The objective of this study is to develop a correlation for IL thermal conductivity during irradiation as a function of the temperature and fission density from the experimentally measured thermal conductivity of the irradiated U-Mo/Al dispersion fuel. The thermal conductivity of IL during irradiation was calculated from the experimentally measured data and a correlation was developed from the thermal conductivity of IL as a function of T and fission density.

  12. 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.

  13. Effects of irradiation on the interface between U-Mo and zirconium diffusion barrier

    Science.gov (United States)

    Jue, Jan-Fong; Keiser, Dennis D.; Miller, Brandon D.; Madden, James W.; Robinson, Adam B.; Rabin, Barry H.

    2018-02-01

    Irradiated fuel plates were characterized by microscopy that focused on the interface between U-Mo and Zr. Before irradiation, there were three major sub-layers identified in the U-Mo/Zr interface, namely, UZr2, Mo2Zr, and U with low Mo. The typical total thickness of this U-Mo/Zr interaction is 2-3 μm. The UZr2 sub-layer formed during fuel plate fabrication remains stable after irradiation, without large bubbles/porosity accumulation. However, this sub-layer becomes increasingly discontinuous as burnup increases. The low-Mo sub-layer exhibits numerous sub-micron bubbles/porosity at low burnup. Larger, interconnected porosity in this sub-layer was observed in a medium-burnup fuel specimen. However, at higher burnup, regions with the extra-large bubbles/porosity (i.e., larger than 5 μm) were observed in the U-Mo fuel foil at least 5 μm away from the original location of this sub-layer. The mechanism for the formation of the extra-large bubbles/porosity is still unclear at this time. In general, the U-Mo/Zr interface in monolithic U-Mo fuels is relatively stable after irradiation. No large detrimental defects, such as large interfacial bubbles or cracks/delamination, were observed in the fuel plates characterized.

  14. Irradiation behavior of the interaction product of U-Mo fuel particle dispersion in an Al matrix

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Hofman, G.L. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2012-06-15

    Highlights: Black-Right-Pointing-Pointer We in-pile tested U-Mo dispersion in Al matrix. Black-Right-Pointing-Pointer We observed interaction layer growth between U-Mo and Al and pore formation there. Black-Right-Pointing-Pointer Pores degrades thermal conductivity and structural integrity of the fueled zone. Black-Right-Pointing-Pointer The amorphous behavior of interaction layers is thought to be the main reason for unstable large pore growth. Black-Right-Pointing-Pointer A mechanism for pore formation and possible remedy to prevent it are proposed. - Abstract: Irradiation performance of U-Mo fuel particles dispersed in Al matrix is stable in terms of fuel swelling and is suitable for the conversion of research and test reactors from highly enriched uranium (HEU) to low enriched uranium (LEU). However, tests of the fuel at high temperatures and high burnups revealed obstacles caused by the interaction layers forming between the fuel particle and matrix. In some cases, fission gas filled pores grow and interconnect in the interdiffusion layer resulting in fuel plate failure. Postirradiation observations are made to examine the behavior of the interdiffusion layers. The interdiffusion layers show a fluid-like behavior characteristic of amorphous materials. In the amorphous interdiffusion layers, fission gas diffusivity is high and the material viscosity is low so that the fission gas pores readily form and grow. Based on the observations, a pore formation mechanism is proposed and potential remedies to suppress the pore growth are also introduced.

  15. An investigation on the irradiation behavior of atomized U-Mo/Al dispersion rod fuels

    International Nuclear Information System (INIS)

    Park, J.M.; Ryu, H.J.; Lee, Y.S.; Lee, D.B.; Oh, S.J.; Yoo, B.O.; Jung, Y.H.; Sohn, D.S.; Kim, C.K.

    2005-01-01

    The second irradiation fuel experiment, KOMO-2, for the qualification test of atomized U-Mo dispersion rod fuels with U-loadings of 4-4.5 gU/cc at KAERI was finished after an irradiation up to 70 at% U 235 peak burn-up and subjected to the IMEF (Irradiation material Examination Facility) for a post-irradiation analysis in order to understand the fuel irradiation performance of the U-Mo dispersion fuel. Current results for PIE of KOMO-2 revealed that the U-Mo/Al dispersion fuel rods exhibited a sound performance without any break-away swelling, but most of the fuel rods irradiated at a high linear power showed an extensive formation of the interaction phase between the U-Mo particle and the Al matrix. In this paper, the analysis of the PIE results, which focused on the diffusion related microstructures obtained from the optical and EPMA (Electron Probe Micro Analysis) observations, will be presented in detail. And a thermal modeling will be carried out to calculate the temperature of the fuel rod during an irradiation. (author)

  16. POST-IRRADIATION ANALYSES OF U-MO DISPERSION FUEL RODS OF KOMO TESTS AT HANARO

    Directory of Open Access Journals (Sweden)

    H.J. RYU

    2013-12-01

    Full Text Available Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4–5 g-U/cm3 were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr, additional protective coatings (silicide or nitride, and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

  17. Observation on the irradiation behavior of U-Mo alloy dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.; Park, Jong-Man

    2000-01-01

    Initial results from the postirradiation examination of high-density dispersion fuel test RERTR-3 are discussed. The U-Mo alloy fuels in this test were irradiated to 40% U-235 burnup at temperature ranging from 140 0 C to 240 0 C. Temperature has a significant effect on overall swelling of the test plates. The magnitude of the swelling appears acceptable and no unstable irradiation behavior is evident. (author)

  18. Neutronic feasibility studies using U-Mo dispersion fuel (9 Wt % Mo, 5.0 gU/cm3) for LEU conversion of the MARIA (Poland), IR-8 (Russia), and WWR-SM (Uzbekistan) research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Deen, J.R.; Hanan, N.A.; Matos, E.

    2000-01-01

    U-Mo alloys dispersed in an Al matrix offer the potential for high-density uranium fuels needed for the LEU conversion of many research reactors. On-going fuel qualification tests by the US RERTR Program show good irradiation properties of U-Mo alloy dispersion fuel containing 7-10 weight percent molybdenum. For the neutronic studies in this paper the alloy was assumed to contain 9 wt % Mo (U-9Mo) with a uranium density in the fuel meat of 5.00 gU/cm 3 which corresponds to 32.5 volume % U-9Mo. Fuels containing U-9Mo have been used in Russian reactors since the 1950's. For the three research reactors analyzed here, LEU fuel element thicknesses are the same as those for the Russian-fabricated HEU reference fuel elements. Relative to the reference fuels containing 80-90% enriched uranium, LEU U-9Mo Al-dispersion fuel with 5.00 gU/cm 3 doubles the cycle length of the MARIA reactor and increases the IR-8 cycle length by about 11%. For the WWR-SM reactor, the cycle length, and thus the number of fuel assemblies used per year, is nearly unchanged. To match the cycle length of the 36% enriched fuel currently used in the WWR-SM reactor will require a uranium density in the LEU U-9Mo Al-dispersion fuel of about 5.4 gU/cm 3 . The 5.00 gU/cm 3 LEU fuel causes thermal neutron fluxes in water holes near the edge of the core to decrease by (6-8)% for all three reactors. (author)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  20. Cross section TEM characterization of high-energy-Xe-irradiated U-Mo

    Energy Technology Data Exchange (ETDEWEB)

    Ye, B., E-mail: bye@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Ave. Lemont, IL 60439 (United States); Jamison, L.; Miao, Y. [Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Ave. Lemont, IL 60439 (United States); Bhattacharya, S. [Department of Materials Science and Engineering, Northwestern University, 2220 Campus Dr. Evanston, IL 60208 (United States); Hofman, G.L.; Yacout, A.M. [Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Ave. Lemont, IL 60439 (United States)

    2017-05-15

    U-Mo alloys irradiated with 84 MeV Xe ions to various doses were characterized with transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) techniques. The TEM thin foils were prepared perpendicular to the irradiated surface to allow a direct observation of the entire region modified by ions. Therefore, depth-selective microstructural information was revealed. Varied irradiation-induced phenomena such as gas bubble formation, phase reversal, and recrystallization were observed at different ion penetration depths in U-Mo. - Highlights: •Three distinct zones were observed along the ion traveling direction in U-7Mo irradiated with 84 MeV Xe ions at 350 °C. •The α-U particles within the Xe-implanted region were reverted to γ-U phase by irradiation. •High-density random intra-granular bubbles in a size of 4–5 nm were found in the irradiated region, coexisting with large inter-granular bubbles. •The high lattice stresses built up during the irradiation-induced phase reversal is probably the driving force for the small grain formation at cell boundaries.

  1. The irradiation behavior of atomized U-Mo alloy fuels at high temperature

    Science.gov (United States)

    Park, Jong-Man; Kim, Ki-Hwan; Kim, Chang-Kyu; Meyer, M. K.; Hofman, G. L.; Strain, R. V.

    2001-04-01

    Post-irradiation examinations of atomized U-10Mo, U-6Mo, and U-6Mo-1.7Os dispersion fuels from the RERTR-3 experiment irradiated in the Advanced Test Reactor (ATR) were carried out in order to investigate the fuel behavior of high uranium loading (8 gU/cc) at a high temperature (higher than 200°C). It was observed after about 40 at% BU that the U-Mo alloy fuels at a high temperature showed similar irradiation bubble morphologies compared to those at a lower temperature found in the RERTR-1 irradiation result, but there was a thick reaction layer with the aluminum matrix which was found to be greatly affected by the irradiation temperature and to a lesser degree by the fuel composition. In addition, the chemical analysis for the irradiated U-Mo fuels using the Electron Probe Micro Analysis (EPMA) method were conducted to investigate the compositional changes during the formation of the reaction product.

  2. Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

    Science.gov (United States)

    Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

    2018-04-01

    A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature- and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS). The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code.

  3. Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

    2018-04-01

    A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature-and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS).The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code. (c) 2018 Elsevier B.V. All rights reserved.

  4. The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Casella, Amanda J.; Casella, Andrew M.

    2017-04-01

    The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed. - Highlights: •Complementary fission gas release events are reported for U-Mo fuel with and without cladding. •Exothermic reaction between Zr diffusion layer and cladding influences fission gas release. •Mechanisms responsible for fission gas release are similar, but with varying timing and magnitude. •Behavior of samples is similar after 800 °C signaling the onset of superlattice destabilization.

  5. Elaboration of mini plates with U-Mo for irradiation in a high flux reactor

    International Nuclear Information System (INIS)

    Pasqualini, Enrique E.

    2005-01-01

    Full text: International new efforts for the reconversion of HEU in research, testing and radioisotopes production reactors, have greatly incremented U-Mo fuels qualification activities. These qualifications require the resolution of undesired interaction at high fluxes between UMo particles and the aluminum matrix in the case of dispersed fuels and the development of U-Mo monolithic fuels. These efforts are being manifested in the planning and execution of additional series of irradiation tests of mini plates and full size plates. Recently, CNEA has elaborated mini plates with different proposals for the irradiation at the ATR reactor (250 MWTH, maximum thermal neutron flux 10 15 n.cm -2 .seg -1 ) at Idaho National Laboratory, USA. Uranium 7% (w/w) molybdenum (U-7Mo) particles were coated with silicon. Chemical vapour deposition (CVD) of silane and high temperature diffusion of silicon were used. Hydrided, milled and dehydrated (HMD) particles heat treated at 1000 C degrees during four hours and centrifugal atomized powder were coated and the results compared. Mini plates were elaborated with both kinds of particles. Mini plates were also elaborated with U-7Mo and silicon particles dispersed in the aluminium matrix. Monolithic mini plates were also developed by co lamination of U-7Mo with a Zircaloy-4 cladding. The different steps of this process are detailed and the method is shown to be versatile, can be easily scaled up and is performed with small modifications of usual equipment in fuel plants. The irradiation experiment is called RERTR-7A, includes a total of 32 mini plates and it is planed to finalize by mid 2006. (author) [es

  6. Annealing tests of in-pile irradiated oxide coated U-Mo/Al-Si dispersed nuclear fuel

    Science.gov (United States)

    Zweifel, T.; Valot, Ch.; Pontillon, Y.; Lamontagne, J.; Vermersch, A.; Barrallier, L.; Blay, T.; Petry, W.; Palancher, H.

    2014-09-01

    U-Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800 °C of in-pile irradiated U-Mo/Al-Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500 °C and 670 °C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500 °C FGs are released from IDL/matrix interfaces. The second peak at 670 °C representing the main release of FGs originates from the interaction between U-Mo and matrix in the vicinity of the cladding.

  7. Interdiffusion between U-Mo alloys and Al or Al alloys at 340 deg. C. Irradiation plan

    International Nuclear Information System (INIS)

    Fortis, A.M.; Mirandou, M.; Ortiz, M.; Balart, S.; Denis, A.; Moglioni, A.; Cabot, P.

    2005-01-01

    Out of reactor interdiffusion experiments between U-Mo alloys and Al alloys made close to fuel operation temperature are needed to validate the results obtained above 500 deg. C. A study of interdiffusion between U-Mo and Al or Al alloys, out and in reactor, has been initiated. The objective is to characterize the interdiffusion layer around 250 deg. C and study the influence of neutron irradiation. Irradiation experiments will be performed in the Argentine RA3 reactor and chemical diffusion couples will be fabricated by Friction Stir Welding (FSW) technique. In this work out-of-pile diffusion experiments performed at 340 deg. C are presented. Friction Stir Welding (FSW) was used to fabricate some of the samples. One of the results is the presence of Si, in the interaction layer, coming from the Al alloy. This is promising in the sense that the absence of Al rich phases may also be expected at low temperature. (author)

  8. Irradiation performance of uranium-molybdenum alloy dispersion fuels; Desempenho sob irradiacao de elementos combustiveis do tipo U-Mo

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Cirila Tacconi de

    2005-07-01

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

  9. Fission gas behaviour and interdiffusion layer growth in in-pile and out-of-pile irradiated U-Mo/Al nuclear fuels

    International Nuclear Information System (INIS)

    Zweifel, Tobias

    2014-01-01

    Worldwide, research and test reactors are to convert their fuel from highly towards lower enriched uranium, among them the FRM II. One prospective fuel is an alloy of uranium and molybdenum (abbr. U-Mo). Test irradiations showed an insufficient irradiation behavior of this new fuel due to the growth of an interdiffusion layer (abbr. IDL) between the U-Mo fuel and the surrounding Al matrix. Furthermore, this layer accumulates fission gases. In this work, heavy ion irradiated U-Mo/Al layer systems were studied and compared to in-reactor irradiated fuel to study the fission gas dynamics. It is demonstrated that the gas behavior is identical for both in-reactor and out-of-reactor approaches.

  10. Thermal behavior analysis of U-Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Mang; Lee, Yoon Sang; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    According to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program, low enriched uranium(LEU) fuel such as uranium silicide dispersion fuels are being used in research reactors. Because of a lower enrichment higher uranium density fuels are required for some high performance research reactors. Some uranium alloys with a high uranium density such as U-Mo alloys have been considered as one of the most promising candidates for a dispersion fuel due to the good irradiation performance. An international qualification program to replace the uranium silicide dispersion fuel with U-Mo dispersion fuel is being carried out under the RERTR program. Although U-Mo powders are conventionally supplied by the mechanical comminuting of as-cast U-Mo alloys, KAERI developed a centrifugal atomization method in order to simplify the preparation process and improve the properties. The centrifugally atomized powders have a rapidly solidified gamma uranium structure and a spherical shape. During the in-reactor operation of a dispersion fuel, interdiffusion or chemical reactions between the fuel particles and the matrix occurr. Intermetallic compounds in the form of UAlx are formed as a result of the diffusional reaction. Because the intermetallic compounds are less dense than the combined reactants, the volume of the fuel element increases after the reaction. In addition to the effect on the swelling performance, the reaction layers between the U-Mo and the Al matrix induces a degradation of the thermal properties of the U-Mo/Al dispersion fuels. It is important to investigate the thermal behavior of U-Mo/Al dispersion fuel according to reaction between the fuel particles and the matrix with the burnup and linear power. In this study, a finite element analysis was used for the calculation of the temperature distribution of the U-Mo/Al dispersion fuel with a burnup and linear power. Kinetics data of the reaction layers such as the growth

  11. Thermal behavior analysis of U-Mo/Al dispersion fuel

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Mang; Lee, Yoon Sang; Kim, Chang Kyu

    2004-01-01

    According to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program, low enriched uranium(LEU) fuel such as uranium silicide dispersion fuels are being used in research reactors. Because of a lower enrichment higher uranium density fuels are required for some high performance research reactors. Some uranium alloys with a high uranium density such as U-Mo alloys have been considered as one of the most promising candidates for a dispersion fuel due to the good irradiation performance. An international qualification program to replace the uranium silicide dispersion fuel with U-Mo dispersion fuel is being carried out under the RERTR program. Although U-Mo powders are conventionally supplied by the mechanical comminuting of as-cast U-Mo alloys, KAERI developed a centrifugal atomization method in order to simplify the preparation process and improve the properties. The centrifugally atomized powders have a rapidly solidified gamma uranium structure and a spherical shape. During the in-reactor operation of a dispersion fuel, interdiffusion or chemical reactions between the fuel particles and the matrix occurr. Intermetallic compounds in the form of UAlx are formed as a result of the diffusional reaction. Because the intermetallic compounds are less dense than the combined reactants, the volume of the fuel element increases after the reaction. In addition to the effect on the swelling performance, the reaction layers between the U-Mo and the Al matrix induces a degradation of the thermal properties of the U-Mo/Al dispersion fuels. It is important to investigate the thermal behavior of U-Mo/Al dispersion fuel according to reaction between the fuel particles and the matrix with the burnup and linear power. In this study, a finite element analysis was used for the calculation of the temperature distribution of the U-Mo/Al dispersion fuel with a burnup and linear power. Kinetics data of the reaction layers such as the growth

  12. Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

    Energy Technology Data Exchange (ETDEWEB)

    Leenaers, A., E-mail: aleenaer@sckcen.be [Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Van den Berghe, S.; Koonen, E.; Kuzminov, V. [Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Detavernier, C. [Department of Solid State Sciences, Ghent University, Krijgslaan 281/S1, 9000 Ghent (Belgium)

    2015-03-15

    In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK• CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% {sup 235}U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL–matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium–Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)–matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

  13. Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

    Science.gov (United States)

    Leenaers, A.; Van den Berghe, S.; Koonen, E.; Kuzminov, V.; Detavernier, C.

    2015-03-01

    In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK•CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% 235U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL-matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)-matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

  14. Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

    2009-08-01

    A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

  15. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-12-15

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%{sup 235}U; the mini-rods were irradiated to an average burnup of ∼ 85%{sup 235}U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  16. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    International Nuclear Information System (INIS)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-01-01

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60% 235 U; the mini-rods were irradiated to an average burnup of ∼ 85% 235 U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%

  17. Post Irradiation TEM Investigation of ZrN Coated U(Mo) Particles Prepared with FIB

    Energy Technology Data Exchange (ETDEWEB)

    Van Renterghem, W.; Leenaers, A.; Van den Berghe, S.; Miller, B. D.; Gan, J.; Madden, J. W.; Keiser, D. D.; Palancher, H.; Hofman, G. L.; Breitkreuz, H.

    2015-10-01

    In the framework of the Selenium project, two dispersion fuel plates were fabricated with Si and ZrN coated fuel particles and irradiated in the Br2 reactor of SCK•CEN to high burn-up. The first analysis of the irradiated plate proved the reduced swelling of the fuel plate and interaction layer growth up to 70% burn-up. The question was raised how the structure of the interaction layer had been affected by the irradiation and how the structure of the fuel particles had evolved. Hereto, samples from the ZrN coated UMo particles were prepared for transmission electron microscopy (TEM) using focused ion beam milling (FIB) at INL. The FIB technique allowed to precisely select the area of the interaction layer and/or fuel to produce a sample that is TEM transparent over an area of 20 by 20 µm. In this contribution, the first TEM results will be presented from the 66% burn-up sample.

  18. COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

    Directory of Open Access Journals (Sweden)

    ALEKSEY. L. IZHUTOV

    2013-12-01

    The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; the mini-rods were irradiated to an average burnup of ∼ 85%235U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  19. COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

    OpenAIRE

    ALEKSEY. L. IZHUTOV; VALERIY. V. IAKOVLEV; ANDREY. E. NOVOSELOV; VLADIMIR. A. STARKOV; ALEKSEY. A. SHELDYAKOV; VALERIY. YU. SHISHIN; VLADIMIR. M. KOSENKOV; ALEKSANDR. V. VATULIN; IRINA. V. DOBRIKOVA; VLADIMIR. B. SUPRUN; GENNADIY. V. KULAKOV

    2013-01-01

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; th...

  20. The Microstructure of Multi-wire U-Mo Monolithic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Sang; Park, Eun Kee; Cho, Woo Hyoung; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    In order to use low-enriched uranium (LEU) instead of highly enriched uranium (HEU) for high performance research reactors, the reduced enrichment for research and test reactors (RERTR) program is developing high uranium density fuel such as U-Mo/Al dispersion fuel. U-Mo alloys have an excellent irradiation performance when compared to other uranium alloys or compounds. But the results from the post-irradiation examination of the U-Mo/Al dispersion fuels indicate that an interaction between the U-Mo alloy fuel and the Al matrix phases occurs readily during an irradiation and it is sensitively dependent on the temperature. In order to lessen these severe interactions, a concept of a multi-wire type fuel was proposed. The fuel configuration is that three to six U-Mo fuel wires (1.5 mm {approx} 2 mm in diameter) are symmetrically arranged at the periphery side in the Al matrix as shown. This multi-wire fuels showed very good fuel performance during the KOMO-3 irradiation test. At the KOMO-3 test, the specimen of the multi-wire fuels were U-7Mo/Al and U-7Mo-1Si/Al. In this study we investigate the microstructure change of the U-7Mo and U-7Mo-1Ti with some variation of annealing conditions. In addition to this, we want to check the effect of adding Ti element to U-7Mo on the gamma phase stability

  1. Swelling Estimation of Multi-wire U-Mo Monolithic Fuel for HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon-Sang; Ryu, Ho-Jin; Park, Jong-Man; Oh, Jong-Myeong; Kim, Chang-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    In order to use low-enriched uranium (LEU) instead of highly enriched uranium (HEU) for high performance research reactors, the reduced enrichment for research and test reactors (RERTR) program is developing high uranium density fuel such as U-Mo/Al dispersion fuel. U-Mo alloys have an excellent irradiation performance when compared to other uranium alloys or compounds. But the results from the post-irradiation examination of the U-Mo/Al dispersion fuels indicate that an interaction between the U-Mo alloy fuel and the Al matrix phases occurs readily during an irradiation and it is sensitively dependent on the temperature. In order to lessen these severe interactions, a concept of a multi-wire type fuel was proposed. The fuel configuration is that three to six U-Mo fuel wires (1.5 mm - 2 mm in diameter) are symmetrically arranged at the periphery side in the Al matrix. In this study temperature calculations and a swelling estimation of a multi-wire monolithic fuel were carried out. Also the results of a post irradiation analysis of this fuel will be introduced.

  2. Analyses on the U-Mo/Al Chemical Interaction and the Effects of Diffusion Barrier Coatings

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Kim, Woo Jeong; Cho, Woo Hyung; Jeong, Yong Jin; Lee, Yoon Sang; Park, Jong Man; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    While many HEU-fueled research reactors have been converted by adopting LEU U{sub 3}Si{sub 2} fuel in harmony with the Reduced Enrichment for Research and Test Reactors (RERTR) program, some high performance research reactors still need the development of advanced fuels with higher uranium densities. Currently, gamma-phase U-Mo alloys are considered promising candidates to be used as high uranium density fuel for the high performance reactors. For the production of UMo alloy powder, the centrifugal atomization technology developed by KAERI has been considered the most promising way because of high yield production and excellent powder quality when compared with other possible methods such as grinding, machining or hydriding-dehydriding. However, severe pore formation associated with an extensive interaction between the U-Mo and Al matrix, although the irradiation performance of U-Mo itself showed most stable, delay the fuel qualification of UMo fuel for high performance research reactors. Because the reaction products, i.e. uranium aluminides (UAlx), is less dense than the mixed reactants, the volume of the fuel meat increases after formation of interaction layer(IL). In addition to the impact on the swelling performance, the reaction layers between the U-Mo and Al matrix induces a degradation of the thermal conductivities of the U-Mo/Al dispersion fuels. The chemical interaction between the U-Mo and Al matrix are analyzed in this study to find remedies to reduce the growth of the interaction layers during irradiation. In addition, various coating technologies for the formation of diffusion barriers on U-Mo particles are proposed as a result of the analyses

  3. 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)

  4. Reduced interaction layer growth of U-Mo dispersion in Al-Si

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Park, Jong Man; Ryu, Ho Jin; Jung, Yang Hong [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong, Daejeon 305-353 (Korea, Republic of); Hofman, G.L. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2012-11-15

    Development of high U-density U-Mo fuel particle dispersion in Al is needed to convert high power research and test reactors from HEU to LEU. Interaction layer growth between U-Mo and Al poses a challenge to this goal. The KOMO-4 test was designed at KAERI and irradiated in the HANARO reactor to {approx}50% burnup of initial 19.75% U-235 enrichment at {approx}200 Degree-Sign C. The main objective of the test was to examine the effect of the Si content in the matrix up to 8 wt.%. U-Mo/Al-Si dispersion samples with a Si addition in the range 0-8 wt.% in the matrix were tested. A sample with pre-irradiation Si-containing interaction layers (ILs) was also tested. As the Si content in the matrix increases, the IL growth was progressively reduced. Contrary to the thermodynamics prediction and out-of-pile observations, however, Si accumulation in the ILs occurred near the IL-matrix interface with only a slight increase in concentration. The effect of the pre-formed ILs was insignificant in reducing IL growth.

  5. TEM and XAS investigation of fission gas behaviors in U-Mo alloy fuels through ion beam irradiation

    Science.gov (United States)

    Zang, Hang; Yun, Di; Mo, Kun; Wang, Kunpeng; Mohamed, Walid; Kirk, Marquis A.; Velázquez, Daniel; Seibert, Rachel; Logan, Kevin; Terry, Jeffrey; Baldo, Peter; Yacout, Abdellatif M.; Liu, Wenbo; Zhang, Bo; Gao, Yedong; Du, Yang; Liu, Jing

    2017-10-01

    In this study, smaller-grained (hundred nano-meter size grain) and larger-grained (micro-meter size grain) U-10Mo specimens have been irradiated (implanted) with 250 keV Xe+ beam and were in situ characterized by TEM. Xe bubbles were not seen in the specimen after an implantation fluence of 2 × 1020 ions/m2 at room temperature. Nucleation of Xe bubbles happened during heating of the specimen to a final temperature of 300 °C. By comparing measured Xe bubble statistics, the nucleation and growth behaviors of Xe bubbles were investigated in smaller-grained and larger-grained U-10Mo specimens. A multi-atom kind of nucleation mechanism has been observed in both specimens. X-ray Absorption spectroscopy showed the edge position in the bubbles to be the same as that of Xe gas. The size of Xe bubbles has been shown to be bigger in larger-grained specimens than in smaller-grained specimens at the same implantation conditions.

  6. Study of relationships between microstructures and service properties, of U(Mo) fissile alloys particles

    International Nuclear Information System (INIS)

    Champion, G.

    2013-01-01

    This thesis enters in the Material and Testing Reactors (MTRs) framework where the necessity to use a Low- Enriched Uranium (LEU) fuel has led to the development of a dense fissile material based on U(Mo) alloys. The designed fuel is a composite material, made of dispersed U(Mo) particles embedded in an Al based matrix. Post- Irradiation Examinations of these LEU fuel plates showed that the irradiation behaviour of the fuel is not fit for purpose yet. This is mainly due to the growth of an interaction layer between the fuel and the matrix and to the bad gas retention efficiency of the fuel particles. This thesis had for purpose the development of several solutions in order to modify and/or decrease or even inhibit the fuel/matrix interaction and to increase the gas retention capacities of the fuel. In order to achieve so, two solutions have been tested during this thesis, (i) optimization of the U(Mo) alloy intrinsic microstructural properties and (ii) modification of the fuel meat/matrix interface, through the deposition of a layer acting as a 'diffusion barrier'. Concerning the first axis of study, a characterization campaign of the reference powders has been performed, as a first step, in order to identify the key parameters for the development of products showing an 'optimized' microstructure. Two novel products have then been developed: one based on a combined process associating 'atomization + grinding' and another, which consists in a magnesiothermy process. These products were subjected to characterization: X-Ray and neutron diffraction, electron backscattered diffraction and transmission electron microscopy have been performed in particular. We managed to show that these powders can be an advantage concerning the issue with the gas retention capacities of the fuel. Concerning the growth of the interaction layer, a third product has been developed: an U(Mo) atomized powder, coated with an alumina layer. We managed to show that a thickness between 100 and

  7. The Effect of Uncertainties on the Operating Temperature of U-Mo/Al Dispersion Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sweidana, Faris B.; Mistarihia, Qusai M.; Ryu Ho Jin [KAIST, Daejeon (Korea, Republic of); Yim, Jeong Sik [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, uncertainty and combined uncertainty studies have been carried out to evaluate the uncertainty of the parameters affecting the operational temperature of U-Mo/Al fuel. The uncertainties related to the thermal conductivity of fuel meat, which consists of the effects of thermal diffusivity, density and specific heat capacity, the interaction layer (IL) that forms between the dispersed fuel and the matrix, fuel plate dimensions, heat flux, heat transfer coefficient and the outer cladding temperature were considered. As the development of low-enriched uranium (LEU) fuels has been pursued for research reactors to replace the use of highly-enriched uranium (HEU) for the improvement of proliferation resistance of fuels and fuel cycle, U-Mo particles dispersed in an Al matrix (UMo/Al) is a promising fuel for conversion of the research reactors that currently use HEU fuels to LEUfueled reactors due to its high density and good irradiation stability. Several models have been developed for the estimation of the thermal conductivity of U–Mo fuel, mainly based on the best fit of the very few measured data without providing uncertainty ranges. The purpose of this study is to provide a reasonable estimation of the upper bounds and lower bounds of fuel temperatures with burnup through the evaluation of the uncertainties in the thermal conductivity of irradiated U-Mo/Al dispersion fuel. The combined uncertainty study using RSS method evaluated the effect of applying all the uncertainty values of all the parameters on the operational temperature of U-Mo/Al fuel. The overall influence on the value of the operational temperature is 16.58 .deg. C at the beginning of life and it increases as the burnup increases to reach 18.74 .deg. C at a fuel meat fission density of 3.50E+21 fission/cm{sup 3}. Further studies are needed to evaluate the behavior more accurately by including other parameters uncertainties such as the interaction layer thermal conductivity.

  8. U-Mo/Al-Si interaction: Influence of Si concentration

    International Nuclear Information System (INIS)

    Allenou, J.; Palancher, H.; Iltis, X.; Cornen, M.; Tougait, O.; Tucoulou, R.; Welcomme, E.; Martin, Ph.; Valot, C.; Charollais, F.; Anselmet, M.C.; Lemoine, P.

    2010-01-01

    Within the framework of the development of low enriched nuclear fuels for research reactors, U-Mo/Al is the most promising option that has however to be optimised. Indeed at the U-Mo/Al interfaces between U-Mo particles and the Al matrix, an interaction layer grows under irradiation inducing an unacceptable fuel swelling. Adding silicon in limited content into the Al matrix has clearly improved the in-pile fuel behaviour. This breakthrough is attributed to an U-Mo/Al-Si protective layer around U-Mo particles appeared during fuel manufacturing. In this work, the evolution of the microstructure and composition of this protective layer with increasing Si concentrations in the Al matrix has been investigated. Conclusions are based on the characterization at the micrometer scale (X-ray diffraction and energy dispersive spectroscopy) of U-Mo7/Al-Si diffusion couples obtained by thermal annealing at 450 deg. C. Two types of interaction layers have been evidenced depending on the Si content in the Al-Si alloy: the threshold value is found at about 5 wt.% but obviously evolves with temperature. It has been shown that for Si concentrations ranging from 2 to 10 wt.%, the U-Mo7/Al-Si interaction is bi-layered and the Si-rich part is located close to the Al-Si for low Si concentrations (below 5 wt.%) and close to the U-Mo for higher Si concentrations. For Si weight fraction in the Al alloy lower than 5 wt.%, the Si-rich sub-layer (close to Al-Si) consists of U(Al, Si) 3 + UMo 2 Al 20 , when the other sub-layer (close to U-Mo) is silicon free and made of UAl 3 and U 6 Mo 4 Al 43 . For Si weight concentrations above 5 wt.%, the Si-rich part becomes U 3 (Si, Al) 5 + U(Al, Si) 3 (close to U-Mo) and the other sub-layer (close to Al-Si) consists of U(Al, Si) 3 + UMo 2 Al 20 . On the basis of these results and of a literature survey, a scheme is proposed to explain the formation of different types of ILs between U-Mo and Al-Si alloys (i.e. different protective layers).

  9. Modeling of the behavior under fuel dispersed irradiation of U-Mo with aluminum matrix from the thermal point of view and its interrelationship with the interdiffusion phase fuel / matrix

    International Nuclear Information System (INIS)

    Moscarda, Maria V.; Taboada, Horacio H.; Rest, J.

    2009-01-01

    Results from postirradiation examinations of U-Mo / Al dispersion fuels plates denotes a strong interrelation and feedback between the fuel-matrix interaction and the fuel temperature, bringing undesired consequences on the total swelling and behavior under irradiation. The present work approaches this problem, modeling the profile of temperatures moment by moment to be able to evaluate the increase of this interaction. The Fast Dart program is used, optimized version of program Dart, developed by Dr. J. Rest in collaboration with Dr. H. Taboada. A subroutine of thermal calculation was implemented in this code, which allowed to calculate the evolution of the interaction between the fuel and the matrix. The results of simulations are compared with the results of postirradiation examinations realized by the Reduced Enrichment for Research and Test Reactors International Program. In particular, a good adjustment in the calculation of the depth of interdiffusion U-Mo/Al is observed, demonstrating a right estimation of the profile of temperatures on the fuel plate. It is considered necessary the inclusion of a model that describes the phases that form in the zone of interaction, denoting its thermal dependency and effects due to the radiation damage. (author)

  10. U-Mo fuel qualification program in HANARO

    International Nuclear Information System (INIS)

    Lee, K.H.; Lee, C.S.; Kim, H.R.; Kuk, I.H.; Kim, C.K.

    2000-01-01

    Atomized U-Mo fuel has shown good performance from the results of previous out-of-pile tests and post-irradiation examinations. A qualification program of rod type U-Mo fuel is in progress and the fuel will be irradiated in HANARO. 6 gU/cm 3 U-7Mo, U-8Mo and U-9Mo are considered in this program. The laboratory test results of porosity, mechanical property, thermal conductivity, and thermal compatibility test are discussed in this paper. In parallel with this qualification program, the feasibility study on the core conversion from the present U 3 Si fuel to U-Mo in HANARO will be initiated to provide technical bases for the policy making. Several options of core conversion for HANARO are proposed and each option will be addressed briefly in terms of the operation policy, fuel management, and licensing of HANARO. (author)

  11. Thermal analysis of LEU modified Cintichem target irradiated in TRIGA reactor

    International Nuclear Information System (INIS)

    Catana, A; Toma, C.

    2009-01-01

    Actions conceived during last years at international level for conversion of Molybdenum fabrication process from HEU to LEU targets utilization created opportunities for INR to get access to information and participating to international discussions under IAEA auspices. Concrete steps for developing fission Molybdenum technology were facilitated. Institute of Nuclear Research bringing together a number of conditions like suitable irradiation possibilities, direct communication between reactor and hot cell facility, handling capacity of high radioactive sources, and simultaneously the existence of an expanding internal market, decided to undertake the necessary steps in order to produce fission molybdenum. Over the course of last years of efforts in this direction we developed the steps for fission Molybdenum technology development based on modified Cintichem process in accordance with the Argonne National Laboratory proved methodology. Progress made by INR to heat transfer computations of annular target using is presented. An advanced thermal-hydraulic analysis was performed to estimate the heat removal capability for an enriched uranium (LEU) foil annular target irradiated in TRIGA reactor core. As a result, the present analysis provides an upper limit estimate of the LEU-foil and external target surface temperatures during irradiation in TRIGA 14 MW reactor. (authors)

  12. U-Mo fuels handbook. Version 1.0

    International Nuclear Information System (INIS)

    Rest, Jeffrey; Kim, Yeon Soo; Hofman, Gerard L.; Meyer, Mitchell K.; Hayes, Steven L.

    2006-01-01

    This handbook provides an overview of property data and fuel performance topics with an emphasis on data available for U-Mo alloys. These data often exist only in report format and have not been widely disseminated in the journal literature. For some topics there is more than one source of data, which are sometimes inconsistent. In this situation, the authors have attempted to select the best dataset to provide a standard for fuel designers and reactor operators. Following the section on unirradiated and irradiated materials properties for the monolithic U-Mo alloy, property data for cladding and matrix aluminum are presented. Property data for cladding aluminum are more widely available, and are not presented in great depth. Finally, some properties of (U-Mo)/Al dispersions are also included in this document. Where no data are available, best estimate correlations are provided. Best fits to the data are presented in order to facilitate use by fuel designers and reactor operators.

  13. 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

  14. Irradiation of MEU and LEU test fuel elements in DR 3

    International Nuclear Information System (INIS)

    Haack, K.

    1984-01-01

    Irradiation of three MEU and three LEU fuel elements in the Danish reactor DR 3. Thermal and fast neutron flux density scans of the core have been made and the results, related to the U235-content of each fuel element, are compared with the values from HEU fuel elements. The test elements were taken to burn-up percentages of 50-60%. Reactivity values of the test elements at charge and at discharge have been measured and the values are compared with those of HEU fuel elements. (author)

  15. 2010 national progress report on R and D on LEU fuel and target technology in Argentina

    International Nuclear Information System (INIS)

    Balart, S.; Blaumann, H.; Cristini, P.; Gonzalez, A.G.; Gonzalez, R.; Hermida, J.D.; Lopez, M.; Mirandou, M.; Taboada, H.

    2010-01-01

    Since last RRFM meeting, CNEA has deployed several related tasks. The RA-6 MTR type reactor, converted its core from HEU to a new LEU silicide one is scaling up the power, according to a protocol requested by the national regulatory body, ARN. CNEA is deploying an intense R and D activity to fabricate both dispersed U-Mo (Al-Si matrix and Al cladding) and monolithic (Zry-4 cladding) miniplates to develop possible solutions to VHD dispersed and monolithic fuels technical problems. Some monolithic 58% enrichment U8%Mo and U10%Mo are being delivered to INL-DoE to be irradiated in ATR reactor core. A conscientious study on compound interphase formation in both cases is being carried out. CNEA, a worldwide leader on LEU technology for fission radioisotope production is providing Brazil with these radiopharmaceutical products and Egypt and Australia with the technology through INVAP SE. CNEA is also committed to improve the diffusion of LEU target and radiochemical technology for radioisotope production and target and process optimization. Future plans include: 1) Fabrication of a LEU dispersed U-Mo fuel prototype following the recommendations of the IAEA's Good Practices document, to be irradiated in a high flux reactor in the frame of the ARG/4/092 IAEA's Technical Cooperation project. 2) Development of LEU very high density monolithic and dispersed U-Mo fuel plates with Zry-4 or Al cladding as a part of the RERTR program. 3) Optimization of LEU target and radiochemical techniques for radioisotope production. (author)

  16. Pore growth in U-Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Jeong, G.Y.; Sohn, D.-S. [Ulsan National Institute of Science and Technology, 50 UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan, 689-798 (Korea, Republic of); Jamison, L.M. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2016-09-15

    U-Mo/Al dispersion fuel is currently under development in the DOE’s Material Management and Minimization program to convert HEU-fueled research reactors to LEU-fueled reactors. In some demanding conditions in high-power and high-performance reactors, large pores form in the interaction layers between the U-Mo fuel particles and the Al matrix, which pose a potential to cause fuel failure. In this study, comprehension of the formation and growth of these pores was explored. As a product, a model to predict pore growth and porosity increase was developed. The model includes three major topics: fission gas release from the U-Mo and the IL to the pores, stress evolution in the fuel meat, and the effect of amorphous IL growth. Well-characterized in-pile data from reduced-size plates were used to fit the model parameters. A data set from full-sized plates, independent and distinctively different from those used to fit the model parameters, was used to examine the accuracy of the model. The model showed fair agreement with the measured data. The model suggested that the growth of the IL has a critical effect on pore growth, as both its material properties and energetics are favorable to pore formation. Therefore, one area of the current effort, focused on suppressing IL growth, appears to be on the right track to improve the performance of this fuel.

  17. INFLUENCE OF FUEL-MATRIX INTERACTION ON THE BREAKAWAY SWELLING OF U-MO DISPERSION FUEL IN AL

    OpenAIRE

    HO JIN RYU; YEON SOO KIM

    2014-01-01

    In order to advance understanding of the breakaway swelling behavior of U-Mo/Al dispersion fuel under a high-power irradiation condition, the effects of fuel-matrix interaction on the fuel performance of U-Mo/Al dispersion fuel were investigated. Fission gas release into large interfacial pores between interaction layers and the Al matrix was analyzed using both mechanistic models and observations of the post-irradiation examination results of U-Mo dispersion fuels. Using the model prediction...

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  19. Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

    2016-06-01

    Low-enriched (U-235 RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

  20. Scanning electron microscopy analysis of fuel/matrix interaction layers in highly-irradiated U-Mo dispersion fuel plates with Al and Al-Si alloy matrices

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, Dennis D. Jr; Jue, Jan Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adom B.; Medvedev, Pavel; Madden, James; Wachs, Dan; Meyer, Mitch [Nuclear Fuels and Materials Division, Idaho National Laboratory (United States)

    2014-04-15

    In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U-7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifically, samples from irradiated U-7Mo dispersion fuel elements with pure Al, Al-2Si and AA4043 (-4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U-7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger) and shape (round) of the fission gas bubbles. Additionally, solid fission product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U-7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al-Si matrices.

  1. Miniplates irradiation in the ATR (Idaho, USA)

    International Nuclear Information System (INIS)

    Pasqualini, Enrique E.

    2007-01-01

    High density U Mo alloys are promising for its utilization in the reconversion of HEU fuels to LEU for research nuclear reactors. Ought to the thermomechanical properties of the alloy U Mo and its interaction with aluminium it is necessary to develop new technologies and fabrication procedures to qualify this material as a nuclear fuel. In this work a review is made about the evolution of the idea and PIE experiments of monolithic LEU U 7 Mo fuel with Zr-4 cladding. The irradiation took place in the frame of international qualification efforts of dispersed and monolithic U Mo fuels. Dispersed and monolithic fuels, elaborated and in intermediate steps of development, are discussed. (author) [es

  2. Analyses of Interaction Phases of U Mo Dispersion Fuel by Synchrotron X ray Diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woo Jeong; Nam, Ji Min; Ryu, Ho Jin; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Herve, Palancher; Charollais, Francois [Saint Paul Lez Durance Cedex, Rhone (France); Bonnin, Anne; Honkimaeki, Veijo [Grenoble Cedex, Grenoble (France); Patrick Lemoined [Gif sur Yvette, Paris (France)

    2012-10-15

    Gamma phase U Mo alloys are one of the promising candidates to be used as advanced high uranium density fuel for high power research reactors due to their excellent irradiation performance. However, formation of interaction layers between the U Mo particles and Al matrix degrades the irradiation performance of U Mo dispersion fuel. One of the remedies to the interaction problem is a Si addition to the Al matrix. Recent irradiation tests have shown that the use of Al (2{approx}5wt%)Si matrices retarded the growth of interaction layers effectively during irradiation. Recently, KAERI has proposed silicide or nitride coated U Mo fuel for the minimization of the interaction layer growth. The silicide or nitride coatings are expected to act as interdiffusion barriers and their out of pile tests showed the improved diffusion barrier performances of the silicide and nitride layers. In order to characterize constituent phases in the coated layers on U Mo particles and the interaction layers of coated U Mo particle dispersed fuel, synchrotron X ray diffraction experiments have been performed at the ESRF (European Synchrotron Radiation Facility), France as a KAERI CEA cooperation program.

  3. Development of U-Mo Research Reactor Fuel for Next Generation

    International Nuclear Information System (INIS)

    Park, Jong Man; Lee, Y. S.; Yang, J. H.; Ryu, H. J.; Kim, C. K.; Chae, H. T.; Seo, C. G.

    2010-08-01

    - Exportation of centrifugal atomized U-Mo powder - Completion of post irradiation examination for KOMO-3 irradiated fuel rods. - Select the dispersion fuel rod candidates for KOMO-4 irradiation test. - Irradiation test to solve the problems of interaction layer formation (KOMO-4) - Set the post irradiation examination of KOMO-4 irradiated fuel rods. - Development and characterization of innovative high U density fuel rods - Obtain and analyze foreign new irradiation test D

  4. Phase transformation of metastable cubic γ-phase in U-Mo alloys

    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 past decade considerable efforts have been put by many fuel designers to develop low enriched uranium (LEU 235 ) base U-Mo alloy as a potential fuel for core conversion of existing research and test reactors which are running on high enriched uranium (HEU > 85%U 235 ) fuel and also for the upcoming new reactors. U-Mo alloy with minimum 8 wt% molybdenum shows excellent metastability with cubic γ-phase in cast condition. However, it is important to characterize the decomposition behaviour of metastable cubic γ-uranium in its equilibrium products for in reactor fuel performance point of view. The present paper describes the phase transformation behaviour of cubic γ-uranium phase in U-Mo alloys with three different molybdenum compositions (i.e. 8 wt%, 9 wt% and 10 wt%). U-Mo alloys were prepared in an induction melting furnace and characterized by X-ray diffraction (XRD) method for phase determination. Microstructures were developed for samples in as cast condition. The alloys were hot rolled in cubic γ-phase to break the cast structure and then they were aged at 500 o C for 68 h and 240 h, so that metastable cubic γ-uranium will undergo eutectoid decomposition to form equilibrium phases of orthorhombic α-uranium and body centered tetragonal U 2 Mo intermetallic compound. U-Mo alloy samples with different ageing history were then characterized by XRD for phase and development of microstructure.

  5. LEU-plate irradiation at FRJ-2 (DIDO) under the German AF-programme

    Energy Technology Data Exchange (ETDEWEB)

    Groos, E; Krug, W; Seferiadis, J; Thamm, G

    1985-07-01

    10 LEU fuel plates (8 with uranium silicides max. U-density 6.1 g/cm{sup 3}) have been irradiated at FRJ-2 (DIDO) of KFA-Juelich till end of October 1984 during 321 full power days up to max. burnup of 2.41x10{sup 27} fissions/m{sup 3} without major interruptions and troubles. PIE began recently in KFA hot cells. Visual inspections revealed no damage or greater deformation for the majority of the plates, but red/brown coloured layers (partially peeled off) on the cladding over the fuel. Aluminium (oxide) is the chief constituent of the layer with smaller portions of Ni and Fe the latter causing the red/brown colour. The major part of the layer ({approx}50 {mu}m) most probably has been formed during 20 h immediately after experiment start-up under abnormal conditions of the coolant water. Gamma scanning has been completed. Dimensional measurements are under way confirming first observations of severe swelling (pillowing) of 1 plate. Density and blister testing as well as metallography and burnup analysis remain to be accomplished end of 1985/beginning of 1986. (author)

  6. XRD and neutron diffraction analyses of heat treated U-Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Ji Min; Kim, Woo Jeong; Ryu, Ho Jin; Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    High density U Mo alloys are regarded as promising candidates for advanced research reactor fuel because they have shown stable irradiation performance when compared to other uranium alloys and compounds. However, interaction layer formation between the U Mo alloys and Al matrix degrades the irradiation performance of U Mo dispersion fuel. Therefore, addition of Ti in U Mo alloys, addition of Si in Al matrix and silicide or nitride coating on the surface of U Mo particles have been proposed in order to inhibit the interaction layer growth. In order to analyze the mechanisms of interaction layer growth inhibition by adding Ti in U Mo alloys or Si in Al matrix, accurate phase characterization of the interaction layers is required. While previous studies using X ray diffraction have been reported, laboratory X ray diffraction method has limitations such as low resolution and small measurement volume. Neutron diffraction method can be a more accurate analysis when compared with X ray diffraction method due to the large penetration depth of neutron. In this study, X ray diffraction and neutron diffraction experiments have been performed by using the laboratory X ray diffractometer and high resolution powder diffractometer (HRPD) of the HANARO research reactor in KAERI.

  7. Influence of fuel-matrix interaction on the breakaway swelling of U-Mo dispersion fuel in Al

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Yeon Soo [Nuclear Engineering Division, Argonne National Laboratory, Arogonne (United States)

    2014-04-15

    In order to advance understanding of the breakaway swelling behavior of U-Mo/Al dispersion fuel under a high-power irradiation condition, the effects of fuel-matrix interaction on the fuel performance of U-Mo/Al dispersion fuel were investigated. Fission gas release into large interfacial pores between interaction layers and the Al matrix was analyzed using both mechanistic models and observations of the post-irradiation examination results of U-Mo dispersion fuels. Using the model predictions, advantageous fuel design parameters are recommended to prevent breakaway swelling.

  8. DART-TM: A thermomechanical version of DART for LEU VHD dispersed and monolithic fuel analysis

    International Nuclear Information System (INIS)

    Saliba, Roberto; Taboada, Horacio; Moscarda, Ma.Virginia; Rest, Jeff

    2003-01-01

    A collaboration agreement between ANL/USDOE and CNEA Argentina, in the area of Low Enriched Uranium Advanced Fuels has been in place since October 16, 1997 under the 'Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy'. An annex concerning DART code optimization has been operative since February 8, 1999. Previously, as a part of this annex a visual thermal FASTDART version was developed that includes mechanistic models for the calculation of the fission-gas-bubble and fuel particle size distribution, reaction layer thickness, and meat thermal conductivity. FASTDART was presented at the last RERTR Meeting that included validation against RERTR 3 irradiation data. The thermal FASTDART version was assessed as an adequate tool for modeling the behavior of LEU U-Mo dispersed fuels under irradiation against PIE RERTR irradiation data. During this past year the development of a 3-D thermo-mechanical version of the code for modeling the irradiation behavior of LEU U-Mo monolithic and dispersion fuel was initiated. Some preliminary results of this work will be shown during RERTR-2003 meeting. (author)

  9. Development of U-Mo/Al dispersion fuel for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Man; Ryu, Ho Jin; Yang, Jae Ho; Jeong, Yong Jin; Lee, Yoon Sang [Korea Atomic Energy Research Inst., Research Reactor Fuel Development Division, Daejeon (Korea, Republic of)

    2012-03-15

    Currently, the KOMO-5 irradiation test for full size U-Mo/Al dispersion fuel rods has been underway since May 23, 2011. The purpose of the KOMO-5 test includes an investigation of the irradiation behaviors of silicide or nitride coated U-7Mo/Al(-Si) dispersion fuels and the effects of pre-formed interaction layers on U-Mo particles. It is expected that the irradiation test will be finished after attaining 60 at% U-235 burnup in May 2012, and the first PIE results of the KOMO-5 will be obtained in September 2012. In addition, an international cooperation program on the qualification of U-Mo dispersion fuels for small and medium size research reactors is going to be proposed in cooperation with the IAEA. Conversion from silicide fuel to U-Mo fuel will increase the cycle length with a smaller number of fuel assemblies and allow more flexible back-end options for spent fuel due to of the reprocessibility of U-Mo. (author)

  10. Progress on LEU very high density fuel and target development in Argentina

    International Nuclear Information System (INIS)

    Balart, S.; Cabot, P.; Calzetta, O.; Duran, A.; Garces, J.; Hermida, J.D.; Manzini, A.; Pasqualini, E.; Taboada, H.

    2006-01-01

    Since last RRFM meeting, CNEA has continued on new LEU fuel and target development activities. Main goals are the plan to convert our RA-6 reactor from HEU to a new LEU core, to get a comprehensive understanding of U-Mo/Al compounds phase formation in dispersed and monolithic fuels, to develop possible solutions to VHD dispersed and monolithic fuels technical problems, to optimize techniques to recover U from silicide scrap samples as cold test for radiowaste separation for final conditioning of silicide spent fuels. and to improve the diffusion of LEU target and radiochemical technology for radioisotope production. Future plans include: - Completion of the RA-6 reactor conversion to LEU; - Improvement on fuel development and production facilities to implement new technologies, including NDT techniques to assess bonding quality; - Irradiation of miniplates and full scale fuel assembly at RA-3 and plans to perform irradiation on higher power and temperature regime reactors; - Optimization of LEU target and radiochemical techniques for radioisotope production. (author)

  11. Evolution of microstructure of U-Mo alloys in as cast and sintered forms

    International Nuclear Information System (INIS)

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

    2009-01-01

    Over the years U 3 Si 2 compound dispersed in aluminium matrix has been successfully used as potential Low Enriched Uranium (LEU 235 ) base dispersion fuel in new research and test reactors and also for converting High Enriched Uranium (HEU > 85% U 235 ) cores to LEU in most of the existing research and test reactors. The maximum density achievable with U 3 Si 2 -AI dispersion fuel is around 4.8 g U cm -3 . To achieve a uranium density of 8.0 to 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 Metallic Fuels Division, R and D efforts are on to develop these high density uranium alloys. Molybdenum plays a crucial role in metastabilising the γ-phase of uranium at room temperature which is very much evident when we see the microstructures of different U-Mo alloys with varying molybdenum concentration as solute atom. The paper describes the role of molybdenum in imparting metastability in U-Mo alloys from their microstructures in as cast and sintered forms. The paper also covers the role of tailored microstructure in U-Mo alloy for the purpose of hydriding and dehydriding treatment to generate alloy powders. (author)

  12. Preliminary investigation of the use of monolithic U-Mo fuel in the MIT reactor

    International Nuclear Information System (INIS)

    Newton, Thomas H. Jr.; Kazimi, Mujid S.; Pilat, Edward E.; Xu Zhiwen

    2003-01-01

    Studies have begun on the use of monolithic LEU U-Mo fuel in the MIT Nuclear Research Reactor (MITR-II) using the Monte Carlo Transport code MCNP. These studies have included model benchmarking, LEU fuel optimization, burnup evaluation, in-core facility design, and determination of safety attributes. Benchmarking studies on the initial core have shown favorable agreement between the calculated and measured reactivity worths of the six control blades. In addition, optimization studies on LEU U7Mo MITR-II fuel have shown that an arrangement of ten to twelve plates per fuel element would have initial reactivity values and thermal neutron fluxes comparable to the current HEU core. Burnup studies which have been made using the MCODE depletion program will be presented. Safety attributes such as temperature coefficients, shutdown margins, and coolant subcooled margin are under evaluation. (author)

  13. 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.

  14. Characterization of U-Mo Foils for AFIP-7

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, Danny J.; Ermi, Ruby M.; Schemer-Kohrn, Alan L.; Overman, Nicole R.; Henager, Charles H.; Burkes, Douglas; Senor, David J.

    2012-11-07

    Twelve AFIP in-process foil samples, fabricated by either Y-12 or LANL, were shipped from LANL to PNNL for potential characterization using optical and scanning electron microscopy techniques. Of these twelve, nine different conditions were examined to one degree or another using both techniques. For this report a complete description of the results are provided for one archive foil from each source of material, and one unirradiated piece of a foil of each source that was irradiated in the Advanced Test Reactor. Additional data from two other LANL conditions are summarized in very brief form in an appendix. The characterization revealed that all four characterized conditions contained a cold worked microstructure to different degrees. The Y-12 foils exhibited a higher degree of cold working compared to the LANL foils, as evidenced by the highly elongated and obscure U-Mo grain structure present in each foil. The longitudinal orientations for both of the Y-12 foils possesses a highly laminar appearance with such a distorted grain structure that it was very difficult to even offer a range of grain sizes. The U-Mo grain structure of the LANL foils, by comparison, consisted of a more easily discernible grain structure with a mix of equiaxed and elongated grains. Both materials have an inhomogenous grain structure in that all of the characterized foils possess abnormally coarse grains.

  15. 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)

  16. Study on HANARO core conversion using U-Mo fuel

    International Nuclear Information System (INIS)

    Lee, K.H.; Lee, C.S.; Seo, C.G.; Park, S.J.; Kim, H.; Kim, C.K.

    2002-01-01

    Two types of fuel rods with different fuel meat diameter and uranium density are considered for HANARO core conversion with high density U-Mo fuel. Arranging standard fuels of 5.0 g U/cc and 6.35 mm in diameter at the inner ring of an assembly and reduced fuels of 4.3 g U/cc and 5.49 mm in diameter at the outer ring of an assembly flattens the assembly power distribution and avoids the increase of linear heat generation rate due to using higher uranium density and less number of fuel rods. The maximum linear heat generation rate is similar with the current reference core and four fuel sites at the outer core in the reflector tank is converted to the irradiation sites to suit more demand on fuel tests and radioisotope production at outer core sites. This new core has 32% longer fuel cycle than the current reference core. (author)

  17. Interdiffusion between U-Mo alloys and Al

    International Nuclear Information System (INIS)

    Mirandou, M.I.; Balart, S.N.; Ortiz, M.; Granovsky, M.S.; Hofman, G.L.

    2002-01-01

    During the fabrication and/or irradiation of the dispersion fuel elements, the fuel particles react with the surrounding Al matrix. This reaction results in the formation of a zone consisting of intermetallic compounds. The low thermal conductivity of these compounds has a major effect on the fuel temperature as well as on the swelling of the fuel. Interdiffusion between U-Mo/Al is being investigated using chemical diffusion couples. In this paper the first results obtained with optical and scanning electron microscope, electron microprobe and X-Ray diffraction are presented. Investigation of the effect on the formation of the interdiffusion zone of small additions of Mg to Al is the primary purpose of this study. (author)

  18. Update on Fresh Fuel Characterization of U-Mo Alloys

    International Nuclear Information System (INIS)

    Burkes, D.E.; Wachs, D.M.; Keiser, D.D.; Okuniewski, M.A.; Jue, J.F.; Rice, F.J.; Prabhakaran, R.

    2009-01-01

    The need to provide more accurate property information on U-Mo fuel alloys to operators, modellers, researchers, fabricators, and government increases as success of the GTRI Reactor Convert program continues. This presentation provides an update on fresh fuel characterization activities that have occurred at the INL since the RERTR 2008 conference in Washington, D.C. The update is particularly focused on properties recently obtained and on the development progress of new measurement techniques. Furthermore, areas where useful and necessary information is still lacking is discussed. The update deals with mechanical, physical, and microstructural properties for both integrated and separate effects. Appropriate discussion of fabrication characteristics, impurities, thermodynamic response, and effects on the topic areas are provided, along with a background on the characterization techniques used and developed to obtain the information. Efforts to measure similar characteristics on irradiated fuel plates are discussed.

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

    International Nuclear Information System (INIS)

    Kim, Y. S.; Hofman, G.; Rest, J.; Shevlyakov, G. V.

    2008-01-01

    This report can be divided into two parts: the first part, which is composed of sections 1, 2, and 3, is devoted to report the analyses of fission gas bubbles; the second part, which is in section 4, is allocated to describe the mechanistic model development. Swelling data of irradiated U-Mo alloy typically show that the kinetics of fission gas bubbles is composed of two different rates: lower initially and higher later. The transition corresponds to a burnup of ∼0 at% U-235 (LEU) or a fission density of ∼3 x 10 21 fissions/cm 3 . Scanning electron microscopy (SEM) shows that gas bubbles appear only on the grain boundaries in the pretransition regime. At intermediate burnup where the transition begins, gas bubbles are observed to spread into the intragranular regions. At high burnup, they are uniformly distributed throughout fuel. In highly irradiated U-Mo alloy fuel large-scale gas bubbles form on some fuel particle peripheries. In some cases, these bubbles appear to be interconnected and occupy the interface region between fuel and the aluminum matrix for dispersion fuel, and fuel and cladding for monolithic fuel, respectively. This is a potential performance limit for U-Mo alloy fuel. Microscopic characterization of the evolution of fission gas bubbles is necessary to understand the underlying phenomena of the macroscopic behavior of fission gas swelling that can lead to a counter measure to potential performance limit. The microscopic characterization data, particularly in the pre-transition regime, can also be used in developing a mechanistic model that predicts fission gas bubble behavior as a function of burnup and helps identify critical physical properties for the future tests. Analyses of grain and grain boundary morphology were performed. Optical micrographs and scanning electron micrographs of irradiated fuel from RERTR-1, 2, 3 and 5 tests were used. Micrographic comparisons between as-fabricated and as-irradiated fuel revealed that the site of

  20. 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

  1. Parametric study of fission-induced U-Mo fuel creep and structural analysis of fuel plates in view of implications for microstructure evolution

    International Nuclear Information System (INIS)

    Kim, Y.S.; Hofman, G.L.; Choo, Y.S.; Robinson, A.B.

    2010-01-01

    U-Mo fuel deformation during irradiation in U-Mo/Al dispersion plates is investigated by using the irradiation data from the RERTR-3 through -9 tests. The observation of fuel particle sintering during irradiation is also presented and its influence for fuel performance is discussed. Structural analysis was also performed to examine the relationship between the stress distribution in the plate and the location of matrix-pore formation in the plate. (author)

  2. Performance of Nb protective diffusion coating on U-Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ji-Hyeon; Sohn, Dong-Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, Sunghwan; Nam, Ji Min; Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    To achieve this aim, it is necessary to increase the volume fraction of fuel particles inside the meat. However, the technical limit is reached at approximately 55 vol.% of fuel particles in the aluminum matrix. As a solution, an uranium compound with an higher uranium density than existing U3Si2 fuel has to be selected. Also alloying the uranium must stabilize γ-phase of uranium at room temperature because adequate properties of the γ -phase of uranium showed a good irradiation behavior in the past. Hence, U-Mo alloys were selected as the best candidates. The formation of interaction phase is a critical problem to apply U-Mo alloys to the high performance research reactor. Different means have been proposed to reduce the interaction between U-Mo fuel and Al matrix. There are three means. : 1. Addition of a diffusion limiting element to the matrix 2. Insertion of a diffusion barrier at the interface between the U-Mo and the Al 3. Alloying of the U-Mo with a third element Here we present the effect of Nb coating as diffusion barrier on formation of interaction layers between UMo powders and Al matrix. We present the effect of Nb coating on formation of interaction layers between U-Mo powders and Al matrix. Centrifugally atomized U-7 wt.% Mo powders were used, and Nb was coated on the surface of U-7 wt.% Mo by sputtering. Subsequently, the Nb-coated U-7 wt.% Mo powders were mixed with pure Al powders, and were made into compacts. The compacts were annealed at 550 .deg. C for 1, 3, 5 hours, respectively, and the result showed that the Nb coating on U-7 wt.% Mo effectively suppressed the growth of interaction layers between U-7 wt.% Mo and Al matrix.

  3. Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

    Energy Technology Data Exchange (ETDEWEB)

    Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.; Holdaway, K. K.; Housley, G. K.; Rabin, B. H.

    2016-10-01

    The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, and other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.

  4. Irradiation testing of LEU fuels in the SILOE Reactor - Progress report

    International Nuclear Information System (INIS)

    Merchie, Francis; Baas, Claude; Martel, Patrick

    1985-01-01

    Irradiation of uranium-silicide fuels has continued in the SILOE reactor during the past year. Thickness vs. fission density data from four U 3 Si plates containing 5.5 and 6.0 g U/cm 3 have been analyzed, and the results are presented. The irradiation of a full 60 g U/cm 3 U 3 Si element has begun. In addition, four U 3 Si 2 plates containing 20 to 54 g U/cm 3 are now being irradiated. These irradiations and future plans are discussed in the paper. (author)

  5. Mechanical Calculations on U-Mo Dispersion fuel plates with MAIA

    International Nuclear Information System (INIS)

    Marelle, V.; Huet, F.; Lemoine, P.

    2005-01-01

    CEA has developed a 2D thermo-mechanical code, called MAIA, for modelling the behaviour of U-Mo dispersion fuel. MAIA uses a finite element method for the resolution of the thermal and mechanical problems. Physical models, issued of the DOE-ANL code PLATE, evaluate the fission products swelling and the volume fraction of the interaction between U-Mo and Al. They allow establishing strains in the meat imposed as loading for the mechanical calculation. MAIA has been validated on the irradiations IRIS 1 and RERTR-3 and a rather good agreement is obtained with post irradiation examinations. MAIA is used to calculate the last irradiation of the French UMo group, IRIS 2. MAIA predicts a maximum temperature of 112 deg. C and meat swelling of 16%. Mechanical calculations are finally performed to evaluate the sensitivity to some mechanical hypotheses such as constitutive laws and the way the meat swelling is applied. (author)

  6. Modeling a failure criterion for U-Mo/Al dispersion fuel

    Science.gov (United States)

    Oh, Jae-Yong; Kim, Yeon Soo; Tahk, Young-Wook; Kim, Hyun-Jung; Kong, Eui-Hyun; Yim, Jeong-Sik

    2016-05-01

    The breakaway swelling in U-Mo/Al dispersion fuel is known to be caused by large pore formation enhanced by interaction layer (IL) growth between fuel particles and Al matrix. In this study, a critical IL thickness was defined as a criterion for the formation of a large pore in U-Mo/Al dispersion fuel. Specifically, the critical IL thickness is given when two neighboring fuel particles come into contact with each other in the developed IL. The model was verified using the irradiation data from the RERTR tests and KOMO-4 test. The model application to full-sized sample irradiations such as IRISs, FUTURE, E-FUTURE, and AFIP-1 tests resulted in conservative predictions. The parametric study revealed that the fuel particle size and the homogeneity of the fuel particle distribution are influential for fuel performance.

  7. Modelling of U-Mo/Al Dispersion fuel fission induced swelling and creep

    International Nuclear Information System (INIS)

    Jeong, Gwan Yoon; Sohn, Dong Seong; Kim, Yeon Soo

    2014-01-01

    In a Dispersion fuel which U-Mo particles are dispersed in Al metal matrix, a similar phenomenon forming a bulge region was observed but it is difficult to quantify and construct a model for explaining creep and swelling because of its complex microstructure change during irradiation including interaction layer (IL) and porosity formation. In a Dispersion fuel meat, fission product induces fuel particles swelling and it has to be accommodated by the deformation of the Al matrix and newly formed IL during irradiation. Then, it is reasonable that stress from fuel swelling in the complex structure should be relaxed by local adjustments of particles, Al matrix, and IL. For analysis of U-Mo/Al Dispersion fuel creep, the creep of U-Mo particle, Al matrix, and IL should be considered. Moreover, not only fuel particle swelling and IL growth, but also fuel and Al matrix consumptions due to IL formation are accounted in terms of their volume fraction changes during irradiation. In this work, fuel particles, Al matrix and IL are treated in a way of homogenized constituents: Fuel particles, Al matrix and IL consist of an equivalent meat during irradiation. Meat volume swelling of two representative plates was measured: One (Plate A) was a pure Al matrix with 6g/cc uranium loading, the other (Plate B) a silicon added Al matrix with 8g/cc uranium loading. The meat swelling of calculated as a function of burnup. The meat swelling of calculation and measurement was compared and the creep rate coefficients for Al and IL were estimated by repetitions. Based on assumption that only the continuous phase of Al-IL combined matrix accommodated the stress from fuel particle swelling and it was allowed to have creep deformation, the homogenization modeling was performed. The meat swelling of two U-Mo/Al Dispersion fuel plates was modeled by using homogenization model

  8. Modelling of U-Mo/Al Dispersion fuel fission induced swelling and creep

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Sohn, Dong Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, Argonne (United States)

    2014-05-15

    In a Dispersion fuel which U-Mo particles are dispersed in Al metal matrix, a similar phenomenon forming a bulge region was observed but it is difficult to quantify and construct a model for explaining creep and swelling because of its complex microstructure change during irradiation including interaction layer (IL) and porosity formation. In a Dispersion fuel meat, fission product induces fuel particles swelling and it has to be accommodated by the deformation of the Al matrix and newly formed IL during irradiation. Then, it is reasonable that stress from fuel swelling in the complex structure should be relaxed by local adjustments of particles, Al matrix, and IL. For analysis of U-Mo/Al Dispersion fuel creep, the creep of U-Mo particle, Al matrix, and IL should be considered. Moreover, not only fuel particle swelling and IL growth, but also fuel and Al matrix consumptions due to IL formation are accounted in terms of their volume fraction changes during irradiation. In this work, fuel particles, Al matrix and IL are treated in a way of homogenized constituents: Fuel particles, Al matrix and IL consist of an equivalent meat during irradiation. Meat volume swelling of two representative plates was measured: One (Plate A) was a pure Al matrix with 6g/cc uranium loading, the other (Plate B) a silicon added Al matrix with 8g/cc uranium loading. The meat swelling of calculated as a function of burnup. The meat swelling of calculation and measurement was compared and the creep rate coefficients for Al and IL were estimated by repetitions. Based on assumption that only the continuous phase of Al-IL combined matrix accommodated the stress from fuel particle swelling and it was allowed to have creep deformation, the homogenization modeling was performed. The meat swelling of two U-Mo/Al Dispersion fuel plates was modeled by using homogenization model.

  9. Interdiffusion among U-Mo-Zr and alloys of Al to 550oC

    International Nuclear Information System (INIS)

    Komar Varela, C.L; Arico, S.F; Gribaudo, L.M

    2006-01-01

    The international community, by means of the project 'Reduced Enrichment for Research and Test Reactors' is interested in the development of a new nuclear fuel of very high density of uranium and low enrichment (≤ 20% de U 235 ) for reactors of investigation and production of radioisotopes, that permit to reach greater neutron flows, with good capacity to be reprocessed One of these assemblies are the alloys of U with Mo contents between 7 and 10% in weight. In the fuels 'dispersed type plate' the particles of U-Mo are mixed with dust of aluminum and are co - laminated between two plates of an alloy of the same material. The existing contact among the particles permits the interdiffusion of the materials with the consequent apparition of new phases. Studies pursuit-irradiation have shown a badly behavior of these new phases. It is for this that is necessary to control the presence of these products of interaction. The aggregate of a third element to the alloys U - Mo has begun to be practiced with this purpose. In this work the modification of the start of the disorder of the phase γU in the alloy U-7%Mo-1%Zr was studied and the interdiffusion between pure aluminum and the same alloy to 550 o C. The results obtained are compared with other obtained for peers U-Mo/Al. The techniques of characterization utilized were: optical microscopy, analysis by diffraction of X-rays and microanalysis quantitative by microprobe electronic. It was observed that the aggregate of Zr refines the grain for a processing of homogenized in composition of Mo to 1000 o C and accelerates the start of the disorder of the phase γU to 550 o C. As for the zone of interaction, was found that the composed identifying do not they differ to them reported in the in peers U-Mo/Al. These are: (U,Mo)Al 4 y UAl 3 (AG)

  10. First-principles study of the surface properties of U-Mo system

    Energy Technology Data Exchange (ETDEWEB)

    Mei, Zhi-Gang; Liang, Linyun; Yacout, Abdellatif M.

    2018-02-01

    U-Mo alloys are promising fuels for future high-performance research reactors with low enriched uranium. Surface properties, such as surface energy, are important inputs for mesoscale simulations (e.g., phase field method) of fission gas bubble behaviors in irradiated nuclear fuels. The lack of surface energies of U-Mo alloys prevents an accurate modeling of the morphology of gas bubbles and gas bubble-induced fuel swelling. To this end, we study the surface properties of U-Mo system, including bcc Mo, alpha-U, gamma-U, and gamma U-Mo alloys. All surfaces up to a maximum Miller index of three and two are calculated for cubic Mo and gamma-U and non-cubic alpha-U, respectively. The equilibrium crystal shapes of bcc Mo, alpha-U and gamma-U are constructed using the calculated surface energies. The dominant surface orientations and the area fraction of each facet are determined from the constructed equilibrium crystal shape. The disordered gamma U-Mo alloys are simulated using the Special Quasirandom Structure method. The (1 1 0) and (1 0 0) surface energies of gamma U-7Mo and U-10Mo alloys are predicted to lie between those of gamma-U and bcc Mo, following a linear combination of the two constituents' surface energies. To better compare with future measurements of surface energies, the area fraction weighted surface energies of alpha-U, gamma-U and gamma U-7Mo and U-10Mo alloys are also predicted. (C) 2017 Published by Elsevier B.V.

  11. Foreign research reactor irradiated nuclear fuel inventories containing HEU and LEU of United States origin

    International Nuclear Information System (INIS)

    Matos, J.E.

    1994-12-01

    This report provides estimates of foreign research reactor inventories of aluminum-based and TRIGA irradiated nuclear fuel elements containing highly enriched and low enriched uranium of United States origin that are anticipated in January 1996, January 2001, and January 2006. These fuels from 104 research reactors in 41 countries are the same aluminum-based and TRIGA fuels that were eligible for receipt under the Department of Energy's Offsite Fuels Policy that was in effect in 1988. All fuel inventory and reactor data that were available as of December 1, 1994, have been included in the estimates of approximately 14,300 irradiated fuel elements in January 1996, 18,800 in January 2001, and 22,700 in January 2006

  12. Rupture of Al matrix in U-Mo/Al dispersion fuel by fission induced creep

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Sohn, Dong Seong [UNIST, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, Argonnge (United States); Lee, Kyu Hong [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This phenomenon was found specifically in the dispersion fuel plate with Si addition in the Al matrix to suppress interaction layer (IL) formation between UMo and Al. It is known that the stresses induced by fission induced swelling in U-Mo fuel particles are relieved by creep deformation of the IL, surrounding the fuel particles, that has a much higher creep rate than the Al matrix. Thus, when IL growth is suppressed, the stress is instead exerted on the Al matrix. The observed rupture in the Al matrix is believed to be caused when the stress exceeded the rupture strength of the Al matrix. In this study, the possibility of creep rupture of the Al matrix between the neighboring U-Mo fuel particles was examined using the ABAQUS finite element analysis (FEA) tool. The predicted rupture time for a plate was much shorter than its irradiation life indicating a rupture during the irradiation. The higher stress leads Al matrix to early creep rupture in this plate for which the Al matrix with lower creep strain rate does not effectively relieve the stress caused by the swelling of the U-Mo fuel particles. For the other plate, no rupture was predicted for the given irradiation condition. The effect of creeping of the continuous phase on the state of stress is significant.

  13. Interdiffusion between U(Mo,Pt) or U(Mo,Zr) and Al or Al A356 alloy

    International Nuclear Information System (INIS)

    Komar Varela, C.; Mirandou, M.; Arico, S.; Balart, S.; Gribaudo, L.

    2009-01-01

    Solid state reactions in chemical diffusion couples U-7 wt.%Mo-0.9 wt.%Pt/Al at 580 deg. C and U-7 wt.%Mo-0.9 wt.%Pt/Al A356 alloy, U-7 wt.%Mo-1 wt.%Zr/Al and U-7 wt.%Mo-1 wt.%Zr/Al A356 alloy at 550 deg. C were characterized. Results were obtained from optical and scanning electron microscopy, electron probe microanalysis and X-ray diffraction. The UAl 3, UAl 4 and Al 20 Mo 2 U phases were identified in the interaction layers of γU(Mo,Pt)/Al and γU(Mo,Zr)/Al diffusion couples. Al 43 Mo 4 U 6 ternary compound was also identified in γU(Mo,Zr)/Al due to the decomposition of γU(Mo,Zr) phase. The U(Al,Si) 3 and U 3 Si 5 phases were identified in the interaction layers of γU(Mo,Pt)/Al A356 and γU(Mo,Zr)/Al A356 diffusion couples. These phases are formed due to the migration of Si to the interaction layer. In the diffusion couple U(Mo,Zr)/Al A356, Zr 5 Al 3 phase was also identified in the interaction layer. The use of synchrotron radiation at Brazilian Synchrotron Light Laboratory (LNLS, CNPq, Campinas, Brazil) was necessary to achieve a complete crystallographic characterization.

  14. Origin and development of the new U-Mo nuclear fuel

    International Nuclear Information System (INIS)

    Boyard, M.; Languille, A.; Thomasson, J.; Hamy, J.M.

    2002-01-01

    Historically most research reactors have used highly enriched nuclear fuels (enrichment > 90 %). Since 1977 the non-proliferation policy has imposed to convert these reactors to far less enriched fuels (< 20 %). An international consensus has evolved towards a nuclear fuel with an enrichment factor of 19,75 %, this fuel is made of a powdered U-Mo alloy scattered in an aluminium die. The external dimensions and the cladding materials of the fuel plate are unchanged in order to minimize development and qualification costs. The U-Mo fuel is expected to maintain or even to increase the performance of reactors and to allow the processing of spent fuels in the same installations as those used for fuels issuing from power plants. Cea, Cogema, Cerca, Framatome, and Technicatome have shared their technical means, their know-how and their financial resources to develop this new nuclear fuel. 2006 is the contract date by which American authorities will stop repatriating the ancient spent fuel (uranium silicide) from research reactors so it is imperative to make available by this date a new nuclear fuel with a satisfactory end of cycle. This article also presents the French program of qualification of the U-Mo fuel. 2 series of irradiation have already been performed, one (Isis-1) in Osiris reactor (Saclay, France) and the second (Umus) in HFR (Petten, Netherlands). A clad failure has led to stop the Umus experiment. 2 new series of irradiation are scheduled to start in 2002. In a parallel way, in the framework of the design of the RJH (Jules Horowitz reactor) Cea will soon perform irradiation of U-Mo fuel plates in BR2 (Mol, Belgium). (A.C.)

  15. Improved performance of U-Mo dispersion fuel by Si addition in Al matrix.

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y S; Hofman, G L [Nuclear Engineering Division

    2011-06-01

    The purpose of this report is to collect in one publication and fit together work fragments presented in many conferences in the multi-year time span starting 2002 to the present dealing with the problem of large pore formation in U-Mo/Al dispersion fuel plates first observed in 2002. Hence, this report summarizes the excerpts from papers and reports on how we interpreted the relevant results from out-of-pile and in-pile tests and how this problem was dealt with. This report also provides a refined view to explain in detail and in a quantitative manner the underlying mechanism of the role of silicon in improving the irradiation performance of U-Mo/Al.

  16. Influence of Fuel-Matrix Interaction on the Deformation of U-Mo Dispersion Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, Chicago (United States)

    2014-05-15

    In order to predict the fuel plate failure leading to breakaway swelling in the meat, an understanding of the effects of the fuel-matrix interaction behavior on the deformation of fuel meat is necessary. However, the effects of IL formation on the development of breakaway swelling have not been studied thoroughly. A mechanism that explains large pore growth that leads to breakaway swelling has not been included in the existing fuel performance models. In this study, the effect of the fuel-matrix interaction on large interfacial porosity development at the IL-Al interface is analyzed using both mechanistic correlations and observations from the post-irradiation examination results of U-Mo Dispersion fuels. The effects of fuel-matrix interaction on the fuel performance of U-Mo/Al Dispersion fuel were investigated. Fuel-matrix interaction bears the causes for breakaway swelling that can lead to a fuel failure under a high-power irradiation condition. Fission gas atoms are released from U-Mo particles to the interaction layer via diffusion and recoil. The fission gases released from the U-Mo and produced in the ILs are further released to the IL-Al interface by diffusion in the IL and recoil. Large pore formation at the IL-Al interface is attributed to the active diffusion of fission gas atoms in the ILs and coalescence between the small bubbles there. A model calculation showed that IL growth increases the probability of forming a breakaway swelling condition. ILs are connected to each other and the Al matrix decreases as ILs grow. When more ILs are interconnected, breakaway swelling can occur when the effective stress from the fission gas pressure in the IL-Al interfacial pore becomes larger than the yield strength of the Al matrix.

  17. MODELING OF INTERACTION LAYER GROWTH BETWEEN U-Mo PARTICLES AND AN Al MATRIX

    OpenAIRE

    YEON SOO KIM; G.L. HOFMAN; HO JIN RYU; JONG MAN PARK; A.B. ROBINSON; D.M. WACHS

    2013-01-01

    Interaction layer growth between U-Mo alloy fuel particles and Al in a dispersion fuel is a concern due to the volume expansion and other unfavorable irradiation behavior of the interaction product. To reduce interaction layer (IL) growth, a small amount of Si is added to the Al. As a result, IL growth is affected by the Si content in the Al matrix. In order to predict IL growth during fabrication and irradiation, empirical models were developed. For IL growth prediction during fabrication an...

  18. CNEA developments in U-Mo-ZrY-4 mini plates and plates fabrication process

    International Nuclear Information System (INIS)

    López, M.; Picchetti, B.; Gonzalez, A.; Taboada, H.

    2013-01-01

    The Uranium Molybdenum alloy was the material chosen to develop the fabrication of high density nuclear fuel, due to its excellent behaviour under irradiation –a consequence of the metastable bcc crystalline structure-. At present, the study is focused on the application of this alloy to monolithic fuel plate development, which fuel core is a thin U-Mo layer. The Zircalloy-4 (Zry-4) alloy used as cladding material is extensively known in the nuclear industry due to its low neutron capture section efficiency and excellent mechanical and corrosion resistance properties. Miniplates fabrication process involves a welded compact made of two Zry-4 covers and a frame surrounding a monolithic U-Mo core, which is co rolled under high temperature. Molybdenum contains of 7% to 10% (mass) in U Mo alloys guarantees the presence of meta-stable bcc gamma phase and, at the same time, does not penalize the neutron economy due to Mo98 presence. In the case of U Mo monolithic miniplates relevant parameters of fabrication, considering the behaviour of the U-Mo alloys reported in many work and in order to optimize the o-rolling process, have been revised: co-rolling temperature, compressive stress and presence of gap. Under this experimental conditions can be studied the the interdiffusion layer, the binding between materials and the Dog Bone. The experimental results shows that 650ºC is an optimal co-rolling temperature; at higher temperatures not only a bigger interdiffusion layer is observed –this phenomenon can lead to a region enriched in Molybdenum- but also a bigger Dog Bone is obtained. Working at higher compressive stress has the same effect in relation to the interdiffusion layer. In addition, the absence of gases in the core is essential for the correct binding of the materials. Concerning the monolithic U-Mo plates fabrication, involved in the ALT FUTURE experiment a new workshop has been conditioned. The aim is to use all the valuable information collected during

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  20. Coated U(Mo) Fuel: As-Fabricated Microstructures

    Energy Technology Data Exchange (ETDEWEB)

    Emmanuel Perez; Dennis D. Keiser, Jr.; Ann Leenaers; Sven Van den Berghe; Tom Wiencek

    2014-04-01

    As part of the development of low-enriched uranium fuels, fuel plates have recently been tested in the BR-2 reactor as part of the SELENIUM experiment. These fuel plates contained fuel particles with either Si or ZrN thin film coating (up to 1 µm thickness) around the U-7Mo fuel particles. In order to best understand irradiation performance, it is important to determine the starting microstructure that can be observed in as-fabricated fuel plates. To this end, detailed microstructural characterization was performed on ZrN and Si-coated U-7Mo powder in samples taken from AA6061-clad fuel plates fabricated at 500°C. Of interest was the condition of the thin film coatings after fabrication at a relatively high temperature. Both scanning electron microscopy and transmission electron microscopy were employed. The ZrN thin film coating was observed to consist of columns comprised of very fine ZrN grains. Relatively large amounts of porosity could be found in some areas of the thin film, along with an enrichment of oxygen around each of the the ZrN columns. In the case of the pure Si thin film coating sample, a (U,Mo,Al,Si) interaction layer was observed around the U-7Mo particles. Apparently, the Si reacted with the U-7Mo and Al matrix during fuel plate fabrication at 500°C to form this layer. The microstructure of the formed layer is very similar to those that form in U-7Mo versus Al-Si alloy diffusion couples annealed at higher temperatures and as-fabricated U-7Mo dispersion fuel plates with Al-Si alloy matrix fabricated at 500°C.

  1. Major results on the development of high density U-Mo fuel and pin-type fuel elements executed under the Russian RERTR program and in cooperation with ANL (USA)

    International Nuclear Information System (INIS)

    Vatulin, A.; Morozov, A.; Stetsky, Y.; Suprun, V.; Dobrikova, I.; Trifonov, Y.; Mishunin, V.; Sorokin, V.

    2003-01-01

    VNIINM is active participant of 'Russian program on Reduced Enrichment for Research and Test Reactors'. Institute Works in two main directions: 1) development of new high-density fuels (HDF) and 2) development of new design of fuel elements with LEU. The development of the new type fuel element is carried out both for existing reactors, and for developing new advanced reactors. The 'TVEL' concern is coordinator of works of this program. The majority enterprises of branch (NIIAR, PIYaF, RRC KI, NZChK) take part in this work. Since 2000 these works are being conducted in cooperation with Argonne National Laboratory (USA) within the RERTR program under VNIINM with ANL contract. At the present, a large set of pre-pile investigations has been completed. All necessary fabrication procedures have been developed for utilization of U-Mo dispersion fuel in Russian-designed research reactors. For irradiation tests the pin-type mini-fuel elements with HDF dispersion fuel with LEU and the uranium density equaled to 4,0 and 6,0 g/cm 3 (up to 40 vol.%) have been manufactured. Their irradiation began in August 2003 in the MIR reactor (NIIAR, Dimitrovgrad). A large set of works for preparation of lifetime tests (WWR-M reactor in Gatchina) of two full-scale fuel assemblies with new pin-type fuel elements on basis LEU UO 2 -Al and UMo-Al fuels has been completed. The in-pile tests of fuel assemblies began in September 2003. The summary of important results of performed works and their near-term future are presented in paper. (author)

  2. Report on the Synchrotron Characterization of U-Mo and U-Zr Alloys and the Modeling Results

    Energy Technology Data Exchange (ETDEWEB)

    Okuniewski, Maria A. [Purdue Univ., West Lafayette, IN (United States); Ganapathy, Varsha [Purdue Univ., West Lafayette, IN (United States); Hamilton, Brenden [Purdue Univ., West Lafayette, IN (United States); Cassutt, Paul [Purdue Univ., West Lafayette, IN (United States); Zhang, Fan [Purdue Univ., West Lafayette, IN (United States); Velaquez, Daniel [Illinois Inst. of Technology, Chicago, IL (United States); Seibert, Rachel [Illinois Inst. of Technology, Chicago, IL (United States); Terry, Jeff [Illinois Inst. of Technology, Chicago, IL (United States); Sprouster, David [Brookhaven National Lab. (BNL), Upton, NY (United States); Ecker, Lynne [Brookhaven National Lab. (BNL), Upton, NY (United States); Elbakhshwan, Mohamed [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-09-01

    ABSTRACT Uranium-molybdenum (U-Mo) and uranium-zirconium (U-Zr) are two promising fuel candidates for nuclear transmutation reactors which burn long-lived minor actinides and fission products within fast spectrum reactors. The objectives of this research are centered on understanding the early stages of fuel performance through the examination of the irradiation induced microstructural changes in U-Zr and U-Mo alloys subjected to low neutron fluences. Specimens that were analyzed include those that were previously irradiated in the Advanced Test Reactor at INL. This most recent work has focused on a sub-set of the irradiated specimens, specifically U-Zr and U-Mo alloys that were irradiated to 0.01 dpa at temperatures ranging from (150-800oC). These specimens were analyzed with two types of synchrotron techniques, including X-ray absorption fine structure and X-ray diffraction. These techniques provide non-destructive microstructural analysis, including phase identification and quantitation, lattice parameters, crystallite sizes, as well as bonding, structure, and chemistry. Preliminary research has shown changes in the phase fractions, crystallite sizes, and lattice parameters as a function of irradiation and temperature. Future data analyses will continue to explore these microstructural changes.

  3. Effect of stress evolution on microstructural behavior in U-Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, G.Y. [Department of Nuclear Engineering, Ulsan National Institute of Science and Technology, 50 UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan 689-798 (Korea, Republic of); Kim, Yeon Soo; Jamison, L.M. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Lee, K.H. [Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Sohn, Dong-Seong, E-mail: dssohn@unist.ac.kr [Department of Nuclear Engineering, Ulsan National Institute of Science and Technology, 50 UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan 689-798 (Korea, Republic of)

    2017-04-15

    U-Mo/Al dispersion fuel irradiated to high burnup at high power (high fission rate) exhibited microstructural changes including deformation of the fuel particles, pore growth, and rupture of the Al matrix. The driving force for these microstructural changes was meat swelling resulting from a combination of fuel particle swelling and interaction layer (IL) growth. In some cases, pore growth in the interaction layers also contributed to meat swelling. The main objective of this work was to determine the stress distribution within the fuel meat that caused these phenomena. A mechanical equilibrium between the stress generated by fuel meat swelling and the stress relieved by fission-induced creep in the meat constituents (U-Mo particles, Al matrix, and IL) was considered. Test plates with well-recorded fabrication data and irradiation conditions were used, and their post-irradiation examination (PIE) data was obtained. ABAQUS finite element analysis (FEA) was utilized to simulate the microstructural evolution of the plates. The simulation results allowed for the determination of effective stress and hydrostatic stress exerted on the meat constituents. The effects of fabrication and irradiation parameters on the stress distribution that drives microstructural evolutions, such as pore growth in the IL and Al matrix rupture, were investigated. - Highlights: •Post-irradiation data for irradiated miniplates were analyzed by using their optical microscopy images. •ABAQUS finite element analysis (FEA) package was utilized to simulate the microstructural evolution of the selected plates. •Stresses were assessed to analyze their effects on microstructural changes during irradiation.

  4. High temperature interdiffusion and phase equilibria in U-Mo

    International Nuclear Information System (INIS)

    Lundberg, L.B.

    1988-01-01

    Experimental data for interdiffusion and phase equilibria in the U-Mo system have been obtained over the temperature range 1400 to 1525 K as a fallout from compatibility experiments in which UO 2 was decomposed by lithium in closed molybdenum capsules. Composition-position, x-ray diffraction and microstructural data from the interdiffusion zones indicate that the intermediate phase U 2 Mo is found in this temperature range, contrary to the currently accepted equilibrium U-Mo phase diagram. The U-Mo interdiffusion data are in good agreement with published values. Inclusion of the U 2 Mo phase in a theoretical correlation of interdiffusion and phase equilibria data using Darken's equation indicate that high temperature interdiffusion of uranium and molybdenum follows the usual thermodynamic rules. Significant changes in the value of the thermodynamic based Darken factor near the U 2 Mo phase boundary on the high uranium side are indicated from both the new and published interdiffusion data. 9 refs., 10 figs., 3 tabs

  5. Corrosion report for the U-Mo fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Bennett, Wendy D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Doherty, Ann L. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Fuller, E. S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hardy, John S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Omberg, Ronald P. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

    2014-08-28

    The Fuel Cycle Research and Development (FCRD) program of the Office of Nuclear Energy (NE) has implemented a program to develop a Uranium-Molybdenum (U-Mo) metal fuel for Light Water Reactors (LWR)s. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties, which includes high thermal conductivity for less stored heat energy. With sufficient development, it may be able to provide the Light Water industry with a melt-resistant accident tolerant fuel with improved safety response. However, the corrosion of this fuel in reactor water environments needs to be further explored and optimized by additional alloying. The Pacific Northwest National Laboratory has been tasked with performing ex-reactor corrosion testing to characterize the performance of U-Mo fuel. This report documents the results of the effort to characterize and develop the U-Mo metal fuel concept for LWRs with regard to corrosion testing. The results of a simple screening test in buffered water at 30°C using surface alloyed U-10Mo is documented and discussed. The screening test was used to guide the selection of several potential alloy improvements that were found and are recommended for further testing in autoclaves to simulate PWR water conditions more closely.

  6. LEU fuel development at CERCA

    International Nuclear Information System (INIS)

    Durand, Jean Pierre; Ottone, J.C.; Mahe, M.; Ferraz, G.

    1998-01-01

    The aim of this paper is to detail the recent progress on both U 3 Si 2 high loaded fuels and new γ phase fuels. Concerning high density density silicide plates up to 6 g Ut/cm 3 , the CEA irradiation programme is completed. Data are still under analysis but one can state that the behaviour was globally similar to conventional fuels known in SILOE and OSIRIS reactors. From the new γ fuel point of view, after demonstration feasibility in 1997 of U Mo thermally stable plates loaded up to 8.3 g Ut/cm3, CERCA has analysed the technical ability of quality inspection means assuming that is of an utmost interest for the insurance of a proper use of high performances fuel in reactors. There are mainly two differences between U Mo fuels (and more generally γ fuels) and conventional ones. Firstly, X-ray diffraction analysis on the fuel powder are needed because the chemical analysis is not sufficient to characterise the γ structure requested. Secondly, the physical limits of the Ultrasonic inspection have been reached due to transitory effect between the meat and the edges. Therefore this technic can not applied in the transitory areas. From that knowledge, the manufacture specifications for a plate dedicated to an irradiation plan can be discussed with a clearer view of the main differences with the U 3 Si 2 fuel reference. (author)

  7. A modelling study of the inter-diffusion layer formation in U-Mo/Al dispersion fuel plates at high power

    Energy Technology Data Exchange (ETDEWEB)

    Ye, B.; Hofman, G. L.; Leenaers, A.; Bergeron, A.; Kuzminov, V.; Van den Berghe, S.; Kim, Y. S.; Wallin, H.

    2018-02-01

    Post irradiation examinations of full-size U-Mo/Al dispersion fuel plates fabricated with ZrN- or Sicoated U-Mo particles revealed that the reaction rate of irradiation-induced U-Mo-Al inter-diffusion, an important microstructural change impacting the performance of this type of fuel, is temperature and fission-rate dependent. In order to simulate the U-Mo/Al inter-diffusion layer (IL) growth behavior in full-size dispersion fuel plates, the existing IL growth correlation was modified with a temperaturedependent multiplication factor that transits around a threshold fission rate. In-pile irradiation data from four tests in the BR2 reactors, including FUTURE, E-FUTURE, SELEMIUM, and SELEMIUM-1a, were utilized to determine and validate the updated IL growth correlation. Irradiation behavior of the plates was simulated with the DART-2D computational code. The general agreement between the calculated and measured fuel meat swelling and constituent volume fractions as a function of fission density demonstrated the plausibility of the updated IL growth correlation. The simulation results also suggested the temperature dependence of the IL growth rate, similar to the temperature dependence of the intermixing rate in ion-irradiated bi-layer systems.

  8. Small-scale Specimen Testing of Monolithic U-Mo Fuel Foils

    Energy Technology Data Exchange (ETDEWEB)

    Ramprashad Prabhakaran; Douglas E. Burkes; James I. Cole; Indrajit Charit; Daniel M. Wachs

    2008-10-01

    The objective of this investigation is to develop a shear punch testing (SPT) procedure and standardize it to evaluate the mechanical properties of irradiated fuels in a hot-cell so that the tensile behavior can be predicted using small volumes of material and at greatly reduced irradiation costs. This is highly important in the development of low-enriched uranium fuels for nuclear research and test reactors. The load-displacement data obtained using SPT can be interpreted in terms of and correlated with uniaxial mechanical properties. In order to establish a correlation between SPT and tensile data, sub-size tensile and microhardness testing were performed on U-Mo alloys. In addition, efforts are ongoing to understand the effect of test parameters (such as specimen thickness, surface finish, punch-die clearance, crosshead velocity and carbon content) on the measured mechanical properties, in order to rationalize the technique, prior to employing it on a material of unknown strength.

  9. Small-scale Specimen Testing of Monolithic U-Mo Fuel Foils

    International Nuclear Information System (INIS)

    Ramprashad Prabhakaran; Douglas E. Burkes; James I. Cole; Indrajit Charit; Daniel M. Wachs

    2008-01-01

    The objective of this investigation is to develop a shear punch testing (SPT) procedure and standardize it to evaluate the mechanical properties of irradiated fuels in a hot-cell so that the tensile behavior can be predicted using small volumes of material and at greatly reduced irradiation costs. This is highly important in the development of low-enriched uranium fuels for nuclear research and test reactors. The load-displacement data obtained using SPT can be interpreted in terms of and correlated with uniaxial mechanical properties. In order to establish a correlation between SPT and tensile data, sub-size tensile and microhardness testing were performed on U-Mo alloys. In addition, efforts are ongoing to understand the effect of test parameters (such as specimen thickness, surface finish, punch-die clearance, crosshead velocity and carbon content) on the measured mechanical properties, in order to rationalize the technique, prior to employing it on a material of unknown strength

  10. Effects of Particle Size and Shape on U-Mo/Al Thermal Conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Tae-Won; Sohn, Dong-Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-10-15

    The thermal conductivity of atomized U-Mo/Al dispersion fuels was measured only by Lee et al. by laser-flash and differential scanning calorimetry (DSC) methods. For the U-Mo particles, they are deformed during manufacturing process such as hot rolling and during irradiation by the creep deformation. Fricke developed a model for the effective thermal conductivity of a dilute suspension of randomly oriented spheroidal particles. In general, the thermal conductivity of composite increase when the particle shape is not sphere. This model is also based on continuum theory which assumes both temperature and heat flux are continuous across the interface. Kapitza, however, showed that there is a discontinuity in temperature across the interface at metal/liquid helium interface. In general, the discontinuity is from the thermal resistance at the interface. If the thermal resistance has a significant impact on the thermal conductivity, particle size is one of the essential parameter for determining the effective thermal conductivity of composite materials. Every, et al modified Bruggeman model to consider the interfacial thermal resistance. The U-Mo/Al dispersion fuel thermal conductivity calculation can be improved by considering the anisotropic effects and interface thermal resistances. There have been various works to analyze the thermal conductivity through Finite Element Method (FEM). Coulson developed a realistic FEM model to calculate the effective thermal conductivity of the fuel meat. This FEM model does not consider the anisotropic effects and interface thermal resistances. Therefore, these effects can be evaluated by comparing the FEM calculated effective thermal conductivity with measured data. In this work, the FEM analysis was done and the anisotropic effects and interface thermal resistances was estimated. From this results, the particle shape and size effects will be discussed. Many thermal conductivity models for the particle dispersed composites have been

  11. Effects of Particle Size and Shape on U-Mo/Al Thermal Conductivity

    International Nuclear Information System (INIS)

    Cho, Tae-Won; Sohn, Dong-Seong

    2014-01-01

    The thermal conductivity of atomized U-Mo/Al dispersion fuels was measured only by Lee et al. by laser-flash and differential scanning calorimetry (DSC) methods. For the U-Mo particles, they are deformed during manufacturing process such as hot rolling and during irradiation by the creep deformation. Fricke developed a model for the effective thermal conductivity of a dilute suspension of randomly oriented spheroidal particles. In general, the thermal conductivity of composite increase when the particle shape is not sphere. This model is also based on continuum theory which assumes both temperature and heat flux are continuous across the interface. Kapitza, however, showed that there is a discontinuity in temperature across the interface at metal/liquid helium interface. In general, the discontinuity is from the thermal resistance at the interface. If the thermal resistance has a significant impact on the thermal conductivity, particle size is one of the essential parameter for determining the effective thermal conductivity of composite materials. Every, et al modified Bruggeman model to consider the interfacial thermal resistance. The U-Mo/Al dispersion fuel thermal conductivity calculation can be improved by considering the anisotropic effects and interface thermal resistances. There have been various works to analyze the thermal conductivity through Finite Element Method (FEM). Coulson developed a realistic FEM model to calculate the effective thermal conductivity of the fuel meat. This FEM model does not consider the anisotropic effects and interface thermal resistances. Therefore, these effects can be evaluated by comparing the FEM calculated effective thermal conductivity with measured data. In this work, the FEM analysis was done and the anisotropic effects and interface thermal resistances was estimated. From this results, the particle shape and size effects will be discussed. Many thermal conductivity models for the particle dispersed composites have been

  12. Neutronic and thermal-hydraulic analysis of a device for irradiation of LEU UAlx-Al targets for 99Mo production in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Nishiyama, Pedro Julio Batista de Oliveira

    2012-01-01

    Technetium-99m ( 99m Tc), the product of radioactive decay of molybdenum-99 ( Mo), is one of the most widely used radioisotope in nuclear medicine, covering approximately 80% of all radiodiagnosis procedures in the world. Nowadays, Brazil requires an amount of about 450 Ci of 99 Mo per week. Due to the crisis and the shortage of 99 Mo supply chain that has been observed on the world since 2008, IPEN/CNEN-SP decided to develop a project to produce 99 Mo through fission of uranium-235. The objective of this dissertation was the development of neutronic and thermal-hydraulic calculations to evaluate the operational safety of a device for 99 Mo production to be irradiated in the IEA-Rl reactor core at 5 MW. In this device will be placed ten targets of UAl x -Al dispersion fuel with low enriched uranium (LEU) and density of 2.889 gU/cm 3 . For the neutronic calculations were utilized the computer codes HAMMER-TECHNION and CITATION and the maximum temperatures reached in the targets were calculated with the code MTRCR-IEA-R1. The analysis demonstrated that the device irradiation will occur without adverse consequences to the operation of the reactor. The total amount of 99 Mo was calculated with the program SCALE and considering that the time needed for the chemical processing and recovering of the 99 Mo will be five days after the irradiation, we have that the 99 Mo activity available for distribution will be 176 Ci for 3 days of irradiation, 236 Ci for 5 days of irradiation and 272 Ci for 7 days of targets irradiation. (author)'

  13. Modeling of interaction layer growth between U-Mo particles and an Al matrix

    International Nuclear Information System (INIS)

    Kim, Yeon Soo; Horman, G. L.; Ryu, Ho Jin; Park, Jong Man; Robinson, A. B.; Wachs, D. M.

    2013-01-01

    Interaction layer growth between U-Mo alloy fuel particles and Al in a dispersion fuel is a concern due to the volume expansion and other unfavorable irradiation behavior of the interaction product. To reduce interaction layer (IL) growth, a small amount of Si is added to the Al. As a result, IL growth is affected by the Si content in the Al matrix. In order to predict IL growth during fabrication and irradiation, empirical models were developed. For IL growth prediction during fabrication and any follow-on heating process before irradiation, out-of-pile heating test data were used to develop kinetic correlations. Two out-of-pile correlations, one for the pure Al matrix and the other for the Al matrix with Si addition, respectively, were developed, which are Arrhenius equations that include temperature and time. For IL growth predictions during irradiation, the out-of-pile correlations were modified to include a fission-rate term to consider fission enhanced diffusion, and multiplication factors to incorporate the Si addition effect and the effect of the Mo content. The in-pile correlation is applicable for a pure Al matrix and an Al matrix with the Si content up to 8 wt%, for fuel temperatures up to 200 .deg. C, and for Mo content in the range of 6 - 10wt%. In order to cover these ranges, in-pile data were included in modeling from various tests, such as the US RERTR-4, -5, -6, -7 and -9 tests and Korea's KOMO-4 test, that were designed to systematically examine the effects of the fission rate, temperature, Si content in Al matrix, and Mo content in U-Mo particles. A model converting the IL thickness to the IL volume fraction in the meat was also developed

  14. MODELING OF INTERACTION LAYER GROWTH BETWEEN U-Mo PARTICLES AND AN Al MATRIX

    Directory of Open Access Journals (Sweden)

    YEON SOO KIM

    2013-12-01

    Full Text Available Interaction layer growth between U-Mo alloy fuel particles and Al in a dispersion fuel is a concern due to the volume expansion and other unfavorable irradiation behavior of the interaction product. To reduce interaction layer (IL growth, a small amount of Si is added to the Al. As a result, IL growth is affected by the Si content in the Al matrix. In order to predict IL growth during fabrication and irradiation, empirical models were developed. For IL growth prediction during fabrication and any follow-on heating process before irradiation, out-of-pile heating test data were used to develop kinetic correlations. Two out-of-pile correlations, one for the pure Al matrix and the other for the Al matrix with Si addition, respectively, were developed, which are Arrhenius equations that include temperature and time. For IL growth predictions during irradiation, the out-of-pile correlations were modified to include a fission-rate term to consider fission enhanced diffusion, and multiplication factors to incorporate the Si addition effect and the effect of the Mo content. The in-pile correlation is applicable for a pure Al matrix and an Al matrix with the Si content up to 8 wt%, for fuel temperatures up to 200 °C, and for Mo content in the range of 6 – 10wt%. In order to cover these ranges, in-pile data were included in modeling from various tests, such as the US RERTR-4, -5, -6, -7 and -9 tests and Korea's KOMO-4 test, that were designed to systematically examine the effects of the fission rate, temperature, Si content in Al matrix, and Mo content in U-Mo particles. A model converting the IL thickness to the IL volume fraction in the meat was also developed.

  15. A study of HANARO core conversion using high density U-Mo fuel

    International Nuclear Information System (INIS)

    Lee, K.H.; Lee, C.S.; Lee, B.C.; Park, S.J.; Kim, H.; Kim, C.K.

    2002-01-01

    Currently, HANARO is using 3.15gU/cc U3Si/Al as a driver fuel. HANARO has seven vertical irradiation holes in the core region. Three of them including a central trap are located in the inner region of the core and mainly being used for material irradiation tests. Four of them are located in the reflector tank but cooled by primary coolant. They are used for fuel irradiation tests or radioisotope development tests. For minimum core modification using high density U-Mo fuels, no dimension change is assumed in the current fuel rods and the cladding thickness remains the same in this study. The high density U-Mo fuel will have up to about twice the linear uranium loading of a current HANARO driver fuel. Using this high density fuel 8 fuel sites can be replaced with irradiation sites. Three kinds of conceptual cores are considered using 5 gU/cc U-7Mo/Al and 16 gU/cc U-7Mo. The increase of the linear heat generation rate due to the decrease of total fuel length can be overcome by more uniform radial and axial power distribution using different uranium densities and different fuel meat diameters are introduced into those cores. The new core has 4.54 times larger surface-to-volume ratio than the reference core. The core uranium loading, linear heat generation rate, excess reactivity, and control rod worth as well as the neutron spectra are analysed for each core. (author)

  16. Fission-induced recrystallization effect on intergranular bubble-driven swelling in U-Mo fuel

    Energy Technology Data Exchange (ETDEWEB)

    Liang, Linyun; Mei, Zhi-Gang; Yacout, Abdellatif M.

    2017-10-01

    We have developed a mesoscale phase-field model for studying the effect of recrystallization on the gas-bubble-driven swelling in irradiated U-Mo alloy fuel. The model can simulate the microstructural evolution of the intergranular gas bubbles on the grain boundaries as well as the recrystallization process. Our simulation results show that the intergranular gas-bubble-induced fuel swelling exhibits two stages: slow swelling kinetics before recrystallization and rapid swelling kinetics with recrystallization. We observe that the recrystallization can significantly expedite the formation and growth of gas bubbles at high fission densities. The reason is that the recrystallization process increases the nucleation probability of gas bubbles and reduces the diffusion time of fission gases from grain interior to grain boundaries by increasing the grain boundary area and decreasing the diffusion distance. The simulated gas bubble shape, size distribution, and density on the grain boundaries are consistent with experimental measurements. We investigate the effect of the recrystallization on the gas-bubble-driven fuel swelling in UMo through varying the initial grain size and grain aspect ratio. We conclude that the initial microstructure of fuel, such as grain size and grain aspect ratio, can be used to effectively control the recrystallization and therefore reduce the swelling in U-Mo fuel.

  17. Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

    Science.gov (United States)

    Burkes, Douglas E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

    2014-07-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

  18. Neutronic and thermal hydraulic analyses of LEU targets irradiated in a research reactor for Molybdenum-99 production

    International Nuclear Information System (INIS)

    Jo, Daeseong; Lee, Kyung-Hoon; Kim, Hong-Chul; Chae, Heetaek

    2014-01-01

    Highlights: • Neutronic and thermal hydraulic analyses of irradiated fuel plates for Molybdenum-99. • Heat production during and after irradiation was evaluated using MCNP and ORIGEN-APR. • Cooling capacities under various cooling conditions were evaluated using TMAP. • Natural convective cooling was adequate for the decay power after 0.03 h from withdrawal. • Maximum temperature of the target decayed for 24 h does not exceed the blistering threshold. - Abstract: Neutronic and thermal hydraulic analyses of irradiated fuel plates for Molybdenum-99 production in a research reactor were performed to investigate (1) the heat production during irradiation, (2) decay heat after irradiation, and (3) cooling capacities under various cooling conditions. The heat production on the target plates irradiated in the core was evaluated using the MCNP code. The decay heat after irradiation was evaluated using the ORIGEN-APR code, and compared against ANSI/ANS-5.1-1979. The cooling capacities of forced convective cooling during irradiation and natural convective cooling after irradiation were estimated using the TMAP code. An equilibrium core with different core statuses i.e., BOC, MOC, and EOC was used to evaluate power released from the targets and the axial power distribution. Based on the neutronic calculations, thermal margins i.e., the maximum wall temperature, minimum ONB temperature margin, and minimum CHF ratio were estimated, and the cooling strategy of the fission Mo targets was discussed. The targets were cooled by forced convective cooling during irradiation, and cooled by natural convective cooling after irradiation. For a further production process, the targets transported to a hot cell were exposed to the air, and cooled by natural convection cooling in air. As a result, the maximum wall temperature remained below the ONB temperature while the targets were under water, and the maximum wall temperature remained under the blistering limit while the targets

  19. Thermal and x-ray studies on Tl2U(MoO4)3 and Tl4U(MoO4)4

    International Nuclear Information System (INIS)

    Dahale, N.D.; Keskar, Meera; Kulkarni, N.K.; Singh Mudher, K.D.

    2006-01-01

    In the quaternary Tl-U(IV)-Mo-O system, two new compounds namely Tl 2 U(MoO 4 ) 3 and Tl 4 U(MoO 4 ) 4 were prepared and characterized by powder X-ray diffraction and thermal methods. These compounds were prepared by solid state reactions of Tl 2 MoO 4 , UMoO 5 and MoO 3 in the required stoichiometric ratio at 500 deg C in evacuated sealed quartz ampoule. The XRD data of Tl 2 U(MoO 4 ) 3 and Tl 4 U(MoO 4 ) 4 were indexed on orthorhombic cell. TG curves of Tl 2 U(MoO 4 ) 3 and Tl 4 U(MoO 4 ) 4 did not show any weight change up to 700 deg C in an inert atmosphere. During heating in an inert atmosphere, Tl 2 U(MoO 4 ) 3 and Tl 4 U(MoO 4 ) 4 showed endothermic Dta peaks due to melting of the compounds at 519 and 565 deg C, respectively. (author)

  20. Computational and experimental analysis of causes for local deformation of research reactor U-Mo fuel pin claddings in case of high burn-ups

    International Nuclear Information System (INIS)

    Popov, V.V.; Khmelevsky, M.Ya.; Lukichev, V.A.; Golosov, O.A.

    2005-01-01

    Post-reactor investigations of (U-Mo) fuel pins irradiated in the IVV-2M reactor have allowed to determine: the change in a fuel pin volume; the dimensions and the kind of the local deformation of fuel pin claddings; the amount of gases released under the cladding from the fuel composition, the thickness and appearance of the interaction layer of between the (U-Mo) particles and aluminium as a matrix material. The computational analysis of the stressed-strained state of fuel pins has shown that the major contribution to the increase of the fuel pin volume is made by the fuel swelling caused by the solid products of fission being formed in the process of operation. The emergence of the (U-Mo) fuel-aluminium matrix interaction layers around the (U-Mo) particles results in formation and evolution of lamination cavities inside the fuel composition under the joint action of the pressure of process gases and gaseous fission products. In case of high burn-up a local bulge of a fuel pin cladding is being formed in the fuel lamination area caused by the pressure of gases in the presence of creep in the fuel pin cladding material. The computational results relating to the local strain in a research reactor (U-Mo) fuel pin are in a good accordance with the results of the post-reactor investigations. (author)

  1. Modeling of high-density U-MO dispersion fuel plate performance

    International Nuclear Information System (INIS)

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

    2002-01-01

    Results from postirradiation examinations (PIE) of highly loaded U-Mo/Al dispersion fuel plates over the past several years have shown that the interaction between the metallic fuel particles and the matrix aluminum can be extensive, reducing the volume of the high-conductivity matrix phase and producing a significant volume of low-conductivity reaction-product phase. This phenomenon results in a significant decrease in fuel meat thermal conductivity during irradiation. PIE has further shown that the fuel-matrix interaction rate is a sensitive function of irradiation temperature. The interplay between fuel temperature and fuel-matrix interaction makes the development of a simple empirical correlation between the two difficult. For this reason a comprehensive thermal model has been developed to calculate temperatures throughout the fuel plate over its lifetime, taking into account the changing volume fractions of fuel, matrix and reaction-product phases within the fuel meat owing to fuel-matrix interaction; this thermal model has been incorporated into the dispersion fuel performance code designated PLATE. Other phenomena important to fuel thermal performance that are also treated in PLATE include: gas generation and swelling in the fuel and reaction-product phases, incorporation of matrix aluminum into solid solution with the unreacted metallic fuel particles, matrix extrusion resulting from fuel swelling, and cladding corrosion. The phenomena modeled also make possible a prediction of fuel plate swelling. This paper presents a description of the models and empirical correlations employed within PLATE as well as validation of code predictions against fuel performance data for U-Mo experimental fuel plates from the RERTR-3 irradiation test. (author)

  2. Characterization of interaction between U-Mo alloy and Al diffusion-couple

    International Nuclear Information System (INIS)

    Liu Yunming; Yin Changgeng; Sun Changlong; Chen Jiangang; Sun Xudong

    2011-01-01

    In this paper, the interaction behavior of U-Mo/Al was studied with the diffusion-couple method, and the couple was continuously jointed by hot-pressing with special device. Annealing experiments were accomplished in a vacuum hot-pressing furnace, and at 550∼570℃ for 5∼21 hours. The results show that the morphology and composition of interaction Layer depend on the interaction layer thickness. The content of U (Mo) and Al is mutational at the interface of U-Mo/interaction layer/Al. The layer close to U-Mo side is mainly composed of product (U, Mo)Al 3 , while the Al side is composed of (U, Mo)Al 4 and UMO 2 Al 20 . Diffusion process of U-Mo/Al is Al immigrating over the Al/U-Mo original interface into U-Mo side and reacting with U-Mo, subsequently the interaction layer is growing into Al. (authors)

  3. Alternative Crucibles for U-Mo Microwave Melting

    Energy Technology Data Exchange (ETDEWEB)

    Kirby, Brent W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-03-31

    The crucibles used currently for microwave melting of U-Mo alloy at the Y-12 Complex contain silicon carbide (SiC) in a mullite (3Al2O3-2SiO2) matrix with an erbia coating in contact with the melt. Due to observed silicon contamination, Pacific Northwest National Laboratory has investigated alternative crucible materials that are susceptible to microwave radiation and are chemically compatible with molten U-Mo at 1400 1500C. Recommended crucibles for further testing are: 1) high-purity alumina (Al2O3); 2) yttria-stabilized zirconia (ZrO2); 3) a composite of alumina and yttria-stabilized zirconia; 4) aluminum nitride (AlN). Only AlN does not require an erbia coating. The recommended secondary susceptor, for heating at low temperature, is SiC in a “picket fence” arrangement.

  4. Powder fabrication of U-Mo alloys for nuclear dispersion fuels

    International Nuclear Information System (INIS)

    Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes

    2011-01-01

    For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

  5. Powder fabrication of U-Mo alloys for nuclear dispersion fuels

    Energy Technology Data Exchange (ETDEWEB)

    Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes, E-mail: mdurazzo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

  6. Fuel performance of rod-type research reactor fuel using a centrifugally atomized U-Mo powder

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Lee, Yoon Sang; Kim, Chang Kyu

    2009-01-01

    A low enriched uranium nuclear fuel for research reactors has been developed in order to replace a highly enriched uranium fuel according to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program. In KAERI, a rod-type U 3 Si dispersion fuel has been developed for a localization of the HANARO fuel and a U 3 Si/Al dispersion fuel of 3.15 gU/cc has been used at HANARO as a driver fuel since 2005. Although uranium silicide dispersion fuels such as U 3 Si 2 /Al and U 3 Si/Al are being used widely, high uranium density dispersion fuels (8-9 g/cm 3 ) are required for some high performance research reactors. U-Mo alloys have been considered as one of the most promising uranium alloys for a dispersion fuel due to their good irradiation performance. An international qualification program on U-Mo fuel to replace a uranium silicide dispersion fuel with a U-Mo dispersion fuel has been carried out

  7. Neutronic analysis for the fission Mo99 production by irradiation of leu targets in TRIGA 14 MW reactor

    International Nuclear Information System (INIS)

    Dulugeac, S. D.; Mladin, M.; Budriman, A. G.

    2013-01-01

    Molybdenum production can be a solution for the future in the utilization of the Romanian TRIGA, taking into account the international market supply needs. Generally, two different techniques are available for Mo 99 production for use in medical Tc 99 generation.The first one is based on neutron irradiation of molybdenum targets of natural isotopic composition or enriched in Mo 98 . In a second process, Mo 99 is obtained as a result of the neutron induced fission of U 235 according to U 235 (n,f) Mo 99 . The objectives of the paper are related to Mo 99 production as a result of fission. Neutron physics parameters are determined and presented, such as: thermal flux axial distribution for the critical reactor at 10 MW inside the irradiation location; reactivity introduced by three Uranium foil containers; neutron fluxes and fission rates in the Uranium foils; released and deposited power in the Uranium foils; Mo 99 activity in the Uranium foils. (authors)

  8. Development of a PVD-based manufacturing process of monolithic LEU irradiation targets for {sup 99}Mo production

    Energy Technology Data Exchange (ETDEWEB)

    Hollmer, Tobias

    2015-08-03

    {sup 99}Mo is the most important radioisotope in nuclear medicine. It is produced by fission of uranium in irradiation targets. The usage of cylindrical monolithic targets can ensure a safe supply of {sup 99}Mo and at the same reduce the amount of highly radioactive waste generated during production. To manufacture these targets, a novel PVD-based technique was developed. Both the feasibility and the high efficiency of this process were demonstrated in a prototype apparatus.

  9. Development of technology of high density LEU dispersion fuel fabrication

    International Nuclear Information System (INIS)

    Wiencek, T.; Totev, T.

    2007-01-01

    Advanced Materials Fabrication Facilities at Argonne National Laboratory have been involved in development of LEU dispersion fuel for research and test reactors from the beginning of RERTR program. This paper presents development of technology of high density LEU dispersion fuel fabrication for full size plate type fuel elements. A brief description of Advanced Materials Fabrication Facilities where development of the technology was carried out is given. A flow diagram of the manufacturing process is presented. U-Mo powder was manufactured by the rotating electrode process. The atomization produced a U-Mo alloy powder with a relatively uniform size distribution and a nearly spherical shape. Test plates were fabricated using tungsten and depleted U-7 wt.% Mo alloy, 4043 Al and Al-2 wt% Si matrices with Al 6061 aluminum alloy for the cladding. During the development of the technology of manufacturing of full size high density LEU dispersion fuel plates special attention was paid to meet the required homogeneity, bonding, dimensions, fuel out of zone and other mechanical characteristics of the plates.

  10. Thermomechanical DART code improvements for LEU VHD dispersion and monolithic fuel element analysis

    International Nuclear Information System (INIS)

    Taboada, H.; Saliba, R.; Moscarda, M.V.; Rest, J.

    2005-01-01

    A collaboration agreement between ANL/US DOE and CNEA Argentina in the area of Low Enriched Uranium Advanced Fuels has been in place since October 16, 1997 under the Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy. An annex concerning DART code optimization has been operative since February 8, 1999. Previously, as a part of this annex a visual FASTDART version and also a DART THERMAL version were presented during RERTR 2000, 2002 and RERTR 2003 Meetings. During this past year the following activities were completed: Optimization of DART TM code Al diffusion parameters by testing predictions against reliable data from RERTR experiments. Improvements on the 3-D thermo-mechanical version of the code for modeling the irradiation behavior of LEU U-Mo monolithic fuel. Concerning the first point, by means of an optimization of parameters of the Al diffusion through the interaction product theoretical expression, a reasonable agreement between DART temperature calculations with reliable RERTR PIE data was reached. The 3-D thermomechanical code complex is based upon a finite element thermal-elastic code named TERMELAS, and irradiation behavior provided by the DART code. An adequate and progressive process of coupling calculations of both codes at each time step is currently developed. Compatible thermal calculation between both codes was reached. This is the first stage to benchmark and validate against RERTR PIE data the coupling process. (author)

  11. Heat Generation Effects on U-Mo/Al through ABAQUS FEM Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Taewon; Jeong, Gwan Yoon; Lee, Cheol Min; Sohn Dongseong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    U-Mo/Al dispersion fuels have been considered a most promising candidate for a replacement of Highly Enriched Uranium (HEU) fuel in many research reactors. Coulson developed a FEM model which show the fuel meat realistically and compared the thermal conductivity results of two and three dimensional model. Williams also developed a FEM model which are different from the former in that it use regularly meshed unit cells. He showed a heat generation effects through FEM simulation and the effective thermal conductivity of the fuel with heat generated in the fuel particles is a little lower than that of the fuel with no heat generated. In the current work, the heat generation effects are analyzed and discussed in a wider range of volume fraction with more realistic models by using ABAQUS finite element package. The FEM model is used to determine the effective thermal conductivity of U-Mo/Al and to simulate the heat generation effects in the study. This model reflected the microscopic morphology of the fuel very well by making random distribution particles although the particle shape is considered as sphere. All simulation results show the heat generation effects although the effects are small when the volume fraction of fuels are high. When the particles are surrounded with interaction layers, the heat transfer from the particle to matrix is disturbed by interaction layers due to the low thermal conductivity of interaction layers. However this effects decreases when the sum of the volume fraction of fuels and interaction layers exceeds 40-50 vol% because a great portion of the heat must pass through fuels and interaction layers although the heat is applied on the surface. Therefore particle size and initial particle volume fractions will be the important factors for the heat generation effects when interaction layers grow during irradiations.

  12. 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)

  13. A modelling study of the inter-diffusion layer formation in U-Mo/Al dispersion fuel plates at high power

    Science.gov (United States)

    Ye, B.; Hofman, G. L.; Leenaers, A.; Bergeron, A.; Kuzminov, V.; Van den Berghe, S.; Kim, Y. S.; Wallin, H.

    2018-02-01

    Post irradiation examinations of full-size U-Mo/Al dispersion fuel plates fabricated with ZrN- or Si- coated U-Mo particles revealed that the reaction rate of irradiation-induced U-Mo-Al inter-diffusion, an important microstructural change impacting the performance of this type of fuel, transited at a threshold temperature/fission rate. The existing inter-diffusion layer (IL) growth correlation, which does not describe the transition behavior of IL growth, was modified by applying a temperature-dependent multiplication factor that transits around a threshold fission rate. In-pile irradiation data from four tests in the BR2 reactors, including FUTURE, E-FUTURE, SELEMIUM, and SELEMIUM-1a, were utilized to determine and validate the updated IL growth correlation. Irradiation behavior of the plates was simulated with the DART-2D computational code. The general agreement between the calculated and measured fuel meat swelling and constituent volume fractions as a function of fission density demonstrated the plausibility of the updated IL growth correlation. The simulation results also suggested the temperature dependence of the IL growth rate, similar to the temperature dependence of the inter-mixing rate in ion-irradiated bi-layer systems.

  14. Program description for the qualification of CNEA - Argentina as a supplier of LEU silicide fuel and post-irradiation examinations plan for the first prototype irradiated in Argentina

    International Nuclear Information System (INIS)

    Rugirello, Gabriel; Adelfang, Pablo; Denis, Alicia; Zawerucha, Andres; Marco, Agustin di; Guillaume, Eduardo; Sbaffoni, Monica; Lacoste, Pablo

    1998-01-01

    In this report we present a description of the ongoing and future stages of the program for the qualification of CNEA, Argentina, as a supplier of low enriched uranium silicide fuel elements for research reactor. Particularly we will focus on the characteristics of the future irradiation experiment on a new detachable prototype, the post-irradiation examinations (PIE) plan for the already irradiated prototype PO4 and an overview of the recently implemented PIE facilities and equipment. The program is divided in several steps, some of which have been already completed. It concludes: development of the uranium silicide fissile material, irradiation and PIE of several full-scale prototypes. Important investments have been already carried out in the facilities for the FE production and PIE. (author)

  15. Development, irradiation testing and PIE of UMo fuel at AECL

    International Nuclear Information System (INIS)

    Sears, D.F.

    2005-01-01

    This paper reviews recent U-Mo dispersion fuel development, irradiation testing and postirradiation examination (PIE) activities at AECL. Low-enriched uranium fuel alloys and powders have been fabricated at Chalk River Labs, with compositions ranging from U-7Mo to U-10Mo. The bulk alloys and powders were characterized using optical and scanning electron microscopy, chemical analysis, X-ray diffraction and neutron diffraction analysis. The analyses confirmed that the powders were of high quality, and in the desired gamma phase. Subsequently, kilogram quantities of DU-Mo and LEU-Mo powder have been manufactured for commercial customers. Mini-elements have been fabricated with LEU-7Mo and LEU-10Mo dispersed in aluminum, with a nominal loading of 4.5 gU/cm 3 . These have been irradiated in the NRU reactor at linear powers up to 100 kW/m. The mini-elements achieved 60 atom% 235 U burnup in 2004 March, and the irradiation is continuing to a planned discharge burnup of 80 atom% 235 U. Interim PIE has been conducted on mini-elements that were removed after 20 atom% 235 U burnup. The PIE results are presented in this paper. (author)

  16. 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

  17. Code structure for U-Mo fuel performance analysis in high performance research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Cho, Tae Won; Lee, Chul Min; Sohn, Dong Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    A performance analysis modeling applicable to research reactor fuel is being developed with available models describing fuel performance phenomena observed from in-pile tests. We established the calculation algorithm and scheme to best predict fuel performance using radio-thermo-mechanically coupled system to consider fuel swelling, interaction layer growth, pore formation in the fuel meat, and creep fuel deformation and mass relocation, etc. In this paper, we present a general structure of the performance analysis code for typical research reactor fuel and advanced features such as a model to predict fuel failure induced by combination of breakaway swelling and pore growth in the fuel meat. Thermo-mechanical code dedicated to the modeling of U-Mo dispersion fuel plates is being under development in Korea to satisfy a demand for advanced performance analysis and safe assessment of the plates. The major physical phenomena during irradiation are considered in the code such that interaction layer formation by fuel-matrix interdiffusion, fission induced swelling of fuel particle, mass relocation by fission induced stress, and pore formation at the interface between the reaction product and Al matrix.

  18. Selenium fuel: Surface engineering of U(Mo) particles to optimise fuel performance

    International Nuclear Information System (INIS)

    Van den Berghe, S.; Leenaers, A.; Detavernier, C.

    2010-01-01

    Recent developments on the stabilisation of U(Mo) in-pile behaviour in plate-type fuel have focussed almost exclusively on the addition of Si to the Al matrix of the fuel. This has now culminated in a qualification effort in the form of the European LEONIDAS initiative for which irradiations will start in 2010. In this framework, many discussions have been held on the Si content of the matrix needed for stabilisation of the interaction phase and the requirement for the formation of Si-rich layers around the particles during the fabrication steps. However, it is clear that the Si needs to be incorporated in the interaction phase for it to be effective, for which the currently proposed methods depend on a diffusion mechanism, which is difficult to control. This has lead to the concept of a Si coated particle as a more efficient way of incorporating the Si in the fuel by putting it immediately where it will be required : at the fuel-matrix interface. As part of the SELENIUM (Surface Engineered Low ENrIched Uranium-Molybdenum fuel) project, SCK CEN has built a sputter coater for PVD magnetron sputter coating of particles in collaboration with the University of Ghent. The coater is equipped with three 3 inch magnetron sputter heads, allowing deposition of 3 different elements or a single element at high deposition speed. The particles are slowly rotated in a drum to produce homogeneous layer thicknesses. (author)

  19. Study on characteristics of U-Mo/Al-Si interaction layers of dispersion fuel plates

    International Nuclear Information System (INIS)

    Liu Lijian; Yin Changgeng; Chen Jiangang; Sun Changlong; Liu Yunming

    2014-01-01

    In this paper, we analyzed the characteristics of U-Mo/Al-Si interaction layers of dispersion fuel plates. The results show that the interaction layers (IL) are with irregular morphology and uneven thickness, and are mainly formed in the internal micro cracks of the dispersion fuel particles or at the interface between the particles and the substrates. The diffusion mechanism of U-Mo/Al-Si is the vacancy diffusion, Al and Si are migrating elements, and the diffusion reaction is that Al and Si diffuse to U-Mo alloy. Inside the interaction layers, the Al content keeps constant basically, but the Si content gradually increases with the substrate-fuel direction, and the maximum content of Si appears interaction layers near the U-Mo side. Adding about 5 wt% Si into Al matrix can restrain the diffusion reaction, and improve the performance of dispersion fuel plates finally. (authors)

  20. Development of dispersion U(Mo)/Al–Si miniplates fabricated at 500 °C with Al 6061 as cladding

    Energy Technology Data Exchange (ETDEWEB)

    Mirandou, M.I., E-mail: mirandou@cnea.gov.ar [Gerencia Materiales-GAEN-CNEA, Avda. Gral. Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina); Aricó, S.F. [Gerencia Materiales-GAEN-CNEA, Avda. Gral. Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina); Instituto Sabato UNSAM-CNEA, Avda. Gral. Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina); Balart, S.N. [Gerencia Materiales-GAEN-CNEA, Avda. Gral. Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina); Fabro, J.O. [Departamento ECRI, Gerencia de Ciclo del Combustible Nuclear, CNEA, Avda. Gral. Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina)

    2015-02-15

    In the frame of U(Mo) dispersion fuel elements qualification, Si additions to Al matrix arose as a promising solution to the unacceptable failures found when pure Al is used. Analysis of as-fabricated fuel plates made with Al–Si matrices demonstrated that good irradiation behavior is correlated with the formation during fabrication of a Si-containing interaction layer around the U(Mo) particles. Thus, the analysis of the influence of fabrication parameters becomes important. Studies on Al–Si dispersion miniplates fabricated in CNEA, Argentina, have been initiated to determine how to obtain the better interaction layer characteristics with the lesser modifications to the fabrication process and the smaller amount of Si in the matrix. In this work results for miniplates made of atomized U–7 wt%Mo particles dispersed in Al–2 wt%Si and Al–4 wt%Si matrices, obtained by mixing pure Al and Si powders, and Al 6061 as cladding are presented. Interaction layer grown during fabrication process (500 °C) consists of Si-containing phases being U(Al, Si){sub 3} its principal component. Its uniformity is not satisfactory due to the formation of an oxide layer.

  1. Investigation of the microstructure influence in the thermo-physical properties of U-Mo alloys through the laser flash method

    Energy Technology Data Exchange (ETDEWEB)

    Pedrosa, Tercio A.; Alves, Fabio F.; Kelmer, Paula F.; Santos, Ana Maria M.; Camarano, Denise das M.; Ferraz, Wilmar B., E-mail: tap@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    The U-Mo alloys are the most investigated and promising nuclear fuel material to be used in research and test reactors, according to the premises of the RERTR program, whose objective is to minimize the threats of nuclear weapons proliferation through the conversion of the nuclear fuels of research and test reactors form a high enrichment grade, HEU (235U>90%, to a low enrichment grade, LEU ({sup 235}U<20%). The high density of the U-Mo alloys associated with its ability to keep the gamma phase metastable at room temperature are the main advantages of these alloys, with Mo contents of 5, 7 and 10 wt% were induction melted and ageing heat treated at 300 and 500 deg C for 72, 120 and 240 h. Microstructural characterization was carried out in the as-cast and aged conditions through XRD and OM techniques. The laser Flash Method at environmental temperature was employed to investigate the variation of the thermal diffusivity as a function of the microstructure obtained in the as-cast and aged conditions. (author)

  2. Investigation of the microstructure influence in the thermo-physical properties of U-Mo alloys through the laser flash method

    International Nuclear Information System (INIS)

    Pedrosa, Tercio A.; Alves, Fabio F.; Kelmer, Paula F.; Santos, Ana Maria M.; Camarano, Denise das M.; Ferraz, Wilmar B.

    2013-01-01

    The U-Mo alloys are the most investigated and promising nuclear fuel material to be used in research and test reactors, according to the premises of the RERTR program, whose objective is to minimize the threats of nuclear weapons proliferation through the conversion of the nuclear fuels of research and test reactors form a high enrichment grade, HEU (235U>90%, to a low enrichment grade, LEU ( 235 U<20%). The high density of the U-Mo alloys associated with its ability to keep the gamma phase metastable at room temperature are the main advantages of these alloys, with Mo contents of 5, 7 and 10 wt% were induction melted and ageing heat treated at 300 and 500 deg C for 72, 120 and 240 h. Microstructural characterization was carried out in the as-cast and aged conditions through XRD and OM techniques. The laser Flash Method at environmental temperature was employed to investigate the variation of the thermal diffusivity as a function of the microstructure obtained in the as-cast and aged conditions. (author)

  3. Analysis of Neutron Flux Distribution in Rsg-Gas Reactor With U-Mo Fuels

    Directory of Open Access Journals (Sweden)

    Taswanda Taryo

    2004-01-01

    Full Text Available The use of U-Mo fuels in research reactors seems to be promising and, recently, world researchers have carried out these such activities actively. The National Nuclear Energy Agency (BATAN which owns RSG-GAS reactor available in Serpong Research Center for Atomic Energy should anticipate this trend. It is, therefore, this research work on the use of U-Mo fuels in RSG-GAS reactor should be carried out. The work was focused on the analysis of neutron flux distribution in the RSG-GAS reactor using different content of molybdenum in U-Mo fuels. To begin with, RSG-GAS reactor core model was developed and simulated into X, Y and Z dimensions. Cross section of materials based on the developed cells of standard and control fuels was then generated using WIMS-D5-B. The criticality calculations were finally carried out applying BATAN-2DIFF code. The results showed that the neutron flux distribution obtained in U-Mo-fuel-based RSG-GAS core is very similar to those achieved in the 300-gram sillicide-fuel-based RSG-GAS reactor core. Indeed, the utilization of the U-Mo RSG-GAS core can be very similar to that of the high-density sillicide reactor core and even could be better in the future.

  4. Improvement of Silicide Coating Method as Diffusion Barrier for U-Mo Dispersion Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Ji Min; Kim, Sunghwan; Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The excessive interaction between the U-Mo alloys and their surrounding Al matrix lead to excessive local swelling called 'pillowing'. For this reason, KAERI suggested several remedies such as alloying U-Mo with Ti, or Al matrix with Si. In addition, silicide or nitride coatings on the surface of U-Mo particles have also been proposed to hinder the growth of the interaction layer. In this study, centrifugally atomized U-Mo-Ti alloy powders were coated with silicide layers. The coating process was improved when compared to the previous coating in terms of the ball milling and heat treatment conditions. Subsequently, silicide coated U-Mo-Ti powders and pure aluminum powders were mixed and made into a compact for the annealing test. The compacts were annealed at 550 .deg. C for 2hr, and characterized using scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDS). 1. Uniform, homogeneous, thickness controllable silicide layers were successfully coated on the surface of U-7wt%Mo-1wt%Ti powders. 2. U{sub 3}Si, U{sub 3}Si{sub 2} silicide layers formed on the surface of U-7wt%Mo-1wt%Ti powders, and were identified by XRD and EDS analyses.

  5. Modeling solute segregation during the solidification of γ-phase U-Mo alloys

    Energy Technology Data Exchange (ETDEWEB)

    Steiner, M.A., E-mail: mas4cw@virginia.edu [University of Virginia, Material Science and Engineering, 395 McCormick Rd, Charlottesville, VA 22904 (United States); Garlea, E. [Y-12 National Security Complex, Oak Ridge, TN 37831 (United States); Agnew, S.R. [University of Virginia, Material Science and Engineering, 395 McCormick Rd, Charlottesville, VA 22904 (United States)

    2016-06-15

    Using first principles calculations, it is demonstrated that solute segregation during U-Mo solidification can be modeled using the classic Brody-Fleming limited diffusion framework. The necessary supporting equations specific to the U-Mo alloy, along with careful verification of the assumptions underpinning the Brody-Fleming model are developed, allowing for concentration profile predictions as a function of alloy composition and cooling rate. The resulting model is compared to experimental solute concentration profiles, showing excellent agreement. Combined with complementary modeling of dendritic feature sizes, the solute segregation model can be used to predict the complete microstructural state of individual U-Mo volume elements based upon cooling rates, informing ideal processing routes.

  6. 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

  7. 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.

  8. 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

  9. Reaction layer growth and reaction heat of U-Mo/Al dispersion fuels using centrifugally atomized powders

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Han, Young Soo; Park, Jong Man; Park, Soon Dal; Kim, Chang Kyu

    2003-01-01

    The growth behavior of reaction layers and heat generation during the reaction between U-Mo powders and the Al matrix in U-Mo/Al dispersion fuels were investigated. Annealing of 10 vol.% U-10Mo/Al dispersion fuels at temperatures from 500 to 550 deg. C was carried out for 10 min to 36 h to measure the growth rate and the activation energy for the growth of reaction layers. The concentration profiles of reaction layers between the U-10Mo vs. Al diffusion couples were measured and the integrated interdiffusion coefficients were calculated for the U and Al in the reaction layers. Heat generation of U-Mo/Al dispersion fuels with 10-50 vol.% of U-Mo fuel during the thermal cycle from room temperature to 700 deg. C was measured employing the differential scanning calorimetry. Exothermic heat from the reaction between U-Mo and the Al matrix is the largest when the volume fraction of U-Mo fuel is about 30 vol.%. The unreacted fraction in the U-Mo powders increases as the volume fraction of U-Mo fuel increases from 30 to 50 vol.%

  10. Construction of a sputtering reactor for the coating and processing of monolithic U-Mo nuclear fuel

    International Nuclear Information System (INIS)

    Schmid, Wolfgang

    2011-01-01

    In the presented thesis sputter deposition was used for the first time to coat monolithic U-Mo nuclear fuel foils with diffusion inhibitive materials. The intention of these coatings is to prevent the formation of an interdiffusion layer between U-Mo and Al cladding during the use of the fuel. A small sputtering reactor was built, in which the method was tested and processing parameters were investigated. In parallel a larger sputtering reactor was constructed, that allows to coat full size monolithic U-Mo nuclear fuel foils and was used to test an industrial application of the technique. As a result a method based on sputter deposition and erosion can be presented, that allows to clean as well as to coat the surface of monolithic U-Mo nuclear fuel foils in excellent quality. It can be included at any time into the manufacturing chain for U-Mo fuel elements, which is currently being developed.

  11. The manufacture of LEU fuel elements at Dounreay

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, J.

    1997-08-01

    Two LEU test elements are being manufactured at Dounreay for test irradiation in the HFR at Petten, The Netherlands. This paper describes the installation of equipment and the development of the fabrication and inspection techniques necessary for the manufacture of LEU fuel plates. The author`s experience in overcoming the technical problems of stray fuel particles, dog-boning, uranium homogeneity and the measurement of uranium distribution is also described.

  12. RERTR progress in Mo-99 production from LEU

    Energy Technology Data Exchange (ETDEWEB)

    Vandegrift, G.F.; Conner, C.; Aase, S.; Bakel, A.; Bowers, D.; Freiberg, E.; Gelis, A.; Quigley, K.J.; Snelgrove, J.L. [Argonne National Laboratory 9700 S. Cass Avenue, Argonne, IL (United States)

    2002-07-01

    The ANL RERTR program is performing R and D supporting conversion of {sup 99}Mo production from HEU to LEU targets. Irradiation and processing of LEU targets were demonstrated at the Argentine Ezeiza Atomic Center. Target irradiation and disassembly were flawless, but the processing is not fully developed. In addition to preparing for, assisting in, and analyzing results of the demonstration, we performed other R and D related to LEU conversion: (1) designing a prototype production dissolver for digesting irradiated LEU foils in alkaline solutions and developing means to simplify digestion, (2) modifying ion-exchange columns used in the CNEA recovery and purification of {sup 99}Mo to deal with the lower volumes generated from LEU-foil digestion, (3) measuring the performance of new inorganic sorbents that outperform alumina for recovering Mo(VI) from nitric acid solutions containing high concentrations of uranium nitrate, and (4) developing means to facilitate the concentration and calcination of waste nitric-acid/LEU-nitrate solutions from {sup 99} Mo production. (author)

  13. Microstructural studies on chemical interactions in U-Mo with Al

    International Nuclear Information System (INIS)

    Martins, Ilson Carlos

    2010-01-01

    This research refers to the study of U-Mo alloy as an alternative material for producing nuclear fuel elements with high density of uranium, for research reactors of high performance. The international non-proliferation of nuclear weapons has enrichment level limited to 20% U 23 '5. U-Mo alloys with 6-10 wt% Mo can lead to a density up to 9 gU/cm 3 , inside the fuel core. The MTR fuel element plates are made from briquettes (U-Mo powder + Al) encapsulated in Al plates, then welded and rolled However, the U-Mo alloy is very reactive in the presence of Al. The reaction products of this interaction are undesirable from the standpoint of nuclear usage, since they cause a chemical interaction layer (IL) formed during thermal cycling and exposure to nuclear fission neutrons. As the IL has low thermal conductivity, they may cause structural failure in the fuel element during operation. The present study provides a new preparation technique for interdiffusion pairs made by hot rolling. The U-Mo alloy, in tablet format, is involved by matrix Al-plates, which is sealed and then hot rolled. This way to prepare the diffusion couples is an ideal condition to avoid the oxidation at the contact interface at U-Mo/Al. The hot rolling preparation also simulates the first reduction pass during MTR fuel plate manufacture. We chose to work with a Mo content of 10 wt% in U-Mo alloy to ensure greater phase formation, since this level favors a greater chemical stability in this phase. The Al alloy matrix was used as the AA1050 since it contains small impurity amounts. The interdiffusion couples U-10Mo/AA1050 were thermally treated in two temperature ranges (1500C and 5500C) and three soaking times (5h, 40h and 80h) to simulate the interdiffusion process and formation of chemical interaction layer. The analysis of the interaction layer U-10Mo/AA1050 was made by SEM/EDS and X-ray diffraction. It revealed a general trend of low interdiffusion of Al (about 8 atomic %) inside U-Mo. There was

  14. Performance of PARR-1 with LEU Fuel

    International Nuclear Information System (INIS)

    Pervez, S.; Latif, M.; Bokhari, I.H.; Bakhtyar, S.

    2005-01-01

    Pakistan Research Reactor (PARR-1) went critical in 1965 with HEU fuel. The reactor core was converted to LEU fuel with power upgradation from 5 MW to 10 MW in 1992. The reactor has been operated with LEU fuel for about 10,000 hours and has produced about 66,000 MWh energy up to now. Average burn up of the irradiated fuel is about 42 %. The fuel performance during the last 12 years has been excellent. Post irradiation visual inspection of the fuel has revealed no abnormality. During operation there have been no signs of releases in the pool water establishing the full integrity of this fuel. The reactor has been mainly utilized for radioisotope production, beam tube experiments including neutron diffraction studies, neutron radiography etc. Studies have been completed to operate the reactor with a mixed core (HEU + LEU) to utilize the less burned HEU fuel elements. A major project of production of fission Moly using PARR-1 is in the final stages. (author)

  15. DENSITY-FUNCTIONAL STUDY OF U-Mo AND U-Zr ALLOYS

    Energy Technology Data Exchange (ETDEWEB)

    Landa, A; Soderlind, P; Turchi, P A

    2010-11-01

    Density-functional theory previously used to describe phase equilibria in U-Zr alloys [A. Landa, P. Soederlind, P.E.A. Turchi, J. Alloys Comp. 478 (2009) 103-110] is extended to investigate the ground-state properties of U-Mo solid solutions. We discuss how the heat of formation in both alloys correlates with the charge transfer between the alloy components, and how the specific behavior of the density of states in the vicinity of the Fermi level promotes the stabilization of the U{sub 2}Mo compound. Our calculations prove that, due to the existence of a single {gamma}-phase over the typical fuel operation temperatures, {gamma}-U-Mo alloys should indeed have much lower constituent redistribution than {gamma}-U-Zr alloys for which binodal decomposition causes a high degree of constituent redistribution.

  16. Status of LEU conversion program at CRNL

    International Nuclear Information System (INIS)

    Kennedy, I.C.

    1991-01-01

    After briefly reviewing the salient features of the NRU Reactor at Chalk River Nuclear Laboratories (CRNL), the progress of our LEU fuel development and testing program is described. The results (to date) of full-size prototype fuel-rod irradiations are reviewed, and the status of the new fuel-fabrication facility on the site is updated. Although development work is proceeding on U 3 Si 2 dispersions, all indications so far are that CRNL's U 3 Si fuel is fully acceptable for reactor operation. Fuel rods from the new fabrication shop will be installed in NRU in 1990, and the complete core conversion of NRU to LEU driver fuel is expected by 1991. (orig.)

  17. Powder production of U-Mo alloy, HMD process (Hydriding- Milling- Dehydriding)

    Energy Technology Data Exchange (ETDEWEB)

    Pasqualini, E. E.; Garcia, J.H.; Lopez, M.; Cabanillas, E.; Adelfang, P. [Dept. Combustibles Nucleares. Comision Nacional de Energia Atomica, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina)

    2002-07-01

    Uranium-molybdenum (U-Mo) alloys can be hydrided massively in metastable {gamma} (gamma) phase. The brittle hydride can be milled and dehydrided to acquire the desired size distributions needed for dispersion nuclear fuels. The developments of the different steps of this process called hydriding-milling- dehydriding (HMD Process) are described. Powder production scales for industrial fabrication is easily achieved with conventional equipment, small man-power and low investment. (author)

  18. Powder production of U-Mo alloy, HMD process (Hydriding- Milling- Dehydriding)

    International Nuclear Information System (INIS)

    Pasqualini, E. E.; Garcia, J.H.; Lopez, M.; Cabanillas, E.; Adelfang, P.

    2002-01-01

    Uranium-molybdenum (U-Mo) alloys can be hydrided massively in metastable γ (gamma) phase. The brittle hydride can be milled and dehydrided to acquire the desired size distributions needed for dispersion nuclear fuels. The developments of the different steps of this process called hydriding-milling- dehydriding (HMD Process) are described. Powder production scales for industrial fabrication is easily achieved with conventional equipment, small man-power and low investment. (author)

  19. Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

    Science.gov (United States)

    Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun

    2015-12-01

    U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

  20. 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)

  1. Neutronic study on conversion of SAFARI-1 to LEU silicide fuel

    International Nuclear Information System (INIS)

    Ball, G.; Pond, R.; Hanan, N.; Matos, J.

    1995-01-01

    This paper marks the initial study into the technical and economic feasibility of converting the SAFARI-1 reactor in South Africa to LEU silicide fuel. Several MTR assembly geometries and LEU uranium densities have been studied and compared with MEU and HEU fuels. Two factors of primary importance for conversion of SAFARI-1 to LEU fuel are the economy of the fuel cycle and the performance of the incore and excore irradiation positions

  2. Thermal expansion studies on UMoO5, UMoO6, Na2U(MoO4)3 and Na4U(MoO4)4

    International Nuclear Information System (INIS)

    Keskar, Meera; Dahale, N.D.; Krishnan, K.

    2009-01-01

    In the present work, thermal expansion behavior of lower valent sodium uranium molybdates, i.e., Na 2 U(MoO 4 ) 3 and Na 4 U(MoO 4 ) 4 were studied under vacuum in the temperature range of 298-873 K using high temperature X-ray diffractometry (HTXRD). Expansion behaviors of UMoO 5 and UMoO 6 were also studied in vacuum from 298 to 873 K and 773 K, respectively. UMoO 5 was synthesized by reacting UO 2 with MoO 3 in equi-molar proportion in evacuated sealed quartz ampoule at 1173 K for 14 h. Na 2 U(MoO 4 ) 3 and Na 4 U(MoO 4 ) 4 were prepared by reacting UMoO 5 and MoO 3 with 1 and 2 moles of Na 2 MoO 4 , respectively, at 873 K in evacuated sealed quartz ampoule. XRD data of UMoO 5 and UMoO 6 were indexed on orthorhombic and monoclinic systems, respectively, whereas, the data of Na 2 U(MoO 4 ) 3 and Na 4 U(MoO 4 ) 4 were indexed on tetragonal system. The lattice parameters and cell volume of all the four compounds, fit into polynomial expression with respect to temperature, showed positive thermal expansion (PTE) up to 873 K.

  3. Characterization of fuel miniplates fabricated with U(Mo) particles dispersed in Al-Si matrices

    International Nuclear Information System (INIS)

    Arico, S F; Mirandou, M I; Balart, S N; Fabro, J O

    2012-01-01

    In 2011 ECRI facility (Depto. ECRI, GCCN, CNEA) restarted the development for the fabrication of dispersion miniplates fuel elements in Al-Si matrix. This miniplates are fabricated with atomized U-7wt%Mo particles dispersed in a matrix formed by a mixture of pure Al and pure Si powders. The first results for an Al-4wt%Si matrix were presented at the AATN 2011 Annual Meeting. In this work, new results from the microstructural characterization of the meat in Al- 2wt%Si and pure Al miniplates are presented and compared with the previous ones. It is the intention to study the influence of the fabrication parameters as well as different Si concentration in the matrix, on the formation and characteristics of the interaction layer formed between the particles and the matrix at the end of the fabrication process. According to the results presented in this work an improvement can be observed on miniplates with Al-Si matrix respect to the one with pure Al. On the miniplates with Al- Si matrix, almost 100 % of the U(Mo) particles presented, at least in some fraction of its surface, an interaction layer composed by phases that contain Si. Moreover its morphological characteristics are independent of the crystallographic state of the U(Mo) particles. However, the oxide layer formed on the U(Mo) during the hot rolling acts as a barrier to the formation of the interaction layer. As a consequence, it is then mandatory to introduce some changes on the fabrication parameters to avoid, or at least minimize, this oxide layer (author)

  4. 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

  5. The genesis of Kurišková U-Mo ore deposit

    International Nuclear Information System (INIS)

    Demko, R.; Biroň, A.; Novotný, L.; Bartalský, B.

    2014-01-01

    The U-Mo ores of the known uranium deposit Kurišková located in the Huta volcano-sedimentary complex (HVC) of lower Permian age belongs to the Petrova Hora Formation of the North-Gemeric tectonic unit (Western Carpathians). The HVC is built up by volcanic rocks of bimodal basalt-rhyolite association, intercalated with sandstones, mudstones and claystones. Based on the sedimentary facies reconstruction, it is supposed paleoenvironment of seasonally flooded shallow lakes of continental fluvial plain with transition to estuaries and shallow marine facies of continental shelf in the upper part of HVC.

  6. 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)

  7. Status of LEU fuel development and conversion of NRU

    International Nuclear Information System (INIS)

    Sears, D.F.; Herbert, L.N.; Vaillancourt, K.D.

    1991-01-01

    This paper reviews the status of the LEU conversion program and the progress made in the fuel development program over the last year. The results from post-irradiation examinations of prototype NRU fuel rods containing Al-U 3 Si dispersion fuel, and of mini-elements containing Al-U 3 Si 2 dispersion fuel, are presented. (orig.)

  8. Development of multilayer perceptron networks for isothermal time temperature transformation prediction of U-Mo-X alloys

    Energy Technology Data Exchange (ETDEWEB)

    Johns, Jesse M., E-mail: jesse.johns@pnnl.gov; Burkes, Douglas, E-mail: douglas.burkes@pnnl.gov

    2017-07-15

    In this work, a multilayered perceptron (MLP) network is used to develop predictive isothermal time-temperature-transformation (TTT) models covering a range of U-Mo binary and ternary alloys. The selected ternary alloys for model development are U-Mo-Ru, U-Mo-Nb, U-Mo-Zr, U-Mo-Cr, and U-Mo-Re. These model's ability to predict 'novel' U-Mo alloys is shown quite well despite the discrepancies between literature sources for similar alloys which likely arise from different thermal-mechanical processing conditions. These models are developed with the primary purpose of informing experimental decisions. Additional experimental insight is necessary in order to reduce the number of experiments required to isolate ideal alloys. These models allow test planners to evaluate areas of experimental interest; once initial tests are conducted, the model can be updated and further improve follow-on testing decisions. The model also improves analysis capabilities by reducing the number of data points necessary from any particular test. For example, if one or two isotherms are measured during a test, the model can construct the rest of the TTT curve over a wide range of temperature and time. This modeling capability reduces the cost of experiments while also improving the value of the results from the tests. The reduced costs could result in improved material characterization and therefore improved fundamental understanding of TTT dynamics. As additional understanding of phenomena driving TTTs is acquired, this type of MLP model can be used to populate unknowns (such as material impurity and other thermal mechanical properties) from past literature sources.

  9. PENGARUH SERBUK U-Mo HASIL PROSES MEKANIK DAN HYDRIDE – DEHYDRIDE – GRINDING MILL TERHADAP KUALITAS PELAT ELEMEN BAKAR U-Mo/Al

    Directory of Open Access Journals (Sweden)

    Supardjo Supardjo

    2015-07-01

    Full Text Available PENGARUH SERBUK U-Mo HASIL PROSES MEKANIK DAN HYDRIDE – DEHYDRIDE – GRINDING MILL TERHADAP KUALITAS PELAT ELEMEN BAKAR U-Mo/Al. Penelitian bahan bakar U-7Mo/Al tipe pelat dilakukan dalam rangka pengembangan bahan bakar U3Si2/Al untuk mendapatkan bahan bakar baru yang memiliki densitas uranium lebih tinggi, stabil selama digunakan sebagai bahan bakar di dalam reaktor dan mudah dilakukan proses olah ulangnya. Lingkup penelitian meliputi pembuatan: paduan U-7Mo dengan teknik peleburan, pembuatan serbuk U-7Mo dengan dikikir dan hydride - dehydride - grinding mill, IEB U-7Mo/Al dengan teknik kompaksi pada tekanan 20 bar, dan PEB U-7Mo/Al dengan teknik pengerolan panas pada temperatur 425oC. Paduan U-7Mo hasil proses peleburan cukup homogen, berat jenis 16,34 g/cm3 dan bersifat ulet, kemudian dibuat menjadi serbuk dengan cara dikikir dan hydride - dehydride - grinding mill. Serbuk U-7Mo hasil proses kikir berbentuk pipih, kontaminan Fe cukup tinggi, sedangkan serbuk hasil proses hydride - dehydride - grinding mill, cenderung equiaxial dengan kontaminan yang rendah. Kedua jenis serbuk U-7Mo tersebut digunakan sebagai bahan baku pembuatan IEB U-7Mo/Al dan PEB U-7Mo/Al dengan densitas uranium 7 gU/cm3 dan diperoleh produk dengan kualitas yang hampir sama. Hasil uji IEB U-7Mo/Al berukuran 25 x 15 x 3,15±0,05 mm, tidak terdapat cacat/retak, distribusi U-7Mo di dalam matriks cukup homogen dan tidak terdapat pengelompokan/aglomerasi U-7Mo yang berdimensi >1 mm. PEB U-7Mo/Al hasil pengerolan dengan tebal akhir 1,45 mm, memiliki ketebalan meat rerata 0,60 mm dan tebal kelongsong 0,4 mm dan terdapat 1 titik pengukuran kelongsong dengan ketebalan 0,15 mm. Dengan membandingkan penggunaan kedua jenis serbuk U-7Mo tersebut, IEB U-7Mo/Al dan PEB U-7Mo/Al yang dihasilkan memiliki kualitas hampir sama. Namun demikian penggunaan serbuk U- 7Mo hasil proses hydride - dehydride - grinding mill lebih baik karena proses pengerjaannya lebih cepat dan impuritas dalam

  10. Characterization of a U-Mo alloy subjected to direct hydriding of the gamma phase

    International Nuclear Information System (INIS)

    Balart, Silvia N.; Bruzzoni, Pablo; Granovsky, Marta S.

    2003-01-01

    The Reduced Enrichment for Research and Test Reactors (RERTR) program has imposed the need to develop plate-type fuel elements based on high density uranium compounds, such as U-Mo alloys. One of the steps in the fabrication of the fuel elements is the pulverization of the fissile material. In the case of the U-Mo alloys, the pulverization can be accomplished through hydriding - dehydriding. Two alternative methods of the hydriding-dehydriding process, namely the selective hydriding in alpha phase (HS-alpha) and the massive hydriding in gamma phase (HM-gamma) are currently being studied at the Comision Nacional de Energia Atomica. The HM-gamma method was reproduced at laboratory scale starting from a U-7 wt % Mo alloy. The hydrided and dehydrided materials were characterized using metallographic techniques, scanning electron microscopy, energy dispersive X-ray analysis and X-ray diffraction. These results are compared with previous results of the HS-alpha method. (author)

  11. Numerical aspects of U-Mo core covered by Zry-4 miniplates co-rolling

    International Nuclear Information System (INIS)

    Picchetti, B.; Moscarda, M.V.; Taboada, H.

    2013-01-01

    The aim of this work is to support through adequate modeling the development of the co-rolling process of miniplates and plates starting with compacts including a monolithic U-Mo core with Zry-4 frame and cladding, Through relevant parameter identification and specific variables calculation a co rolling process model was set. The goal is to design a co-rolling optimal strategy related to the expected results through the use of such model. To that end the rolling process is depicted and some elements of strain stress theory on metals are employed. Plastic strain depends on deviator components of the stress tensor but no on the hydrostatic one. Metal sheet co-rolling is a plastic strain by planar compression at constant volume. During the co-rolling process the width constancy is assumed, being the piece of metal free to flow along its length. In this work the relationship between constitutive materials shield stresses U-Mo core and Zry-4 cladding under T= 650°C co-rolling is determined. This allows to modeling the reduction that exist in each co-rolling step for each one of phases present, which enables the design of a loop control lace optimizing the co rolling process. (author)

  12. Isothermal section of diagram of U-Mo-B and U-Re-B systems

    International Nuclear Information System (INIS)

    Val'ovka, I.P.; Kuz'ma, Yu.B.

    1986-01-01

    The methods of X-ray analysis are used to study the U-Mo-B and U-Re-B systems and to plot phase equilibrium diagrams at 1000 and 800 deg C, respectively. A formation of boride UMoB 4 (structure of the ThMoB 4 type) is confirmed in the U-Mo-B system and new compounds are found: U 2 MoB 6 (rhombic structure of the Y 2 ReB 6 type, a=0.9301(9), b=1.1434(11), c=0.3678(4) nm), ∼UMo 2 B 6 and ∼ UMo 4 B 4 with unknown structures. In the U-Re-B system besides previously known boride UReB 4 (the ThMoB 4 structure type), new ones are obtained: U 2 ReB 6 (Y 2 ReB 6 type, a=0.9373(9), b=1.1529(13), c0.3653(4) nm) and UReB 3 (hexagonal structure of the proper type, a=0.5083(1), c=0.5095(1) nm)

  13. Neutronic and thermal-hydraulic analysis of a device for irradiation of LEU UAl{sub x}-Al targets for {sup 99}Mo production in the IEA-R1 reactor; Analises neutronica e termo-hidraulica de um dispositivo para irradiacao de alvos tipo LEU de UAl{sub x}-Al para producao de {sup 99}MO no reator IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Nishiyama, Pedro Julio Batista de Oliveira

    2012-07-01

    Technetium-99m ({sup 99m}Tc), the product of radioactive decay of molybdenum-99 ( Mo), is one of the most widely used radioisotope in nuclear medicine, covering approximately 80% of all radiodiagnosis procedures in the world. Nowadays, Brazil requires an amount of about 450 Ci of {sup 99}Mo per week. Due to the crisis and the shortage of {sup 99}Mo supply chain that has been observed on the world since 2008, IPEN/CNEN-SP decided to develop a project to produce {sup 99}Mo through fission of uranium-235. The objective of this dissertation was the development of neutronic and thermal-hydraulic calculations to evaluate the operational safety of a device for {sup 99}Mo production to be irradiated in the IEA-Rl reactor core at 5 MW. In this device will be placed ten targets of UAl{sub x}-Al dispersion fuel with low enriched uranium (LEU) and density of 2.889 gU/cm{sup 3}. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION and CITATION and the maximum temperatures reached in the targets were calculated with the code MTRCR-IEA-R1. The analysis demonstrated that the device irradiation will occur without adverse consequences to the operation of the reactor. The total amount of {sup 99}Mo was calculated with the program SCALE and considering that the time needed for the chemical processing and recovering of the {sup 99}Mo will be five days after the irradiation, we have that the {sup 99}Mo activity available for distribution will be 176 Ci for 3 days of irradiation, 236 Ci for 5 days of irradiation and 272 Ci for 7 days of targets irradiation. (author)'.

  14. Interim Report on Mixing During the Casting of LEU-10Mo Plates in the Triple Plate Molds

    Energy Technology Data Exchange (ETDEWEB)

    Aikin, Jr., Robert M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-04-12

    LEU-10%Mo castings are commonly produced by down blending unalloyed HEU with a DU-12.7%Mo master-alloy. This work uses process modeling to provide insight into the mixing of the unalloyed uranium and U-Mo master alloy during melting and mold filling of a triple plate casting. Two different sets of situations are considered: (1) mixing during mold filling from a compositionally stratified crucible and (2) convective mixing of a compositionally stratified crucible during mold heating. The mold filling simulations are performed on the original Y-12 triple plate mold and the horizontal triple plate mold.

  15. PIE Report on the KOMO-3 Irradiation Test Fuels

    International Nuclear Information System (INIS)

    Park, Jong Man; Ryu, H. J.; Yang, J. H.

    2009-04-01

    In the KOMO-3, in-reactor irradiation test had been performed for 12 kinds of dispersed U-Mo fuel rods, a multi wire fuel rod and a tube fuel rod. In this report we described the PIE results on the KOMO-3 irradiation test fuels. The interaction layer thickness between fuel particle and matrix could be reduced by using a large size U-Mo fuel particle or introducing Al-Si matrix or adding the third element in the U-Mo particle. Monolithic fuel rod of multi-wire or tube fuel was also effective in reducing the interaction layer thickness

  16. Nitride Coating Effect on Oxidation Behavior of Centrifugally Atomized U-Mo Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Jin; Cho, Woo Hyoung; Park, Jong Man; Lee, Yoon Sang; Yang, Jae Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    Uranium metal and uranium compounds are being used as nuclear fuel materials and generally known as pyrophoric materials. Nowadays the importance of nuclear fuel about safety is being emphasized due to the vigorous exchanges and co-operations among the international community. According to the reduced enrichment for research and test reactors (RERTR) program, the international research reactor community has decided to use low-enriched uranium instead of high-enriched uranium. As a part of the RERTR program, KAERI has developed centrifugally atomized U-Mo alloys as a promising candidate of research reactor fuel. Kang et al. studied the oxidation behavior of centrifugally atomized U-10wt% Mo alloy and it showed better oxidation resistance than uranium. In this study, the oxidation behavior of nitride coated U-7wt% Mo alloy is investigated to enhance the safety against pyrophoricity

  17. Scaling up the production capacity of U-Mo powder by HMD process

    International Nuclear Information System (INIS)

    Pasqualini, E.E.; Lopez, M.; Helzel Garcia, L.J.; Echenique, P.; Adelfang, P.

    2002-01-01

    The recent discovery that uranium alloys in metastable gamma phase can be hydrided at low temperatures and pressures have allowed developing the method of commuting bulk materials by milling the hydride to desired size and then dehydriding the powder. This process is called HMD (hydriding-milling-dehydriding) and needs an initial step of hydrogen incorporation to allow the alloy to be hydrided. This four step process has been conveniently set up for the production of U-7Mo powder for its use in nuclear fuels. Low equipment investment and low man power are needed for this achievement. The process is being analyzed in its scaling up for one kilogram batches and a 50 kilogram per year production capacity of U-Mo powder. (author)

  18. 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)

  19. Prediction of U-Mo dispersion nuclear fuels with Al-Si alloy using artificial neural network

    International Nuclear Information System (INIS)

    Susmikanti, Mike; Sulistyo, Jos

    2014-01-01

    Dispersion nuclear fuels, consisting of U-Mo particles dispersed in an Al-Si matrix, are being developed as fuel for research reactors. The equilibrium relationship for a mixture component can be expressed in the phase diagram. It is important to analyze whether a mixture component is in equilibrium phase or another phase. The purpose of this research it is needed to built the model of the phase diagram, so the mixture component is in the stable or melting condition. Artificial neural network (ANN) is a modeling tool for processes involving multivariable non-linear relationships. The objective of the present work is to develop code based on artificial neural network models of system equilibrium relationship of U-Mo in Al-Si matrix. This model can be used for prediction of type of resulting mixture, and whether the point is on the equilibrium phase or in another phase region. The equilibrium model data for prediction and modeling generated from experimentally data. The artificial neural network with resilient backpropagation method was chosen to predict the dispersion of nuclear fuels U-Mo in Al-Si matrix. This developed code was built with some function in MATLAB. For simulations using ANN, the Levenberg-Marquardt method was also used for optimization. The artificial neural network is able to predict the equilibrium phase or in the phase region. The develop code based on artificial neural network models was built, for analyze equilibrium relationship of U-Mo in Al-Si matrix

  20. LEU WWR-M2 fuel assemblies burnable test

    International Nuclear Information System (INIS)

    Kirsanov, G.A.; Konoplev, K.A.; Pikulik, R.G.; Sajkov, Yu. P.; Tchmshkyan, D.V.; Tedoradze, L.V.; Zakharov, A.S.

    2000-01-01

    The results of in-pile irradiation tests of LEU WWR-M2 fuel assemblies with reduced enrichment of fuel are submitted in the report. The tests are made according to the Russian Program on Reduced Enrichment for Research and Test Reactors (RERTR). United States Department of Energy and the Ministry of Atomic Energy of Russian Federation jointly fund this Program. The irradiation tests of 5 WWR-M2 experimental assemblies are carried out at WWR-M reactor of the Petersburg Nuclear Physics Institute (PNPI). The information on assembly design and technique of irradiation tests is presented. In the irradiation tests the integrity of fuel assemblies is periodically measured. The report presents the data for the integrity maintained during the burnup of 5 fuel assemblies up to 45%. These results demonstrate the high reliability of the experimental fuel assemblies within the guaranteed burnup limits specified by the manufacturer. The tests are still in progress; it is planned to test and analyze the change in integrity for burnup of up to 70% - 75% or more. LEU WWR-M2 fuel assemblies are to be offered for export by their Novosibirsk manufacturer. Currently, HEU WWR-M2 fuel assemblies are used in Hungary, Ukraine and Vietnam. LEU WWR-M2 fuel assemblies were designed as a possible replacement for the HEU WWR-M2 fuel assemblies in those countries, but their use can be extended to other research reactors. (author)

  1. 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)

  2. LEU fuel cycle analyses for the Belgian BR2 Research Reactor

    International Nuclear Information System (INIS)

    Deen, J.R.; Snelgrove, J.L.

    1988-01-01

    Equilibrium fuel cycle characteristics were calculated for reference HEU and two proposed LEU fuel cycles using an 11-group diffusion-theory neutron flux solution in hexagonal-Z geometry. The diffusion theory model was benchmarked with a detailed Monte Carlo core model. The two proposed LEU fuel designs increased the 235 U loading 20% and the fuel meat volume 51%. The first LEU design used 10 B as a burnable absorber. Either proposed LEU fuel element would provide equilibrium fuel cycle characteristics similar to those of the HEU fuel cycle. Irradiation rates of Co control followers and Ir disks in the center of the core were reduced 6 ± 1% in the LEU equilibrium core compared to reference HEU core. 11 refs., 4 figs., 5 tabs

  3. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

    Energy Technology Data Exchange (ETDEWEB)

    Collette, R. [Colorado School of Mines, Nuclear Science and Engineering Program, 1500 Illinois St, Golden, CO 80401 (United States); King, J., E-mail: kingjc@mines.edu [Colorado School of Mines, Nuclear Science and Engineering Program, 1500 Illinois St, Golden, CO 80401 (United States); Buesch, C. [Oregon State University, 1500 SW Jefferson St., Corvallis, OR 97331 (United States); Keiser, D.D.; Williams, W.; Miller, B.D.; Schulthess, J. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

    2016-07-15

    The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends when comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. The results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program. - Highlights: • Automated image processing is used to extract fission gas bubble data from irradiated U−Mo fuel samples. • Verification and validation tests are performed to ensure the algorithm's accuracy. • Fission bubble parameters are predictably difficult to compare across samples of varying compositions. • The 2-D results suggest the need for more homogenized fuel sampling in future studies. • The results also demonstrate the value of 3-D reconstruction techniques.

  4. U-Mo Alloy Powder Obtained Through Selective Hydriding. Particle Size Control

    International Nuclear Information System (INIS)

    Balart, S.N.; Bruzzoni, P.; Granovsky, M.S.

    2002-01-01

    Hydride-dehydride methods to obtain U-Mo alloy powder for high-density fuel elements have been successfully tested by different authors. One of these methods is the selective hydriding of the α phase (HSα). In the HSα method, a key step is the partial decomposition of the γ phase (retained by quenching) to α phase and an enriched γ phase or U 2 Mo. This transformation starts mainly at grain boundaries. Subsequent hydrogenation of this material leads to selective hydriding of the α phase, embrittlement and intergranular fracture. According to this picture, the particle size of the final product should be related to the γ grain size of the starting alloy. The feasibility of controlling the particle size of the product by changing the γ grain size of the starting alloy is currently investigated. In this work an U-7 wt% Mo alloy was subjected to various heat treatments in order to obtain different grain sizes. The results on the powder particle size distribution after applying the HSα method to these samples show that there is a strong correlation between the original γ grain size and the particle size distribution of the powder. (author)

  5. Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

    Directory of Open Access Journals (Sweden)

    Aldawahra Saadou

    2015-06-01

    Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

  6. Evidence for the presence of U-Mo-Al ternary compounds in the U-Mo/Al interaction layer grown by thermal annealing: a coupled micro X-ray diffraction and micro X-ray absorption spectroscopy study

    International Nuclear Information System (INIS)

    Palancher, H.; Martin, P.; Nassif, V.

    2007-01-01

    The systematic presence of the ternary phases U 6 Mo 4 Al 43 and UMo 2 Al 20 is reported in a U-Mo/Al interaction layer grown by thermal annealing. This work shows, therefore, the low Mo solubility in UAl 3 and UAl 4 binary phases; it contradicts the hypothesis of the formation of (U,Mo)Al 3 and (U,Mo)Al 4 solid solutions often admitted in the literature. Using μ-XAS (micro X-ray absorption spectroscopy) at the Mo K edge and μ-XRD (micro X-ray diffraction), the heterogeneity of the interaction layer obtained on a γ-U 0.85 Mo 0.15 /Al diffusion couple has been precisely investigated. The UMo 2 Al 20 phase has been identified at the closest location from the Al side. Moreover, μ-XRD mapping performed on an annealed fuel plate enabled the characterization of the four phases resulting from the γ-U 0.85 Mo 0.15 /Al and (U 2 Mo+α-U)/Al interactions. A strong correlation between the concentrations of UAl 4 and UMo 2 Al 20 and those of UAl 3 and U 6 Mo 4 Al 43 has been shown. (orig.)

  7. A neutronic feasibility study for LEU conversion of the Brookhaven Medical Research Reactor (BMRR).

    Energy Technology Data Exchange (ETDEWEB)

    Hanan, N. A.

    1998-01-14

    A neutronic feasibility study for converting the Brookhaven Medical Research Reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with Brookhaven National Laboratory. Two possible LEU cores were identified that would provide nearly the same neutron flux and spectrum as the present HEU core at irradiation facilities that are used for Boron Neutron Capture Therapy and for animal research. One core has 17 and the other has 18 LEU MTR-type fuel assemblies with uranium densities of 2.5g U/cm{sup 3} or less in the fuel meat. This LEU fuel is fully-qualified for routine use. Thermal hydraulics and safety analyses need to be performed to complete the feasibility study.

  8. 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)

  9. 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)

  10. Status of LEU fuel development and conversion of NRU

    International Nuclear Information System (INIS)

    Sears, D.F.; Herbert, L.N.; Vaillancourt, K.D.

    1989-11-01

    The status of the low-enrichment uranium (LEU) fuel development and NRU conversion program at Chalk River Nuclear Laboratories is reviewed. Construction of a new fuel fabrication facility is essentially completed and installation of LEW fuel manufacturing equipment has begun. The irradiation of 31 prototype Al-61 wt% U 3 Si dispersion fuel rods, approximately one third of a full NRU core, is continuing without incident. Recent post-irradiation examination of spent fuel rods revealed that the prototype LEU fuel achieved the design burnup (80 at%) in excellent condition, confirming that the Al-U 3 Si 2 dispersion fuel to complement out Al-U 3 Si capability. Three full-size NRU rods containing Al-U 3 Si 2 dispersion fuel have been fabricated for a qualification irradiation in NRU. Post-irradiation examinations of mini-elements containing Al-U 3 Si 2 fuel revealed that the U 3 Si 2 behaved similarly to U 3 Si 2 fuel revealed that the U 3 Si 2 particles and the aluminum matrix, and fission gas bubbles up to 10 μm in diameter, could be seen in the particles after 60 at% and 80 at% burnup. The mini-elements contained a variety of silicide particle sizes; however, no significant swelling dependence on particle size distribution was observed

  11. Interdiffusion and reactions between U-Mo and Zr at 650 °C as a function of time

    Science.gov (United States)

    Park, Y.; Keiser, D. D.; Sohn, Y. H.

    2015-01-01

    Development of monolithic U-Mo alloy fuel (typically U-10 wt.%Mo) for the Reduced Enrichment for Research and Test Reactors (RERTR) program entails a use of Zr diffusion barrier to eliminate the interdiffusion-reactions between the fuel alloy and Al-alloy cladding. The application of Zr barrier to the U-Mo fuel system requires a co-rolling process that utilizes a soaking temperature of 650 °C, which represents the highest temperature the fuel system is exposed to during both fuel manufacturing and reactor application. Therefore, in this study, development of phase constituents, microstructure and diffusion kinetics of U-10 wt.%Mo and Zr was examined using solid-to-solid diffusion couples annealed at 650 °C for 240, 480 and 720 h. Phase constituents and microstructural development were analyzed by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). Concentration profiles were mapped as diffusion paths on the isothermal ternary phase diagram. Within the diffusion zone, single-phase layers of β-Zr and β-U were observed along with a discontinuous layer of Mo2Zr between the β-Zr and β-U layers. In the vicinity of Mo2Zr phase, islands of α-Zr phases were also found. In addition, acicular α-Zr and U6Zr3Mo phases were observed within the γ-U(Mo) terminal alloy. Growth rate of the interdiffusion-reaction zone was determined to be 7.75 (± 5.84) × 10-16 m2/s at 650 °C, however with an assumption of a certain incubation period.

  12. Fuel conversion of JRR-4 from HEU to LEU

    International Nuclear Information System (INIS)

    Ichikawa, Hiroki; Nakajima, Teruo

    1997-01-01

    Japanese JRR-4 (Japan Research Reactor No.4) is a pool type, light water moderated and cooled, ETR type fuel reactor used for Shielding experiments, isotope production, neutron activation analyses, Si doping, reactor students training. It acieved first criticality on January 28, 1965 with maximum thermal power 3.5MW. The standard core consistes of 20 Fuel elements, 7 control rods 5 Irradiation holes, neutron source, graphite reflectors. Available thermal flux is 7x1013 n/cm2/s. Within the RERTR program plans are made for core conversion from HEU to LEU

  13. The Fabrication Technology Development of Uniform U and U-Mo Foil by Twin Roll Casting

    International Nuclear Information System (INIS)

    Kim, C. K.; Kim, K. H.; Lee, Y. S.; Woo, Y. M.; Kim, J. D.; Oh, J. M.; Sim, M. S.

    2012-01-01

    Uranium foil samples, of which the technology was developed by KAERI around 2000, were distributed to 6 countries including USA in connection with CRP of IAEA. A problem of thickness irregularity was issued so that cold work was done on it. Due to the pin hole and preferred orientation occurrence an additional development project was raised. It was presumed that the irregularity would be influenced by the eddy flow of the melt. So the melt feeding system was changed from pressurized melt flow to gravity-forced flow for more stable melt flow. And then It was tried that the bulgies on the foil surface were eliminated by deforming with a pressing roll. To save the production cost the expensive quartz crucible was replaced with a common graphite plugging crucible system with repeatable use. The loss of very expensive LEU material from melt leak of open nozzle in quartz crucible could be excluded. A new foil collection winding system was adopted so that the quickly coming-out foil could be taken without rumpling. The equipment was test-run with Cu as surrogate. Some drawbacks found during test-running were solved by modifying several parts. Cu foils could be produced with optimized conditions successfully. DU metal was also used for test-running the modified equipment and then some related modifications were done. Finally DU foils meeting the requesting specification could be produced. The length was longer than 10 m. The foil thickness ranged from 140 μm to 300 μm. On observation and measurement the thickness homogeneity was evaluated to be improved a little

  14. Reprocessing of LEU silicide fuel at Dounreay

    International Nuclear Information System (INIS)

    Cartwright, P.

    1996-01-01

    UKAEA have recently reprocessed two LEU silicide fuel elements in their MTR fuel reprocessing plant at Dounreay. The reprocessing was undertaken to demonstrate UKAEA's commitment to the world-wide research reactor communities future needs. Reprocessing of LEU silicide fuel is seen as a waste treatment process, resulting in the production of a liquid feed suitable for conditioning in a stable form of disposal. The uranium product from the reprocessing can be used as a blending feed with the HEU to produce LEU for use in the MTR cycle. (author)

  15. Followup calculations for the UVAR LEU conversion

    International Nuclear Information System (INIS)

    Rydin, R.; Hosticka, B.; Burns, T; Hubbard, T.; Mulder, R

    2004-01-01

    The UVAR reactor was successfully converted to LEU fuel in April 1994. Void coefficient measurements were made on the 4- by-4 fully-graphite-reflected LEU-1 core configuration, and an isothermal temperature coefficient measurement was made on the operational 4-by-5 partially-graphite-reflected LEU-2 core configuration. Both of these experiments have now been modeled in their critical configurations using the 3DBUM code. The LEU cores were also modeled using the Monte Carlo code MCNP in order to obtain a neutron/gamma source for BNCT filter design calculations. Advanced BNCT filters have been evaluated using both MCNP and the discrete ordinates code DORT. The results indicate that the UVAR would be an ideal source for the BNCT treatment of brain tumors. (author)

  16. Followup calculations for the UVAR LEU conversion

    International Nuclear Information System (INIS)

    Rydin, R.A.; Hosticka, B.; Burns, T.

    1995-01-01

    The UVAR reactor was successfully converted to LEU fuel in April 1994. Void coefficient measurements were made on the 4-by-4 fully-graphite-reflected LEU-1 core configuration, and an isothermal temperature coefficient measurement was made on the operational 4-by-5 partially-graphite-reflected LEU-2 core configuration. Both of these experiments have now been modeled in their critical configurations using the 3DBUM code. The LEU cores were also modeled using the Monte Carlo code MCNP in order to obtain a neutron/gamma source for BNCT filter design calculations. Advanced BNCT filters have been evaluated using both MCNP and the discrete ordinates code DORT. The results indicate that the UVAR would be an ideal source for the BNCT treatment of brain tumors

  17. 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)

  18. Neutronic performance of high-density LEU fuels in water-moderated and water-reflected research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Matos, J.E.

    1996-01-01

    At the Reduced Enrichment for Research and Test Reactors (RERTR) meeting in September 1994, Durand reported that the maximum uranium loading attainable with U 3 Si 2 fuel is about 6.0 g U/cm 3 . The French Commissariat a l'Energie Atomique (CEA) plan to perform irradiation tests with 5 plates at this loading. Compagnie pour L'Etude et La Realisation de Combustibles Atomiques (CERCA) has also fabricated a few uranium nitride (UN) plates with a uranium density in the fuel meat of 7.0 g/cm 3 and found that UN is compatible with the aluminum matrix at temperatures below 500 C. High density dispersion fuels proposed for development include U-Zr(4 wt%)-Nb(2 wt%), U-Mo(5 wt%), and U-Mo(9 wt%). The purpose of this note is to examine the relative neutronic behavior of these high density fuels in a typical light water-reflected and water-moderated MTR-type research reactor. The results show that a dispersion of the U-Zr-Nb alloy has the most favorable neutronic properties and offers the potential for uranium densities greater than 8.0 g/cm 3 . On the other hand, UN is the least reactive fuel because of the relatively large 14 N(n,p) cross section. For a fixed value of k eff , the required 235 U loading per fuel element is least for the U-Zr-Nb fuel and steadily increases for the U-Mo(5%), U-Mo(9%), and UN fuels. Because of volume fraction limitations, the UO 2 dispersions are only useful for uranium densities below 5.0 g/cm 3 . In this density range, however, UO 2 is more reactive than U 3 Si 2

  19. 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

  20. 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.)

  1. Feasibility study for LEU conversion of the WWR-K reactor at the Institute of Nuclear Physics in Kazakhstan using a 5-tube fuel assembly

    International Nuclear Information System (INIS)

    Hanan, N.A.; Liaw, J.R.; Matos, J.E.

    2005-01-01

    A feasibility study by the RERTR program for possible LEU conversion of the 6 MW WWR-K reactor concludes that conversion is feasible using an LEU 5-tube Russian fuel assembly design. This 5-tube design is one of several LEU fuel assembly designs being studied (Ref. 1) for possible use in this reactor. The 5-tube assembly contains 200 g 235 U with an enrichment of 19.7% in four cylindrical inner tubes and an outer hexagonal tube with the same external dimensions as the current HEU (36%) 5-tube fuel assembly, which contains 112.5 g 235 U. The fuel meat material, LEU UO 2 -Al dispersion fuel with ∼ 2.5 g U/cm 3 , has been extensively irradiation tested in a number of reactors with uranium enrichments of 36% and 19.7%. Since the 235 U loading of the LEU assemblies is much larger than the HEU assemblies, a smaller LEU core with five rows of fuel assemblies is possible (instead of six rows of fuel assemblies in the HEU core). This smaller LEU core would consume about 60% as many fuel assemblies per year as the current HEU core and provide thermal neutron fluxes in the inner irradiation channels that are ∼ 17% larger than with the present HEU core. The current 21 day cycle length would be maintained and the average discharge burnup would be ∼ 42%. Neutron fluxes in the five outer irradiation channels would be smaller in the LEU core unless these channels can be moved closer to the LEU fuel assemblies. Results show that the smaller LEU core would meet the reactor's shutdown margin requirements and would have an adequate thermal-hydraulic safety margin to onset of nucleate boiling. (author)

  2. SMOPY, a new NDA tool for safeguards of LEU and MOX spent fuel

    International Nuclear Information System (INIS)

    Lebrun, A.; Merelli, M.; Szabo, J.-L.; Huver, M.; Arenas-Carrasco, J.

    2001-01-01

    Upon IAEA request, the French support program to IAEA Safeguards has developed a new device for control of the irradiated LEU and MOX fuels. The Safeguards Mox Python (SMOPY) is the achievement of a 4 years R and D program supported by CEA and COGEMA in partnership with Eurisys Mesures. The SMOPY system is based on the combination of 2 NDA techniques (passive neutron and room temperature gamma spectrometry) and on line interpretation tools (automatic gamma spectrum interpretation, depletion code EVO). Through the measurement managing software, all this contributes to the fully automatic measurement, interpretation and characterization of any kind of spent fuel. The device is transportable (50 kg, 60 cm) and is composed of four parts: 1. the measurement head with one high efficiency fission chamber and a micro room temperature gamma spectrometric probe; 2. the carrier which carries the measurement head. The carrier bottom fits the racks for accurate positioning and its top fits operator's fuel moving tool; 3. the portable electronic cabinet which includes both neutron and gamma electronic cards; 4. the portable PC which gets inspectors data, controls the measurement, get measured values, interprets them and immediately provides the inspector with worthwhile info for appropriate on the field decisions. Main features of SMOPY are: Discrimination of MOX versus LEU irradiated fuels in any case (conservative case is one cycle MOX versus three cycles LEU after short cooling time); Full characterization of irradiated LEU (burnup, cooling time, Pu amounts ...); Partial Defect Test on LEU fuels. A first version of SMOPY has been tested in industrial condition during summer 2000. This tests shown a need of shielding improvement around the gamma detector. A new version has been build a will be qualified during a new field test and then the system will be ready for routine operation in IAEA and commercial delivery. After giving details about the system itself, this paper

  3. Evaluation of AS-CAST U-Mo alloys processed in graphite crucible coated with boron nitride

    Energy Technology Data Exchange (ETDEWEB)

    Marra, Kleiner M., E-mail: kleiner.marra@prof.una.br [Centro Universitario UNA, Belo Horizonte, MG (Brazil). Curso de Engenharia Mecânica; Reis, Sérgio C.; Paula, João B. de; Pedrosa, Tércio A., E-mail: reissc@cdtn.br, E-mail: jbp@cdtn.br, E-mail: tap@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    This paper reports the production of uranium-molybdenum alloys, which have been considered promising fuel for test and research nuclear reactors. U-Mo alloys were produced in three molybdenum contents: 5w%, 7w%, and 10w%, using an electric vacuum induction furnace. A boron nitride-coated graphite crucible was employed in the production of the alloys and, after melting, the material was immediately poured into a boron nitride-coated graphite mold. The incorporation of carbon was observed, but it happened in a lower intensity than in the case of the non-coated crucible/mold. It is observed that the carbon incorporation increased and alloys density decreased with Mo addition. It was also noticed that the increase in the carbon or molybdenum content did not seem to change the as-cast structure in terms of granulation. The three alloys presented body-centered cubic crystal structure (γ-phase), after solidification, besides a seeming negative microsegregation of molybdenum, from the center to the periphery of the grains. There were signs of macrosegregation, from the base to the top of the ingots. (author)

  4. 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.

  5. Full core operation in JRR-3 with LEU fuels

    International Nuclear Information System (INIS)

    Murayama, Y.; Issiki, M.

    1995-01-01

    The new JRR-3 a 20MWT swimming pool type research reactor, is made up of plate type LEU fuel elements with U-Al x fuel at 2.2 gU/cm 3 . Reconstruction work for the new JR-3 was a good success, and common operation started in November 1990, and 7 cycles (26 days operation/cycle) have passed. We have no experience in using such a high uranium density fuel element with aluminide fuel. So we plan to examine the condition of the irradiated fuel elements with three methods, that is, measurement of the value of FFD in operation, observation of external view of the fuels in refueling work and postirradiation examination after maximum burn-up will be established. In the results of the first two methods, the fuel elements of JRR-3 is burned up normally and have no evidence of failure. (author)

  6. The French UMo group contribution to new LEU fuel development

    International Nuclear Information System (INIS)

    Hamy, J.M.; Lemoine, P.; Huet, F.; Jarousse, C.; Emin, J.L.

    2005-01-01

    The French UMo Group was based on a close collaboration between CEA and AREVA's companies strongly involved in the MTR field. The aim of this program was to deliver industrially a high performance LEU UMo fuel able to be reprocessed, and suitable for a wide range of Research Reactor, covering the expected needs for MTR next generation. Since 1999, the program has been focused on industrial aspects with the intention to deal with the whole fuel cycle: manufacturing, irradiation behaviour, fuel characterisation, code development and reprocessing validation. It has been based on the fabrication of full-sized U-7%Mo fuel plates with a density up to 8 gU/cm 3 . The dedicated and advanced R and D means provided by the CEA have been used intensively with the contribution of HFR and BR2 facilities in Europe. This paper presents a synthesis of the program and the corresponding significant results obtained. These results have played a major role as regards the UMo dispersion fuel qualification route by issuing, for the first time, evidence of severe performance limitations. Consequently, the global international effort to develop and qualify a high density LEU UMo fuel has been definitively re-routed and forced to overcome these discrepancies by exploring new technical solutions. A French extended program sustained by a CEA and CERCA collaboration has been launched in 2004 in order to develop a suitable UMo fuel solution. UMo dispersion and monolithic fuel are both investigated through three new full-sized plate irradiations planned in OSIRIS. (author)

  7. Neutronic and thermal-hydraulic analysis of devices for irradiation of LEU targets type of UALx-Al and U-Ni to production of 99Mo in reactor IEA-R1 and RMB

    International Nuclear Information System (INIS)

    Domingos, Douglas Borges

    2014-01-01

    In this work neutronic and thermal-hydraulic analyses were made to compare three types of targets (UAl 2 -Al, U-Ni cylindrical and U-Ni plate) used for the production of 99 Mo by fission of 235 U. Some experiments were conducted to validate the neutronic and thermal-hydraulics methodologies used in this work. For the neutronic calculations the computational programs NJOY99.0, AMPX-II and HAMMERTECHNION were used to generate the cross sections. SCALE 6.0 and CITATION computational programs were used for three-dimensional calculations of the reactor cores, fuel burning and the production of 99 Mo. The computational programs MTRCR-IEAR1 and ANSYS CFX were used to calculate the thermal and hydraulic parameters of the irradiation devices and for comparing them to limits and design criteria. First were performed neutronic and thermal-hydraulic analyzes for the reactor IEA-R1 with the targets of UAl 2 -Al (10 mini plates). Analyses have shown that the total activity obtained for 99 Mo on the mini plates does not meet the demand of Brazilian hospitals (450 Ci/week) and that no limit of thermo-hydraulic design is overtaken. Next, the same calculations were performed for the three target types in Multipurpose Brazilian Reactor (MBR). The neutronic analyzes demonstrated that the three targets meet the demand of Brazilian hospitals. The thermal hydraulic analysis shows that a minimum speed of 7 m/s for the target UAl 2 -Al, 8 m/s for the cylindrical target U-Ni and 9 m/s for the target U-Ni plate will be necessary in the irradiation device to not exceed the design limits. Were performed experiments using a test bench for validate the methodologies for the thermal-hydraulic calculation. The experiments performed to validate the neutronic calculations were made in the reactor IPEN/MB-01. All experiments were simulated with the methodologies described above and the results compared. The simulations results showed good agreement with experimental results. (author)

  8. Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W.; Ross Finlay, M.; Moore, Glenn; Medvedev, Pavel; Meyer, Mitch

    2017-05-01

    The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

  9. HEU to LEU fuel conversion. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.

    1994-10-01

    The Nuclear Regulatory Commission issued a ruling, effective March 27, 1986, that all U.S. non-power reactors convert from HEU fuel to LEU fuel. A Reduced Enrichment for Research and Test Reactors Program was conducted by the Department of Energy at Argonne National Laboratory to coordinate the development of the high density LEU fuel and assist in the development of Safety Analysis Reports for the smaller non-power reactors. Several meetings were held at Argonne in 1987 with the non-power reactor community to discuss the conversion and to set up a conversion schedule for university reactors. EG&G at Idaho was assigned the coordination of the fuel element redesigns. The fuel elements were manufactured by the Babcock & Wilcox Company in Lynchburg, Virginia. The University of Virginia was awarded a grant by the DOE Idaho Operations Office in 1988 to perform safety analysis studies for the LEU conversion for its 2 MW UVAR and 100 Watt CAVALIER reactors. The University subsequently decided to shut down the CAVALIER reactor. A preliminary SAR on the UVAR, along with Technical Specification changes, was submitted to the NRC in November, 1990. An updated SAR was approved by the NRC in January, 1991. In September, 1992, representatives from the fuel manufacturer (B&W) and the fuel designer (EG&G, Idaho) came to the UVAR facility to observe trial fittings of new 22 plate LEU mock fuel elements. B&W fabricated two non-fuel bearing elements, a regular 22 plate element and a control rod element. The elements were checked against the drawings and test fitted in the UVAR grid plate. The dimensions were acceptable and the elements fit in the grid plate with no problems. The staff made several suggestions for minor construction changes to the end pieces on the elements, which were incorporated into the final design of the actual fuel elements. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  10. HEU to LEU fuel conversion. Final report

    International Nuclear Information System (INIS)

    Mulder, R.U.

    1994-10-01

    The Nuclear Regulatory Commission issued a ruling, effective March 27, 1986, that all U.S. non-power reactors convert from HEU fuel to LEU fuel. A Reduced Enrichment for Research and Test Reactors Program was conducted by the Department of Energy at Argonne National Laboratory to coordinate the development of the high density LEU fuel and assist in the development of Safety Analysis Reports for the smaller non-power reactors. Several meetings were held at Argonne in 1987 with the non-power reactor community to discuss the conversion and to set up a conversion schedule for university reactors. EG ampersand G at Idaho was assigned the coordination of the fuel element redesigns. The fuel elements were manufactured by the Babcock ampersand Wilcox Company in Lynchburg, Virginia. The University of Virginia was awarded a grant by the DOE Idaho Operations Office in 1988 to perform safety analysis studies for the LEU conversion for its 2 MW UVAR and 100 Watt CAVALIER reactors. The University subsequently decided to shut down the CAVALIER reactor. A preliminary SAR on the UVAR, along with Technical Specification changes, was submitted to the NRC in November, 1990. An updated SAR was approved by the NRC in January, 1991. In September, 1992, representatives from the fuel manufacturer (B ampersand W) and the fuel designer (EG ampersand G, Idaho) came to the UVAR facility to observe trial fittings of new 22 plate LEU mock fuel elements. B ampersand W fabricated two non-fuel bearing elements, a regular 22 plate element and a control rod element. The elements were checked against the drawings and test fitted in the UVAR grid plate. The dimensions were acceptable and the elements fit in the grid plate with no problems. The staff made several suggestions for minor construction changes to the end pieces on the elements, which were incorporated into the final design of the actual fuel elements. Selected papers are indexed separately for inclusion in the Energy Science and Technology

  11. Accelerating the design and testing of LEU fuel assemblies for conversion of Russian-designed research reactors outside Russia

    International Nuclear Information System (INIS)

    Matos, J.E

    2003-01-01

    This paper identifies proposed geometries and loading specifications of LEU tube-type and pin-type test assemblies that would be suitable for accelerating the conversion of Russian-designed research reactors outside of Russia if these fuels are manufactured, qualified by irradiation testing, and made commercially available in Russia. (author)

  12. [Leu31, Pro34]neuropeptide Y

    DEFF Research Database (Denmark)

    Fuhlendorff, J; Gether, U; Aakerlund, L

    1990-01-01

    Two types of binding sites have previously been described for 36-amino acid neuropeptide Y (NPY), called Y1 and Y2 receptors. Y2 receptors can bind long C-terminal fragments of NPY-e.g., NPY-(13-36)-peptide. In contrast, Y1 receptors have until now only been characterized as NPY receptors that do...... not bind such fragments. In the present study an NPY analog is presented, [Leu31, Pro34]NPY, which in a series of human neuroblastoma cell lines and on rat PC-12 cells can displace radiolabeled NPY only from cells that express Y1 receptors and not from those expressing Y2 receptors. The radiolabeled analog......, [125I-Tyr36] monoiodo-[Leu31, Pro34]NPY, also binds specifically only to cells with Y1 receptors. The binding of this analog to Y1 receptors on human neuroblastoma cells is associated with a transient increase in cytoplasmic free calcium concentrations similar to the response observed with NPY. [Leu31...

  13. 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)

  14. 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)

  15. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

  16. HEU/LEU-conversion of BER II successfully finished

    International Nuclear Information System (INIS)

    Haas, K.; Fischer, C.-O.; Krohn, H.

    2000-01-01

    The BER II (Berliner Experimental Reactor) research reactor is a swimming pool type reactor located in Berlin, Germany. The reactor operates with a thermal power of 10 MW and is primarily used to produce neutrons for neutron scattering experiments. The conversion from HEU- to LEU-fuel elements began in August, 1997. At the last RERTR Meeting 1999 in Budapest, Hungary, Hahn-Meitner-Institut (HMI) presented a 'Status Report' on the conversion of 10 HEU/LEU mixed cores. In February 2000, HMI finished the HEU/LEU-conversion. Hereby, the first pure LEU-standard-core went into operation. Our second LEU-core just ends its operation at the end of July. The third LEU-core will be built up in the beginning of August. The average burn-up rate was improved from 50 - 55% (HEU) to 60 - 65% (LEU). Therefore, only 14 elements/year are now used instead of 28/year. The following report describes our first steps in building pure LEU-cores from mixed HEU/LEU-cores, as well as our initial experience using the pure LEU-cores. (author)

  17. Recent observations at the post-irradiation examination of low-enriched U-Mo miniplates irradiated to high burn-up

    International Nuclear Information System (INIS)

    Hofman, G.L.; Kim, Y.S.; Finlay, M.R.; Snelgrove, J.L.; Hayes, S.L.; Meyer, M.K.; Clark, C.R.

    2003-01-01

    High-density dispersion fuel experiment, RERTR-4, was removed from the Advanced Test Reactor (ATR) after reaching a peak U-235 burnup of ∼80% and is presently undergoing postirradiation examination at the ANL Alpha-Gamma Hot Cell Facility. This test consists of 32 mini fuel plates of which 27 were fabricated with nominally 6 and 8 g cm -3 atomized and machined uranium alloy powders containing 6.5 wt% to 10 wt% molybdenum. In addition, two miniplates contained solid U-10wt%Mo foils. Recent results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from a companion test performed to 50% burnup, RERTR-5, are presented. (author)

  18. LEU fuel powder technology at Babcock and Wilcox (USA)

    International Nuclear Information System (INIS)

    Bogacik, K.E.

    1984-01-01

    This paper traces BandW involvement in HEU fuel manufacturing to the current work directed at LEU reactor technology. Past work at BandW in areas such as alloying, fuel handling and core manufacturing has been of significant benefit to the current LEU fuel processing requirements. Recent investigations and process developments for production of LEU aluminide and silicide fuels are discussed. Techniques for alloying by vacuum are melting, followed by comminution methods after alloying, are presented for both the LEU aluminide and silicide fuel powders. Powder processing discussions include compacting techniques used by BandW for these alloys. This overview of BandW's LEU i nvolvement provides details of specific modifications and process developments in powdered fuels. Product attributes such as powder chemistry, size, and other physical properties of each LEU fuel are presented. (author)

  19. An alternative LEU design for the FRM-II

    International Nuclear Information System (INIS)

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

    1997-02-01

    The Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm[sup 3] and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime (50 days) and the same neutron flux performance (8 x 10[sup 14] n/cm[sup 2]/s in the reflector). LEU silicide fuel with 4.5 g/cm[sup 3] has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. Computer models for the HEU and LEU designs have been exchanged between TUM and ANL and discrepancies have been resolved. The following issues are addressed: qualification of HEU and LEU silicide fuels, stability of the fuel plates, gamma heating in the heavy water reflector, a hypothetical accident involving the configuration of the reflector, a loss of primary coolant flow transient due to an interrupted power supply, the radiological consequences of larger fission product and plutonium inventories in the LEU core, and cost and schedule. Calculations were also done to address the possibility that new high density LEU fuels could be developed that would allow conversion of the TUM HEU design to LEU fuel. Based on the excellent results for the Alternative LEU Design that were obtained in these analyses, the RERTR Program concludes that all of the major technical issues regarding use of LEU fuel instead of HEU fuel in the FRM-II have been successfully resolved and that it is definitely feasible to use LEU fuel in the FRM-II without compromising the safety or performance of the facility

  20. Fuel element burnup determination in HEU-LEU mixed TRIGA research reactor core

    International Nuclear Information System (INIS)

    Zagar, Tomaz; Ravnik, Matjaz

    2000-01-01

    This paper presents the results of a burnup calculations and burnup measurements for TRIGA FLIP HEU fuel elements and standard TRIGA LEU fuel elements used simultaneously in small TRIGA Mark II research reactor in Ljubljana, Slovenija. The fuel element burnup for approximately 15 years of operation was calculated with two different in house computer codes TRIGAP and TRIGLAV (both codes are available at OECD NEA Data Bank). The calculation is performed in one-dimensional radial geometry in TRIGAP and in two-dimensional (r,φ) geometry in TRIGLAV. Inter-comparison of results shows important influence of in-core water gaps, irradiation channels and mixed rings on burnup calculation accuracy. Burnup of 5 HEU and 27 LEU fuel elements was also measured with reactivity method. Measured and calculated burnup values are inter-compared for these elements (author)

  1. Neutron flux measurement in the central channel (XC-1) of TRIGA 14 MW LEU core

    International Nuclear Information System (INIS)

    BARBOS, D.; BUSUIOC, P.; ROTH, Cs.; PAUNOIU, C.

    2008-01-01

    The TRIGA 14 MW reactor, operated by Institute for Nuclear Research Pitesti, Romania, is a pool type reactor, and has a rectangular shape which holds fuel bundles and is surrounded with beryllium reflectors. Each fuel bundle is composed of 25 nuclear fuel rods. The TRIGA 14 MW reactor was commissioned 28 years ago with HEU fuel rods. The conversion was gradually achieved, starting in February 1992 and completed in March 2006. The full conversion of the 14 MW TRIGA Research Reactor was completed in May 2006 and each step of the conversion was achieved by removal of HEU fuel, replaced by LEU fuel, accompanied by a large set of theoretical evaluation and physical measurements intended to confirm the performances of gradual conversion. After the core full conversion, a program of measurements and comparisons with previous results of core physics and measurements is underway, allowing data acquisition for normal operation, demonstration of safety and economics of the converted core. Neutron flux spectrum measurements in the XC in the XC-1 water 1 water-filled channel were performed using multi multi-foil activation techniques. The neutron spectra and flux are obtained by unfolding from measured reaction rates using SAND II computer code. The integral neutron flux value for LEU core is greater of 13% than for the standard HEU core. Also thermal neutron flux value for converted LEU core is smaller by 0.38% than for the standard HEU core. These differences appear because the foil activation detectors have been irradiated using a pneumatic rabbit having a diameter of 32 mm, whereas foil irradiations in standard HEU core has been performed with a pneumatic rabbit having a diameter of 14 mm, and therefore the neutron spectra in LEU core is less thermalized and the weight of fast neutron is greater

  2. Irradiation behavior of uranium-molybdenum dispersion fuel: Fuel performance data from RERTR-1 and RERTR-2

    International Nuclear Information System (INIS)

    Meyer, M.K.; Clark, C.R.; Hayes, S.L.; Strain, R.V.; Hofman, G.L.; Snelgrove, J.L.; Park, J.M.; Kim, K.H.

    1999-01-01

    This paper presents quantitative data on the irradiation behavior of uranium-molybdenum fuels from the low temperature RERTR-1 and -2 experiments. Fuel swelling measurements of U-Mo fuels at ∼40% and ∼70% burnup are presented. The rate of fuel-matrix interaction layer growth is estimated. Microstructures of fuel in the pre- and postirradiation condition were compared. Based on these data, a qualitative picture of the evolution of the U-Mo fuel microstructure during irradiation has been developed. Estimates of uranium-molybdenum fuel swelling and fuel-matrix interaction under high-power research reactor operating conditions are presented. (author)

  3. 31 CFR 540.308 - Low Enriched Uranium (LEU).

    Science.gov (United States)

    2010-07-01

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

  4. 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.

  5. 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

  6. RERTR-12 Insertion 1 Irradiation Summary Report

    International Nuclear Information System (INIS)

    Perez, D.M.; Lillo, M.A.; Chang, G.S.; Woolstenhulme, N.E.; Roth, G.A.; Wachs, D.M.

    2012-01-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications. RERTR-12 insertion 1 includes the capsules irradiated during the first two irradiation cycles. These capsules include Z, X1, X2 and X3 capsules. The following report summarizes the life of the RERTR-12 insertion 1 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  7. RERTR-12 Insertion 2 Irradiation Summary Report

    International Nuclear Information System (INIS)

    Perez, D.M.; Chang, G.S.; Wachs, D.M.; Roth, G.A.; Woolstenhulme, N.E.

    2012-01-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications.1 RERTR-12 insertion 2 includes the capsules irradiated during the last three irradiation cycles. These capsules include Z, Y1, Y2 and Y3 type capsules. The following report summarizes the life of the RERTR-12 insertion 2 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  8. 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

  9. Status of core conversion with LEU silicide fuel in JRR-4

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Teruo; Ohnishi, Nobuaki; Shirai, Eiji [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-08-01

    Japan Research Reactor No.4 (JRR-4) is a light water moderated and cooled, 93% enriched uranium ETR-type fuel used and swimming pool type reactor with thermal output of 3.5MW. Since the first criticality was achieved on January 28, 1965, JRR-4 has been used for shielding experiments, radioisotope production, neutron activation analyses, training for reactor engineers and so on for about 30 years. Within the framework of the RERTR Program, the works for conversion to LEU fuel are now under way, and neutronic and thermal-hydraulic calculations emphasizing on safety and performance aspects are being carried out. The design and evaluation for the core conversion are based on the Guides for Safety Design and Evaluation of research and testing reactor facilities in Japan. These results show that the JRR-4 will be able to convert to use LEU fuel without any major design change of core and size of fuel element. LEU silicide fuel (19.75%) will be used and maximum neutron flux in irradiation hole would be slightly decreased from present neutron flux value of 7x10{sup 13}(n/cm{sup 2}/s). The conversion works are scheduled to complete in 1998, including with upgrade of the reactor building and utilization facilities.

  10. Status of core conversion with LEU silicide fuel in JRR-4

    International Nuclear Information System (INIS)

    Nakajima, Teruo; Ohnishi, Nobuaki; Shirai, Eiji

    1997-01-01

    Japan Research Reactor No.4 (JRR-4) is a light water moderated and cooled, 93% enriched uranium ETR-type fuel used and swimming pool type reactor with thermal output of 3.5MW. Since the first criticality was achieved on January 28, 1965, JRR-4 has been used for shielding experiments, radioisotope production, neutron activation analyses, training for reactor engineers and so on for about 30 years. Within the framework of the RERTR Program, the works for conversion to LEU fuel are now under way, and neutronic and thermal-hydraulic calculations emphasizing on safety and performance aspects are being carried out. The design and evaluation for the core conversion are based on the Guides for Safety Design and Evaluation of research and testing reactor facilities in Japan. These results show that the JRR-4 will be able to convert to use LEU fuel without any major design change of core and size of fuel element. LEU silicide fuel (19.75%) will be used and maximum neutron flux in irradiation hole would be slightly decreased from present neutron flux value of 7x10 13 (n/cm 2 /s). The conversion works are scheduled to complete in 1998, including with upgrade of the reactor building and utilization facilities

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  12. Metallographic analysis of irradiated RERTR-3 fuel test specimens

    International Nuclear Information System (INIS)

    Meyer, M. K.; Hofman, G. L.; Strain, R. V.; Clark, C. R.; Stuart, J. R.

    2000-01-01

    The RERTR-3 irradiation test was designed to investigate the irradiation behavior of aluminum matrix U-MO alloy dispersion fuels under high-temperature, high-fission-rate conditions. Initial postirradiation examination of RERTR-3 fuel specimens has concentrated on binary U-MO atomized fuels. The rate of matrix aluminum depletion was found to be higher than predictions based on low temperature irradiation data. Wavelength Dispersive X-ray Spectroscopy (WDS) indicates that aluminum is present in the interior of the fuel particles. WDS data is supported by a mass and volume balance calculation performed on the basis of image analysis results. The depletion of matrix aluminum seems to have no detrimental effects on fuel performance under the conditions tested to date

  13. An alternative LEU design for the FRM-II

    International Nuclear Information System (INIS)

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

    1996-01-01

    The Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm 3 and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime (50 days) and the same neutron flux performance. LEU silicide fuel with 4.5 g/cm 3 has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. The following issues raised by TUM were addressed in Ref. 1: qualification of HEU and LEU silicide fuels, gamma heating in the heavy water reflector, radiological consequences of larger fission product and plutonium inventories in the LEU core, and cost and schedule. The conclusions of these analyses are summarized below. This paper addresses three additional safety issues that were raised by TUM in Ref. 2: stability of the involute fuel plates, a hypothetical accident involving the configuration of the reflector, and a loss of primary coolant flow transient due to an interrupted power supply. Based on the excellent results for the Alternative LEU Design that were obtained in these analyses, the RERTR Program concludes that all of the major technical issues regarding use of LEU fuel instead of HEU fuel in the FRM-II have been successfully resolved and that it is definitely feasible to use LEU fuel in the FRM-II without compromising the safety or performance of the facility

  14. Status of HEU-LEU conversion of FRJ-2

    International Nuclear Information System (INIS)

    Damm, G.; Nabbi, R.

    2002-01-01

    The operator of the German FRJ-2 research reactor, 'Research Center Juelich', has participated from the beginning in the RERTR programme and made comprehensive contributions to the test and use of LEU fuel for HEU-LEU-conversion measures. The originally planned time scale for the conversion of FRJ-2 was significantly delayed because of a change of the manufacturer of the LEU fuel elements and a 4 years shutdown of the reactor for refurbishment purposes. In the meantime the new LEU fuel elements are qualified and tested in the reactor. In the moment calculations for the safety report are made and it is planned to apply for the license of FRJ-2 operation with LEU fuel at the beginning of 2003. In order to get most reliable results a sophisticated computational method based on a MCNP model coupled with the depletion code BURN was developed for reactor physical calculations, core conversion studies and fuel element performance analysis and applied to the mixed and LEU core. The licensing schedule and results of latest calculations for the conversion study will be presented. The simulations shows that the thermal flux in the LEU core is about 19% resulting in a lower burnup rate. But in the reflector area around the core and in the center of the cold n source the neutron flux reduction remains limited to 6%. Due to a harder neutron spectrum in the LEU core the kinetic and safety related parameters are slightly reduced. Using the ORIGEN code it could be shown that the increase of the total fission products inventory amounts to about 6% compared to a HEU core. As a consequence of the high amount of U-238, the amount of U-235 in the LEU core has to be about 27% higher than in the HEU core but the U-235 burnup is approx. 5% lower due to the contribution of fissile plutonium. (author)

  15. Radiological consequence analysis with HEU and LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

    1984-01-01

    A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and non-site specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. 10 references, 3 figures, 7 tables.

  16. Preproghrelin Leu72Met polymorphism in obese Korean children.

    Science.gov (United States)

    Jo, Dae-Sun; Kim, Se-Lim; Kim, Sun-Young; Hwang, Pyoung Han; Lee, Kee-Hyoung; Lee, Dae-Yeol

    2005-11-01

    Ghrelin is a novel gut-brain peptide that has somatotropic, orexigenic, and adipogenic effects. We examined the preproghrelin Leu72Met polymorphism in 222 obese Korean children to determine whether it is associated with obesity. The frequencies of the Leu72Met polymorphism were 29.3% in obese, 32.3% in overweight, and 32.5% in lean Korean children. No significant difference was found between Met72 carrier and non-carrier obese children with respect to BMI, total body fat, serum triglycerides, total cholesterol, or LDL-cholesterol levels. Our data suggest that the preproghrelin Leu72Met polymorphism is not associated with obesity in children.

  17. Radiological consequence analysis with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

    1984-01-01

    A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and non-site specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. 10 references, 3 figures, 7 tables

  18. A radiological consequence analysis with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

    1985-01-01

    A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and nonsite specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. (author)

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

    International Nuclear Information System (INIS)

    Malkoske, G.R.

    2003-01-01

    The availability of isotope grade, Highly Enriched Uranium (HEU), from the United States for use in the manufacture of targets for molybdenum-99 production in AECL's NRU research reactor has been a key factor to enable MDS Nordion to develop a reliable, secure supply of medical isotopes for the international nuclear medicine community. The molybdenum extraction process from HEU targets is a proven and established method that has reliably produced medical isotopes for several decades. The HEU process provides predictable, consistent yields for our high-volume, molybdenum-99 production. Other medical isotopes such as I-131 and Xe-133, which play an important role in nuclear medicine applications, are also produced from irradiated HEU targets as a by-product of the molybdenum-99 process. To ensure a continued reliable and timely supply of medical isotopes, MDS Nordion is completing the commissioning of two MAPLE reactors and an associated isotope processing facility (the New Processing Facility). The new MAPLE facilities, which will be dedicated exclusively to medical isotope production, will provide an essential contribution to a secure, robust global healthcare system. Design and construction of these facilities has been based on a life cycle management philosophy for the isotope production process. This includes target irradiation, isotope extraction and waste management. The MAPLE reactors will operate with Low Enriched Uranium (LEU) fuel, a significant contribution to the objectives of the RERTR program. The design of the isotope production process in the MAPLE facilities is based on an established process - extraction of isotopes from HEU target material. This is a proven technology that has been demonstrated over more than three decades of operation. However, in support of the RERTR program and in compliance with U.S. legislation, MDS Nordion has undertaken a LEU Target Development and Conversion Program for the MAPLE facilities. This paper will provide an

  20. Fuel cycle cost study with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Matos, J.E.; Freese, K.E.

    1984-01-01

    Fuel cycle costs are compared for a range of 235 U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors

  1. A fuel cycle cost study with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Matos, J.E.; Freese, K.E.

    1985-01-01

    Fuel cycle costs are compared for a range of 235 U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors. (author)

  2. A fuel cycle cost study with HEU and LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Matos, J E; Freese, K E [Argonne National Laboratory, Argonne, IL (United States)

    1985-07-01

    Fuel cycle costs are compared for a range of {sup 235}U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors. (author)

  3. Neutronic and thermal-hydraulic analysis of devices for irradiation of LEU targets type of UAL{sub x}-Al and U-Ni to production of {sup 99}Mo in reactor IEA-R1 and RMB; Analises neutronicas e termo-hidraulica de dispositivos para irradiacao de alvos tipo LEU de UAL{sub x}-Al e U-Ni para producao de {sup 99}Mo nos reatores IEA-R1 e RMB

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas Borges

    2014-07-01

    In this work neutronic and thermal-hydraulic analyses were made to compare three types of targets (UAl{sub 2}-Al, U-Ni cylindrical and U-Ni plate) used for the production of {sup 99}Mo by fission of {sup 235}U. Some experiments were conducted to validate the neutronic and thermal-hydraulics methodologies used in this work. For the neutronic calculations the computational programs NJOY99.0, AMPX-II and HAMMERTECHNION were used to generate the cross sections. SCALE 6.0 and CITATION computational programs were used for three-dimensional calculations of the reactor cores, fuel burning and the production of {sup 99}Mo. The computational programs MTRCR-IEAR1 and ANSYS CFX were used to calculate the thermal and hydraulic parameters of the irradiation devices and for comparing them to limits and design criteria. First were performed neutronic and thermal-hydraulic analyzes for the reactor IEA-R1 with the targets of UAl{sub 2}-Al (10 mini plates). Analyses have shown that the total activity obtained for {sup 99}Mo on the mini plates does not meet the demand of Brazilian hospitals (450 Ci/week) and that no limit of thermo-hydraulic design is overtaken. Next, the same calculations were performed for the three target types in Multipurpose Brazilian Reactor (MBR). The neutronic analyzes demonstrated that the three targets meet the demand of Brazilian hospitals. The thermal hydraulic analysis shows that a minimum speed of 7 m/s for the target UAl{sub 2}-Al, 8 m/s for the cylindrical target U-Ni and 9 m/s for the target U-Ni plate will be necessary in the irradiation device to not exceed the design limits. Were performed experiments using a test bench for validate the methodologies for the thermal-hydraulic calculation. The experiments performed to validate the neutronic calculations were made in the reactor IPEN/MB-01. All experiments were simulated with the methodologies described above and the results compared. The simulations results showed good agreement with experimental

  4. No association of the neuropeptide Y (Leu7Pro) and ghrelin gene (Arg51Gln, Leu72Met, Gln90Leu) single nucleotide polymorphisms with eating disorders.

    Science.gov (United States)

    Kindler, Jochen; Bailer, Ursula; de Zwaan, Martina; Fuchs, Karoline; Leisch, Friedrich; Grün, Bettina; Strnad, Alexandra; Stojanovic, Mirjana; Windisch, Julia; Lennkh-Wolfsberg, Claudia; El-Giamal, Nadja; Sieghart, Werner; Kasper, Siegfried; Aschauer, Harald

    2011-06-01

    Genetic factors likely contribute to the biological vulnerability of eating disorders. Case-control association study on one neuropeptide Y gene (Leu7Pro) polymorphism and three ghrelin gene (Arg51Gln, Leu72Met and Gln90Leu) polymorphisms. 114 eating disorder patients (46 with anorexia nervosa, 30 with bulimia nervosa, 38 with binge eating disorder) and 164 healthy controls were genotyped. No differences were detected between patients and controls for any of the four polymorphisms in allele frequency and genotype distribution (P > 0.05). Allele frequencies and genotypes had no significant influence on body mass index (P > 0.05) in eating disorder patients. Positive findings of former case-control studies of associations between ghrelin gene polymorphisms and eating disorders could not be replicated. Neuropeptide Y gene polymorphisms have not been investigated in eating disorders before.

  5. Future U.S. supply of Mo-99 production through fission based LEU/LEU technology

    International Nuclear Information System (INIS)

    James Welsh; Bigles, C.I.; Alejandro Valderrabano

    2015-01-01

    Coqui RadioPharmaceuticals Corp. (Coqui) has the goal of establishing a medical isotope production facility for securing a continuous domestic supply of the radioisotope molybdenum-99 for U.S. citizens. Coqui will use an LEU/LEU proven and implemented open pool, light-water, 10 MW, reactor design. The facility is being designed with twin reactors for reliability an on-site hot lab chemical processing and a waste conditioning area and a possible generator producing radio-chemistry lab. Coqui identified a 25 acre site adjacent to an existing industrial park in northern central Florida. This land was gifted and transferred to Coqui by the University of Florida Foundation. We are in the process of developing licensing documents related to the facility. The construction permit application for submission to the U.S. Nuclear Regulatory Commission is currently being prepared. Submission is scheduled for mid to late 2015. Community reaction to the proposed development has been positive. We expect to create 220 permanent jobs and we have an anticipated to be operational by 2020. (author)

  6. 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.

  7. 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

  8. Future U.S. supply of Mo-99 production through fission based LEU/LEU technology.

    Science.gov (United States)

    Welsh, James; Bigles, Carmen I; Valderrabano, Alejandro

    Coquí RadioPharmaceuticals Corp. (Coquí) has the goal of establishing a medical isotope production facility for securing a continuous domestic supply of the radioisotope molybdenum-99 for U.S. citizens. Coquí will use an LEU/LEU proven and implemented open pool, light-water, 10 MW, reactor design. The facility is being designed with twin reactors for reliability an on-site hot lab chemical processing and a waste conditioning area and a possible generator producing radio-chemistry lab. Coquí identified a 25 acre site adjacent to an existing industrial park in northern central Florida. This land was gifted and transferred to Coquí by the University of Florida Foundation. We are in the process of developing licensing documents related to the facility. The construction permit application for submission to the U.S. Nuclear Regulatory Commission is currently being prepared. Submission is scheduled for mid to late 2015. Community reaction to the proposed development has been positive. We expect to create 220 permanent jobs and we have an anticipated to be operational by 2020.

  9. LEU fuel fabrication program for the RECH-1 reactor. Status report

    International Nuclear Information System (INIS)

    Chavez, J.C.; Barrera, M.; Jimenez, O.; Lisboa, J.; Marin, J.

    2000-01-01

    In 1995 a 50 LEU U 3 Si 2 fuel elements fabrication program for the RECH-1 research reactor was established at the Comision Chilena de Energia Nuclear, CCHEN. After a fabrication process qualification stage, in 1998, four elements were early delivered to the reactor in order to start an irradiation qualification stage. The irradiation has reached an estimated 10% burn-up and no fabrication problems have been detected up to this burn-up level. During 1999 and up to the first quarter of 2000, 19 fuel elements were produced and 7 fuel elements are expected for the end of 2000. This report presents an updated summary of the main results obtained in this fuel fabrication program. A summary of other activities generated by this program, such as in core follow-up of the four leader fuel elements, ISO 9001 implementation for the fabrication process and a fabrication and qualification optimization planning, is also presented here. (author)

  10. Irradiation performance of uranium-molybdenum alloy dispersion fuels

    International Nuclear Information System (INIS)

    Almeida, Cirila Tacconi de

    2005-01-01

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

  11. 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

  12. 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

  13. The whole-core LEU fuel demonstration in the ORR

    International Nuclear Information System (INIS)

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

    1985-01-01

    A whole-core demonstration of LEU fuel in the ORR is expected to begin during November 1985. Fuel elements will contain U 3 Si 2 at 4.8 Mg U/m 3 and shim rod fuel followers will contain U 3 Si 2 at 3.5 Mg U/m 3 . Fuel fabrication is underway at B and W, CERCA, and NUKEM, with shipments scheduled to commence in October. The primary objectives of the demonstration are to provide data for validation of LEU and mixed-core fuel cycle calculations and to provide a large-scale demonstration of the acceptable performance of production-line U 3 Si 2 fuel elements. It is planned to approach the full LEU core through a series of mixed cores. Measurements to be made include flux distribution, reactivity swing, control rod worth, cycle length, fuel discharge burn-up, gamma heating rate, β eff /l, and isothermal temperature coefficient. Measurements will also be made on fresh LEU and fresh HEU critical configurations. Preliminary safety approval has been received and the final safety assessment is being reviewed. Key issues being addressed in the safety assessment are fuel performance, radiological consequences, margin to burnout and transient behavior. The LEU core is comparable in all safety aspects to the HEU core and the transition core is only marginally worse owing to higher power seeking factors. (author)

  14. 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.

  15. Synthesis of high specific active tritiated Leu-enkephalin in the leucine residue

    Energy Technology Data Exchange (ETDEWEB)

    Baba, S.; Hasegawa, H.; Shinohara, Y. (Tokyo Coll. of Pharmacy (Japan))

    1989-12-01

    Leu-enkephalin labelled with tritium in the Leu residue has been prepared. Synthesis of the precursor peptide, (4,5-dehydroLeu{sup 5}-)Leu-enkephalin, was carried out by solid phase synthesis using Fmoc amino acid derivatives. The peptide was tritiated catalytically yielding {sup 3}H-Leu-enkephalin with a specific radioactivity of 4.39 TBq/mmol. The distribution of tritium label was investigated by reversed-phase high performance liquid chromatography with a synchronized accumulating radioisotope detector following acidic and enzymatic hydrolysis, which confirmed that the tritium label was entirely located at the Leu residue. (author).

  16. Effect of in-pile degradation of the meat thermal conductivity on the maximum temperature of the plate-type U-Mo dispersion fuels

    International Nuclear Information System (INIS)

    Medvedev, Pavel G.

    2009-01-01

    Effect of in-pile degradation of thermal conductivity on the maximum temperature of the plate-type research reactor fuels has been assessed using the steady-state heat conduction equation and assuming convection cooling. It was found that due to very low meat thickness, characteristic for this type of fuel, the effect of thermal conductivity degradation on the maximum fuel temperature is minor. For example, the fuel plate featuring 0.635 mm thick meat operating at heat flux of 600 W/cm2 would experience only a 20 C temperature rise if the meat thermal conductivity degrades from 0.8 W/cm-s to 0.3 W/cm-s. While degradation of meat thermal conductivity in dispersion-type U-Mo fuel can be very substantial due to formation of interaction layer between the particles and the matrix, and development of fission gas filled porosity, this simple analysis demonstrates that this phenomenon is unlikely to significantly affect the temperature-based safety margin of the fuel during normal operation.

  17. Mo-99 production on a LEU solution reactor

    International Nuclear Information System (INIS)

    Brown, R.W.; Thome, L.A.; Khvostionov, V.Y.

    2005-01-01

    A pilot homogenous reactor utilizing LEU has been developed by the Kurchatov Institute in Moscow along with their commercial partner TCI Medical. This solution reactor operates at levels up to 50 kilowatts and has successfully produced high quality Mo-99 and Sr-89. Radiochemical extraction of medical radionuclides from the reactor solution is performed by passing the solution across a series of inorganic sorbents. This reactor has commercial potential for medical radionuclide production using LEU UO 2 SO 4 fuel. Additional development work is needed to optimize multiple 50 kilowatt cores while at the same time, optimizing production efficiency and capital expenditure. (author)

  18. Whole-core LEU fuel demonstration in the ORR

    International Nuclear Information System (INIS)

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

    1985-01-01

    A whole-core demonstration of LEU fuel in the ORR is expected to begin during November 1985. Fuel elements will contain U 3 Si 2 at 4.8 Mg U/m 3 and shim rod fuel followers will contain U 3 Si 2 at 3.5 Mg U/m 3 . Fuel fabrication is underway at B and W, CERCA, and NUKEM, with shipments scheduled to commence in October. The primary objectives of the demonstration are to provide data for validation of LEU and mixed-core fuel cycle calculations and to provide a large-scale demonstration of the acceptable performance of production-line U 3 Si 2 fuel elements. It is planned to approach the full LEU core through a series of mixed cores. Measurements to be made include flux distribution, reactivity swing, control rod worths, cycle length, fuel discharge burnup, gamma heating rates, β/sub eff/l, and isothermal temperature coefficient. Measurements will also be made on fresh LEU and fresh HEU critical configurations. Preliminary safety approval has been received and the final safety assessment is being reviewed

  19. Facility safeguards at an LEU fuel fabrication facility in Japan

    International Nuclear Information System (INIS)

    Kuroi, H.; Osabe, T.

    1984-01-01

    A facility description of a Japanese LEU BWR-type fuel fabrication plant focusing on safeguards viewpoints is presented. Procedures and practices of MC and A plan, measurement program, inventory taking, and the report and record system are described. Procedures and practices of safeguards inspection are discussed and lessons learned from past experiences are reviewed

  20. A mixed core conversion study with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Matos, J.E.; Freese, K.E.

    1985-01-01

    The results of a mixed core study are presented for gradual replacement of HEU fuel with LEU fuel using the IAEA generic 10 MW reactor as an example. The key parameters show that the transition can be accomplished safely and economically. (author)

  1. Mixed core conversion study with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Matos, J.E.; Freese, K.E.

    1984-01-01

    The results of a mixed core study are presented for gradual replacement of HEU fuel with LEU fuel using the IAEA generic 10 MW reactor as an example. The key parameters show that the transition can be accomplished safely and economically

  2. The Environment Shapes the Inner Vestibule of LeuT

    DEFF Research Database (Denmark)

    Sohail, Azmat; Jayaraman, Kumaresan; Venkatesan, Santhoshkannan

    2016-01-01

    Human neurotransmitter transporters are found in the nervous system terminating synaptic signals by rapid removal of neurotransmitter molecules from the synaptic cleft. The homologous transporter LeuT, found in Aquifex aeolicus, was crystallized in different conformations. Here, we investigated t...... showed TM1A movements, consistent with the simulations, confirming a substantially different inward-open conformation in lipid bilayer from that inferred from the crystal structure....... the inward-open state of LeuT. We compared LeuT in membranes and micelles using molecular dynamics simulations and lanthanide-based resonance energy transfer (LRET). Simulations of micelle-solubilized LeuT revealed a stable and widely open inward-facing conformation. However, this conformation was unstable...... in a membrane environment. The helix dipole and the charged amino acid of the first transmembrane helix (TM1A) partitioned out of the hydrophobic membrane core. Free energy calculations showed that movement of TM1A by 0.30 nm was driven by a free energy difference of ~15 kJ/mol. Distance measurements by LRET...

  3. Analysis of the Ford Nuclear Reactor LEU core

    Energy Technology Data Exchange (ETDEWEB)

    Rathkopf, J A; Drumm, C R; Martin, W R; Lee, J C [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

    1983-09-01

    This paper has summarized the current status of the effort to analyze the FNR HEU/LEU cores and to compare the calculated results with measurements. In general, calculated predictions of experimental results are quite good, especially for global parameters such as reactivity, as seen in the single HEU/LEU element substitution experiment and the LEU full core critical loading. Shim rod worths are predicted well for two of the rods but too high for a third rod possibly due to inaccurate thermal flux distribution calculation. The calculated thermal flux maps show excellent agreement with experiment throughout the FNR core. In the heavy water tank, however, experimental values for the thermal flux obtained by different methods are inconsistent among themselves as well as with the calculated finding. Work is under.way to use our computational tools to correct the discrepancies between the various measurement techniques and to improve the computational results for flux distribution and the rod worth experiment. Although uncertainties exist in our analysis, as evidenced by the discrepancies mentioned above, we consider our present calculational package to be a useful, reasonably accurate, and efficient system for performing analyses of MTR LEU/HEU core configurations.

  4. The ORR Whole-Core LEU Fuel Demonstration

    International Nuclear Information System (INIS)

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

    1990-01-01

    The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U 3 Si 2 -Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged 235 U burnups validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of 235 U burnup support the corresponding measured quantities. In general, calculations for reaction rate distributions, control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 17 refs

  5. TEM investigation of irradiated U-7 weight percent Mo dispersion fuel

    International Nuclear Information System (INIS)

    Van den Berghe, S.

    2009-01-01

    In the FUTURE experiment, fuel plates containing U-7 weight percent Mo atomized powder were irradiated in the BR2 reactor. At a burn-up of approximately 33 percent 235 U (6.5 percent FIMA or 1.41 10 21 fissions/cm 3 meat), the fuel plates showed an important deformation and the irradiation was stopped. The plates were submitted to detailed PIE at the Laboratory for High and Medium level Activity. The results of these examinations were reported in the scientific report of last year and published in open literature. Since then, the microstructural aspects of the FUTURE fuel were studied in more detail using transmission electron microscopy (TEM), in an attempt to understand the nature of the interaction phase and the fission gas behavior in the atomized U(Mo) fuel. The FUTURE experiment is regarded as the definitive proof that the classical atomized U(Mo) dispersion fuel is not stable under irradiation, at least in the conditions required for normal operation of plate-type fuel. The main cause for the instability was identified to be the irradiation behavior of the U(Mo)-Al interaction phase which is formed between the U(Mo) particles and the pure aluminum matrix during irradiation. It is assumed to become amorphous under irradiation and as such cannot retain the fission gas in stable bubbles. As a consequence, gas filled voids are generated between the interaction layer and the matrix, resulting in fuel plate pillowing and failure. The objective of the TEM investigation was the confirmation of this assumption of the amorphisation of the interaction phase. A deeper understanding of the actual nature of this layer and the fission gas behaviour in these fuels in general can allow a more oriented search for a solution to the fuel failures

  6. Genetic interaction between the ero1-1 and leu2 mutations in Saccharomyces cerevisiae

    DEFF Research Database (Denmark)

    López-Mirabal, H Reynaldo; Winther, Jakob R; Kielland-Brandt, Morten C

    2007-01-01

    of the ero1-1 mutation were carried out in a leu2 mutant. The ero1-1 leu2 strain does not grow in standard synthetic complete medium at 30 degrees C, a defect that can be remedied by increasing the L-leucine concentration in the medium or by transforming the ero1-1 leu2 strain with the LEU2 wild-type allele...

  7. US Progress on Property Characterization to Support LEU U-10 Mo Monolithic Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Cole, James Irvin [Idaho National Laboratory; Rabin, Barry H [Idaho National Laboratory; Smith, James Arthur [Idaho National Laboratory; Scott, Clark Landon [Idaho National Laboratory; Benefiel, Bradley Curtis [Idaho National Laboratory; Larsen, Eric David [Idaho National Laboratory; Lind, Robert Paul [Idaho National Laboratory; Sell, David Alan [Idaho National Laboratory

    2016-03-01

    The US High Performance Research Reactor program is pursuing development and qualification of a new high density monolithic LEU fuel to facilitate conversion of five higher power research reactors located in the US (ATR, HFIR, NBSR, MIT and MURR). In order to support fabrication development and fuel performance evaluations, new testing capabilities are being developed to evaluate the properties of fuel specimens. Residual stress and fuel-cladding bond strength are two characteristics related to fuel performance that are being investigated. In this overview, new measurement capabilities being developed to assess these characteristics in both fresh and irradiated fuel are described. Progress on fresh fuel testing is summarized and on-going hot-cell implementation efforts to support future PIE campaigns are detailed. It is anticipated that benchmarking of as-fabricated fuel characteristics will be critical to establishing technical bases for specifications that optimize fuel fabrication and ensure acceptable in-reactor fuel performance.

  8. Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, Dennis D., E-mail: dennis.keiser@inl.gov [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States); Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States); Ross Finlay, M. [Australian Nuclear Science and Technology Organization, PMB 1, Menai, NSW 2234 (Australia); Moore, Glenn; Medvedev, Pavel; Meyer, Mitch [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States)

    2017-05-15

    The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

  9. Reclamation and reuse of LEU silicide fuel from manufacturing scrap

    International Nuclear Information System (INIS)

    Gale, G.R.; Pace, B.W.; Evans, R.S.

    2004-01-01

    In order to provide an understanding of the organization which is the sole supplier of United States plate type research and test reactor fuel and LEU core conversions, a brief description of the structure and history is presented. Babcock and Wilcox (B and W) is a part of McDermott International, Inc. which is a large diversified corporation employing over 20,000 people primarily in engineering and construction for the off-shore oil and power generation industries throughout the world. B and W provides many energy related products requiring precision machining and high quality systems. This is accomplished by using state-of-the-art equipment, technology and highly skilled people. The RTRFE group within B and W has the ability to produce various complexly shaped fuel elements with a wide variety of fuels and enrichments. B and W RTRFE has fabricated over 200,000 plates since 1981 and gained the diversified experience necessary to satisfy many customer requirements. This accomplishment was possible with the support of McDermott International and all of its resources. B and W has always had a commitment to high quality and integrity. This is apparent by the success and longevity (125 years) of the company. A lower cost to convert cores to LEU provides direct support to RERTR and demonstrates Babcock and Wilcox's commitment to the program. As a supporter of RERTR reactor conversion from HEU to LEU, B and W has contributed a significant amount of R and D money to improve the silicide fuel process which ultimately lowers the LEU core costs. In the most recent R and D project, B and W is constructing a LEU silicide reclamation facility to re-use the unirradiated fuel scrap generated from the production process. Remanufacturing use of this fuel completes the fuel cycle and provides a contribution to LEU cores by reducing scrap inventory and handling costs, lowering initial purchase of fuel due to increasing the process yields, and lowering the replacement costs. This

  10. Peptide (Lys-Leu) and amino acids (Lys and Leu) supplementations improve physiological activity and fermentation performance of brewer's yeast during very high-gravity (VHG) wort fermentation.

    Science.gov (United States)

    Yang, Huirong; Zong, Xuyan; Cui, Chun; Mu, Lixia; Zhao, Haifeng

    2017-12-22

    Lys and Leu were generally considered as the key amino acids for brewer's yeast during beer brewing. In the present study, peptide Lys-Leu and a free amino acid (FAA) mixture of Lys and Leu (Lys + Leu) were supplemented in 24 °P wort to examine their effects on physiological activity and fermentation performance of brewer's yeast during very high-gravity (VHG) wort fermentation. Results showed that although both peptide Lys-Leu and their FAA mixture supplementations could increase the growth and viability, intracellular trehalose and glycerol content, wort fermentability, and ethanol content for brewer's yeast during VHG wort fermentation, and peptide was better than their FAA mixture at promoting growth and fermentation for brewer's yeast when the same dose was kept. Moreover, peptide Lys-Leu supplementation significantly increased the assimilation of Asp, but decreased the assimilation of Gly, Ala, Val, (Cys)2, Ile, Leu, Tyr, Phe, Lys, Arg, and Pro. However, the FAA mixture supplementation only promoted the assimilation of Lys and Leu, while reduced the absorption of total amino acids to a greater extent. Thus, the peptide Lys-Leu was more effective than their FAA mixture on the improvement of physiological activity, fermentation performance, and nitrogen metabolism of brewer's yeast during VHG wort fermentation. © 2017 International Union of Biochemistry and Molecular Biology, Inc.

  11. 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.

  12. Immunoregulatory T cells in man. Histamine-induced suppressor T cells are derived from a Leu 2+ (T8+) subpopulation distinct from that which gives rise to cytotoxic T cells

    International Nuclear Information System (INIS)

    Sansoni, P.; Silverman, E.D.; Khan, M.M.; Melmon, K.L.; Engleman, E.G.

    1985-01-01

    One mechanism of histamine-mediated inhibition of the immune response in man is to activate T suppressor cells that bear the Leu 2 (OKT8) marker. The current study was undertaken to characterize the histamine-induced suppressor cell using a monoclonal antibody (9.3) shown previously to distinguish cytotoxic T cells from antigen-specific suppressor T cells. Leu 2+ cells isolated from peripheral blood were further separated with antibody 9.3 into Leu 2+, 9.3+, and Leu 2+, 9.3- subsets and each subset was incubated with different concentrations of histamine before determining their ability to suppress immune responses in vitro. The results indicate that the Leu 2+, 9.3- subpopulation includes all histamine-induced suppressor cells, that 10(-4) M histamine is the optimal concentration for suppressor cell induction, and that exposure of Leu 2+, 9.3- cells to histamine for 30 s is sufficient to initiate the induction process. After treatment with histamine these cells inhibit both phytohemagglutinin-induced T cell proliferation and pokeweed mitogen-induced B cell differentiation. The suppression of phytohemagglutinin-induced proliferation was resistant to x-irradiation with 1,200 rad, either before or after histamine exposure, suggesting that Leu 2+, 9.3- cells need not proliferate to become suppressor cells or exert suppression. Moreover, suppression by these cells was not due to altered kinetics of the response. Finally, a histamine type 2 receptor antagonist (cimetidine) but not a type 1 receptor antagonist (mepyramine) blocked the induction of suppressor cells. On the basis of these results and our previous studies of antigen specific suppressor cells, we conclude that Leu 2+ suppressor cells in man are derived from a precursor pool that is phenotypically distinct from cells that can differentiate into cytotoxic T cells

  13. Establishing a LEU MTR fuel manufacturing facility in South Africa

    International Nuclear Information System (INIS)

    Jamie, R.W.; Kocher, A.

    2010-01-01

    The South African MTR Fuel Manufacturing Facility was established in the 1970's to supply SAFARI-1 with Fuel Elements and Control Rods. South African capability was developed in parallel with the uranium enrichment program to meet the needs of the Reactor. Further to the July 2005 decision by the South African Governmnent to convert both SAFARI-1 and the Fuel Plant to LEU, the SAFARI-1 phase has been successfully completed and Necsa has commenced with the conversion of the MTR Fuel Manufacturing Facility. In order to establish, validate and qualify the facility, Necsa has entered into a co-operation and technology transfer agreement with AREVA CERCA, the French manufacturer of Research Reactor fuel elements. Past experiences, conversion challenges and the status of the MTR Fuel Facility Project are discussed. On-going co-operation with AREVA CERCA to implement the local manufacture of LEU fuel is explained and elaborated on. (author)

  14. HEU and Leu FueL Shielding Comparative Study Applied for Spent Fuel Transport

    International Nuclear Information System (INIS)

    Margeanu, C.A.; Margeanu, S.; Barbos, D.

    2009-01-01

    INR Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies. The dual core concept involves the operation of a 14 MW TRIGA steady-state, high flux research and material testing reactor at one end of a large pool, and the independent operation of an annular-core pulsing reactor (TRIGA-ACPR) at the other end of the pool. The steady-state reactor is mostly used for long term testing of power reactor fuel components (pellets, pins, subassemblies and fuel assemblies) followed by post-irradiation examination. Following the general trend to replace the He fuel type (High Enriched Uranium) by Leu fuel type (Low Enriched Uranium), in the light of international agreements between IAEA and the states using He fuel in their nuclear reactors, Inr Past's have been accomplished the TRIGA research reactor core full conversion on May 2006. The He fuel repatriation in US in the frame of Foreign Research Reactor Spent Nuclear Fuel Return Programme effectively started in 1999, the final stage being achieved in summer of 2008. Taking into account for the possible impact on the human and environment, in all activities associated to nuclear fuel cycle, the spent fuel or radioactive waste characteristics must be well known. Shielding calculations basic tasks consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper is a comparative study of Leu and He fuel utilization effects for the shielding analysis during spent fuel transport. A comparison against the measured data for He spent fuel, available from the last stage of the spent fuel repatriation, is presented. All the geometrical and material data related on the spent fuel shipping cask were considered according to the Nac-Lt Cask approved model. The shielding analysis estimates radiation doses to shipping cask wall surface

  15. Production of leu high density fuels at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Freim, J.B.

    1983-01-01

    A large number of fuel elements of all types are produced for both international and domestic customers by Nuclear Fuel Division of Babcock and Wilcox. A brief history of the division, included previous and present research reactor fuel element fabrication experience is discussed. The manufacturing facilities are briefly described. The fabrication of LEU fuels and economic analysis of the production are included. (A.J.)

  16. Thermal hydraulic analysis of the JMTR improved LEU-core

    Energy Technology Data Exchange (ETDEWEB)

    Tabata, Toshio; Nagao, Yoshiharu; Komukai, Bunsaku; Naka, Michihiro; Fujiki, Kazuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takeda, Takashi [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokai, Ibaraki (Japan)

    2003-01-01

    After the investigation of the new core arrangement for the JMTR reactor in order to enhance the fuel burn-up and consequently extend the operation period, the ''improved LEU core'' that utilized 2 additional fuel elements instead of formerly installed reflector elements, was adopted. This report describes the results of the thermal-hydraulic analysis of the improved LEU core as a part of safety analysis for the licensing. The analysis covers steady state, abnormal operational transients and accidents, which were described in the annexes of the licensing documents as design bases events. Calculation conditions for the computer codes were conservatively determined based on the neutronic analysis results and others. The results of the analysis, that revealed the safety criteria were satisfied on the fuel temperature, DNBR and primary coolant temperature, were used in the licensing. The operation license of the JMTR with the improved LEU core was granted in March 2001, and the reactor operation with new core started in November 2001 as 142nd operation cycle. (author)

  17. Neutronics Study on LEU Nuclear Thermal Rocket Fuel Options

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, Paolo; Kim, Yong Hee [KAIST, Daejeon (Korea, Republic of); Howe, Steven [CSNR, Idaho (United States)

    2014-10-15

    This has resulted in a non-trivial simplification of the tasks needed to develop such an engine and the quick initial development of the concept. There are, however, a series of key core-design choices that are currently under scrutiny in the field that have to be resolved in order for the LEU-NTR to be fully developed. The most important of these is the choice of fuel: carbide composite or tungsten cermet. This study presents a first comparison of the two fuel types specifically in the neutronic application to the LEU-NTR, keeping in mind the unique neutronic environment and the system requirements of the system. The scope of the study itself is limited to a neutronics study of the two fuels and only a cursory overview of the material properties of the fuels themselves... The results of this study have led to two major conclusions. First of all is that the carbide composite fuel is, from a neutronics standpoint, a much better fuel. It has a low absorption cross-section, is inherently a strong moderator, is able to achieve a higher reactivity using smaller amounts of fissile material, and can potentially enable a smaller reactor. Second is that despite its neutronic difficulties (high absorption, inferior moderating abilities, and lower k-infinity values) the tungsten cermet fuel is still able to perform satisfactorily in an LEU-NTR, largely due to its ability to have an extremely high fuel loading.

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  19. 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.

  20. Induction of CD4 suppressor T cells with anti-Leu-8 antibody

    International Nuclear Information System (INIS)

    Kanof, M.E.; Strober, W.; James, S.P.

    1987-01-01

    To characterize the conditions under which CD4 T cells suppress polyclonal immunoglobulin synthesis, we investigated the capacity of CD4 T cells that coexpress the surface antigen recognized by the monoclonal antibody anti-Leu-8 to mediate suppression. In an in vitro system devoid of CD8 T cells, CD4, Leu-8+ T cells suppressed pokeweed mitogen-induced immunoglobulin synthesis. Similarly, suppressor function was induced in unfractionated CD4 T cell populations after incubation with anti-Leu-8 antibody under cross-linking conditions. This induction of suppressor function by anti-Leu-8 antibody was not due to expansion of the CD4, Leu-8+ T cell population because CD4 T cells did not proliferate in response to anti-Leu-8 antibody. However, CD4, Leu-8+ T cell-mediated suppression was radiosensitive. Finally, CD4, Leu-8+ T cells do not inhibit immunoglobulin synthesis when T cell lymphokines were used in place of helper CD4 T cells (CD4, Leu-8- T cells), suggesting that CD4 T cell-mediated suppression occurs at the T cell level. We conclude that CD4 T cells can be induced to suppress immunoglobulin synthesis by modulation of the membrane antigen recognized by anti-Leu-8 antibody

  1. Transmission electron microscopy investigation of neutron irradiated Si and ZrN coated UMo particles prepared using FIB

    Science.gov (United States)

    Van Renterghem, W.; Miller, B. D.; Leenaers, A.; Van den Berghe, S.; Gan, J.; Madden, J. W.; Keiser, D. D.

    2018-01-01

    Two fuel plates, containing Si and ZrN coated U-Mo fuel particles dispersed in an Al matrix, were irradiated in the BR2 reactor of SCK•CEN to a burn-up of ∼70% 235U. Five samples were prepared by INL using focused ion beam milling and transported to SCK•CEN for transmission electron microscopy (TEM) investigation. Two samples were taken from the Si coated U-Mo fuel particles at a burn-up of ∼42% and ∼66% 235U and three samples from the ZrN coated U-Mo at a burn-up of ∼42%, ∼52% and ∼66% 235U. The evolution of the coating, fuel structure, fission products and the formation of interaction layers are discussed. Both coatings appear to be an effective barrier against fuel matrix interaction and only on the samples having received the highest burn-up and power, the formation of an interaction between Al and U(Mo) can be observed on those locations where breaches in the coatings were formed during plate fabrication.

  2. 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)

  3. 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)

  4. 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

  5. Nuclear criticality assessment of LEU and HEU fuel element storage

    International Nuclear Information System (INIS)

    Pond, R.B.; Matos, J.E.

    1984-01-01

    Criticality aspects of storing LEU (20%) and HEU (93%) fuel elements have been evaluated as a function of 235 U loading, element geometry, and fuel type. Silicide, oxide, and aluminide fuel types have been evaluated ranging in 235 U loading from 180 to 620 g per element and from 16 to 23 plates per element. Storage geometry considerations have been evaluated for fuel element separations ranging from closely packed formations to spacings of several centimeters between elements. Data are presented in a form in which interpolations may be made to estimate the eigenvalue of any fuel element storage configuration that is within the range of the data. (author)

  6. Greenfield Alternative Study LEU-Mo Fuel Fabrication Facility

    Energy Technology Data Exchange (ETDEWEB)

    Washington Division of URS

    2008-07-01

    This report provides the initial “first look” of the design of the Greenfield Alternative of the Fuel Fabrication Capability (FFC); a facility to be built at a Greenfield DOE National Laboratory site. The FFC is designed to fabricate LEU-Mo monolithic fuel for the 5 US High Performance Research Reactors (HPRRs). This report provides a pre-conceptual design of the site, facility, process and equipment systems of the FFC; along with a preliminary hazards evaluation, risk assessment as well as the ROM cost and schedule estimate.

  7. Behaviour of uranium under irradiation

    International Nuclear Information System (INIS)

    Adda, Y.; Mustelier, J.P.; Quere, Y.; Commissariat a l'Energie Atomique, Fontenay-aux-Roses

    1964-01-01

    The main results obtained in a study of the formation of defects caused in uranium by fission at low temperature are reported. By irradiation at 20 K. it was possible to determine the number of Frenkel pairs produced by one fission. An analysis of the curves giving the variations in electrical resistivity shows the size of the displacement spikes and the mechanism of defect creation due to fission. Irradiations at 77 K gave additional information, showing behaviour differences in the case of recrystallised and of cold worked uranium. The diffusion of rare gases was studied using metal-rare gas alloys obtained by electrical discharge, and samples of irradiated uranium. Simple diffusion is only responsible for the release of the rare gases under vacuum in cases where the rare gas content is very low (very slightly irradiated U). On the other hand when the concentration is higher (samples prepared by electrical discharge) the gas is given off by the formation, growth and coalescence of bubbles; the apparent diffusion coefficient is then quite different from the true coefficient and cannot be used in calculations on swelling. The various factors governing the phenomenon of simple diffusion were examined. It was shown in particular that a small addition of molybdenum could reduce the diffusion coefficient by a factor of 100. The precipitation of gas in uranium (Kr), in silver (Kr) and in Al-Li alloy (He) have been followed by measurement of the crystal parameter and of the electrical resistivity, and by electron microscope examination of thin films. The important part played by dislocations in the generation and growth of bubbles has been demonstrated, and it has been shown also that precipitation of bubbles on the dislocation lattice could block the development of recrystallisation. The results of these studies were compared with observations made on the swelling of uranium and uranium alloys U Mo and U Nb strongly irradiated between 400 and 700 C. In the case of Cubic

  8. Preproghrelin Leu72Met polymorphism in Chinese subjects with coronary artery disease and controls.

    Science.gov (United States)

    Tang, Na-Ping; Wang, Lian-Sheng; Yang, Li; Gu, Hai-Juan; Zhu, Huai-Jun; Zhou, Bo; Sun, Qing-Min; Cong, Ri-Hong; Wang, Bin

    2008-01-01

    Ghrelin, a novel endogenous ligand for the growth hormone secretagogue receptor, is considered to exert a protective effect against atherosclerosis. The Leu72Met (+408C>A) polymorphic variant of the preproghrelin, the gene for the ghrelin precursor, has been linked to obesity, diabetes and metabolic syndrome. However, it is unclear whether this polymorphism is associated with coronary artery disease (CAD). We conducted a case-control study with 317 CAD patients and 323 controls to investigate the potential association of the Leu72Met polymorphism with the occurrence of CAD and CAD-related phenotypes in Chinese population. No significant difference in the Leu72Met genotype frequency was observed between CAD patients and controls (P=NS). The Leu72Met polymorphism was not associated with hypertension, diabetes, dyslipidemia, the number of diseased vessels, plasma total cholesterol, triglyceride, high density lipoprotein cholesterol, low density lipoprotein cholesterol or fasting glucose levels in CAD patients. However, among CAD patients, those with variant genotypes (Leu72Met and Met72Met) had lower BMI (24.4+/-0.3 kg/m(2)) than Leu72Leu carriers (25.4+/-0.2 kg/m(2), adjusted P=0.033). Our data indicate that the preproghrelin Leu72Met polymorphism is not associated with CAD in Chinese population. However, the Leu72Met variant is associated with BMI among CAD patients.

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

    International Nuclear Information System (INIS)

    Alwaer, Sami M.

    2015-01-01

    The object of this work was to assist the IAEA by providing the Libyan country report about the Coordination Research Project (CRP), on the subject of “Developing techniques for small scale indigenous Mo-99 production using LEU-foil” which took place over five years and four RCMs. A CRP on this subject was approved in early 2005. The objectives of this CRP are to: transfer know-how in the area of 99 Mo production using LEU targets based on reference technologies from leading laboratories in the field to the participating laboratories in the CRP; develop national work plans based on various stages of technical development and objectives in this field; establish the procedures and protocols to be employed, including quality control and assurance procedures; establish the coordinated activities and programme for preparation, irradiation, and processing of LEU targets [a]; and to compare results obtained in the implementation of the technique in order to provide follow up advice and assistance. Technetium-99m ( 99m Tc), the daughter product of molybdenum-99 ( 99 Mo), is the most commonly utilized medical radioisotope in the world, used for approximately 20-25 million medical diagnostic procedures annually, comprising some 80% of all diagnostic nuclear medicine procedures. National and international efforts are underway to shift the production of medical isotopes from highly enriched uranium (HEU) to low enriched uranium (LEU) targets. A small but growing amount of the current global 99 Mo production is derived from the irradiation of LEU targets. The IAEA became aware of the interest of a number of developing Member States that are seeking to become small scale, indigenous producers of 99 Mo to meet local nuclear medicine requirements. The IAEA initiated Coordinated Research Project (CRP) T.1.20.18 “Developing techniques for small-scale indigenous production of Mo-99 using LEU or neutron activation” in order to assist countries in this field. The more

  10. 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

  11. Data Compilation for AGR-1 Baseline Coated Particle Composite LEU01-46T

    International Nuclear Information System (INIS)

    Hunn, John D.; Lowden, Richard Andrew

    2006-01-01

    This document is a compilation of characterization data for the AGR-1 baseline coated particle composite LEU01-46T, a composite of four batches of TRISO-coated 350 (micro)m 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with a ∼ 50% dense carbon buffer layer (100 (micro)m nominal thickness) followed by a dense inner pyrocarbonlayer (40 (micro)m nominal thickness) followed by a SiC layer (35 (micro)m nominal thickness) followed by another dense outer pyrocarbon layer (40 (micro)m nominal thickness). The coated particles, were produced by ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program to be put into compacts for insertion in the first irradiation test capsule, AGR-1. The kernels were obtained from BWXT and identified as composite (G73D-20-69302). The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified as LEU01-?? (where ?? is a series of integers beginning with 01). Additional particle batches were coated with only buffer or buffer plus inner pyrocarbon (IPyC) layers using similar process conditions as used for the full TRISO batches comprising the LEU01-46T composite. These batches were fabricated in order to qualify that the process conditions used for buffer and IPyC would produce acceptable densities, as described in sections 8 and 9. These qualifying batches used 350 (micro)m natural uranium oxide/uranium carbide kernels (NUCO). The kernels were obtained from BWXT and identified as composite G73B-NU-69300. The use of NUCO surrogate kernels is not expected to significantly effect the densities of the buffer and IPyC coatings. Confirmatory batches using LEUCO kernels from G73D-20-69302 were coated and characterized to verify this assumption. The AGR-1 Fuel Product Specification and Characterization Guidance (INL EDF-4380, Rev. 6) provides the requirements necessary for acceptance

  12. Preparation results for lifetime test of conversion LEU fuel in plutonium production reactors

    International Nuclear Information System (INIS)

    Vatulin, A.; Stetskiy, Yu.; Kukharkin, N.; Kalougin, A.; Gavrilov, P.; Ivanov, A.

    1999-01-01

    The program of converting Russian production reactors for the purpose to stop their plutonium fabrication is currently in progress. The program also provides for operation of these reactors under the conversion mode with using of low-enriched fuel (LEU). LEU fuel elements were developed and activities related to their preparation for reactor tests were carried out. (author)

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

    International Nuclear Information System (INIS)

    Mueller, H.

    1997-01-01

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

  14. 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

  15. Neutronic analysis of a reference LEU core for Pakistan research reactor using oxide fuel

    International Nuclear Information System (INIS)

    Akhtar, K.M.; Qazi, M.K.; Bokhari, I.H.; Khan, L.A.; Pervez, S.

    1988-07-01

    Neutronic analysis of a 10 MW reference core for PARR, having 28 fresh LEU fuel elements arranged in a 6x5 configuration has been carried out using standard computer codes WIMS-D, EXTERMINATOR-II, and CITATION. Total nuclear power peaking of 3.2 has bee found to occur in the fuel plate adjacent to the water filled central flux trap at the depth of 43.8 cm from the top of the active core. Replacement of water in central flux trap with an aluminum block, having a 50 mm diameter water filled irradiation channel changes the flux profiles in fuel, core side flux trap and reflector. The thermal flux in the central flux trap decreases by about 53%. Therefore some of the fuel elements will have to be removed and the new configuration has to be analysed to determine the first operating core. However, after achieving some burn-up and confirmation from thermal hydraulic analysis, the core configuration analysed, will be the final working core. (orig./A.B.)

  16. 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)

  17. 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)

  18. 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.

  19. Safety analysis of JMTR LEU fuel core, (3)

    International Nuclear Information System (INIS)

    Tsuchida, Noboru; Shiraishi, Tadao; Takahashi, Yutaka; Inada, Seiji; Saito, Minoru; Futamura, Yoshiaki; Kitano, Kyoshiro.

    1992-10-01

    Dose analysis in the safety evaluation and the site evaluation were performed for the JMTR core conversion from MEU fuel to LEU fuel. In the safety evaluation, the effective dose equivalents for the public surrounding the site were estimated in fuel handling accident and flow blockage to coolant channel which were selected as the design basis accidents with release of radioactive fission products to the environment. In the site evaluation, the flow blockage to coolant channel was selected as siting basis events, since this accident had the possibility of spreading radioactive release. Maximum exposure doses for the public were estimated assuming large amounts of fission products to release. It was confirmed that risk of radiation exposure of the public is negligible and the siting is appropriate. (author)

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

    International Nuclear Information System (INIS)

    Snelgrove, J.L.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

    After presentation of the foregoing paper by Dr. Nelson Hanan of Argonne National Laboratory (ANL) proposing an alternative LEU core with one fuel ring and a power level of 33 MW, a presentation was made by Dr. Klaus Boning of the Technical University of Munich comparing the FRM-II HEU design with an LEU design by Tlm that had two fuel rings and a power level of 40 MW. Dr. Boning raised the following issues concerning the use of LEU fuel in FRM-H reactor designs: (1) qualification of HEU and LEU silicide fuels, (2) gamma heating in the heavy water reflector, (3) the radiological consequences of hypothetical accidents, and (4) cost and schedule. These issues are addressed in this Attachment. In his presentation, Dr. Hanan mentioned that ANL was also investigating other LEU designs. This work led to a second alternative LEU design that has the same neutron flux performance (8 x 10 14 n/cm 2 /s peak neutron flux in the reflector) and the same fuel lifetime (50 full power days) as the HEU design, but uses LEU silicide fuel with a uranium density of only 4.5 g/cm 3 . This design was achieved by using a fuel plate that has a fuel meat thickness of 0.76 mm, a cladding thickness of 0.38 mm, and a water channel gap of 2.2 mm. A comparison is shown of the main characteristics of this second alternative LEU design with those of the FRM-II HEU design. The ANL core again has one fuel ring with the same dimensions. With this LEU design, a two stage process is no longer necessary because LEU silicide fuel with a uranium density of 4.5 g/cm 3 is fully qualified, licensable, and available now for use in a high flux reactor such as the FRM-II

  2. Distinguishing Isomeric Peptides: The Unimolecular Reactivity and Structures of (LeuPro)M+ and (ProLeu)M+ (M = Alkali Metal).

    Science.gov (United States)

    Jami-Alahmadi, Yasaman; Linford, Bryan D; Fridgen, Travis D

    2016-12-29

    The unimolecular chemistries and structures of gas-phase (ProLeu)M + and (LeuPro)M + complexes when M = Li, Na, Rb, and Cs have been explored using a combination of SORI-CID, IRMPD spectroscopy, and computational methods. CID of both (LeuPro)M + and (ProLeu)M + showed identical fragmentation pathways and could not be differentiated. Two of the fragmentation routes of both peptides produced ions at the same nominal mass as (Pro)M + and (Leu)M + , respectively. For the litiated peptides, experiments revealed identical IRMPD spectra for each of the m/z 122 and 138 ions coming from both peptides. Comparison with computed IR spectra identified them as the (Pro)Li + and (Leu)Li + , and it is concluded that both zwitterionic and canonical forms of (Pro)Li + exist in the ion population from CID of both (ProLeu)Li + and (LeuPro)Li + . The two isomeric peptide complexes could be distinguished using IRMPD spectroscopy in both the fingerprint and the CH/NH/OH regions. The computed IR spectra for the lowest energy structures of each charge solvated complexes are consistent with the IRMPD spectra in both regions for all metal cation complexes. Through comparison between the experimental spectra, it was determined that in lithiated and sodiated ProLeu, metal cation is bound to both carbonyl oxygens and the amine nitrogen. In contrast, the larger metal cations are bound to the two carbonyls, while the amine nitrogen is hydrogen bonded to the amide hydrogen. In the lithiated and sodiated LeuPro complexes, the metal cation is bound to the amide carbonyl and the amine nitrogen while the amine nitrogen is hydrogen bonded to the carboxylic acid carbonyl. However, there is no hydrogen bond in the rubidiated and cesiated complexes; the metal cation is bound to both carbonyl oxygens and the amine nitrogen. Details of the position of the carboxylic acid C═O stretch were especially informative in the spectroscopic confirmation of the lowest energy computed structures.

  3. A leukocyte antigen, Leu-13, is involved in induction of resistance of human cells to x-ray cell killing by interferon-α

    International Nuclear Information System (INIS)

    Kita, Kazuko; Zhai, Ling; Sugaya, Shigeru; Suzuki, Nobuo

    2003-01-01

    We previously reported on human interferon (HuIFN)-induced resistance of human cells to X-ray and UV cell killing. In this study, we searched for the genes whose expression is responsible for the resistance, using a PCR-based mRNA differential display method and Northern blotting analysis. RSa cells were used for this analysis, because they show increased resistance to X-ray- and UV-caused cell killing by HuIFN-α treatment prior to irradiation. Messenger RNA expression levels for Leu-13, a leukocyte antigen, were markedly up-regulated in RSa cells after HuIFN-α treatment. Furthermore, pretreatment of RSa cells with antisense oligonucleotides for Leu-13 mRNA resulted in the suppression of the HuIFN-α-induced resistance of the cells to X-ray cell killing, but did not modulate HuIFN-α-induced resistance to UV cell killing. These results suggest that Leu-13 is involved in HuIFN-α-induced resistance of human cells to X-ray cell killing, but not to UV cell killing. (author)

  4. Nuclear and radiological safety in the substitution process of the fuel HEU to LEU 30/20 in the Reactor TRIGA Mark III of the ININ

    International Nuclear Information System (INIS)

    Hernandez G, J.

    2012-10-01

    Inside the safety initiative in the international ambit, with the purpose of reducing the risks associated with the use of high enrichment nuclear fuels (HEU) for different proposes to the peaceful uses of the nuclear energy, Mexico contributes by means of the substitution of the high enrichment fuel HEU for low enrichment fuel LEU 30/20 in the TRIGA Mark III Reactor, belonging to Instituto Nacional de Investigaciones Nucleares (ININ). The conversion process was carried out by means of the following activities: analysis of the proposed core, reception and inspection of the fuel LEU 30/20, the discharge of the fuels of the mixed reactor core, shipment of the fuels HEU fresh and irradiated to the origin country, reload activities with the fuels LEU 30/20 and parameters measurement of the core operation. In order to maintaining the personnel's integrity and infrastructure associated to the Reactor, during the whole process the measurements of nuclear and radiological safety were controlled to detail, in execution with the license requirements of the installation. This work describes the covering activities and radiological inspections more relevant, as well as the measurements of radiological control implemented with base in the estimate of the equivalent dose of the substitution process. (Author)

  5. SEM Characterization of an Irradiated Monolithic U-10Mo Fuel Plate

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B.

    2010-01-01

    Results of scanning electron microscopy (SEM) characterization of irradiated U-7Mo dispersion fuel plates with differing amounts of matrix Si have been reported. However, to date, no results of SEM analysis of irradiated U-Mo monolithic fuel plates have been reported. This paper describes the first SEM characterization results for an irradiated monolithic U-10Mo fuel plate. Two samples from this fuel plate were characterized. One sample was produced from the low-flux side of the fuel plate, and another was produced at the high-flux side of the fuel plate. This characterization focused on the microstructural features present at the U-10Mo foil/cladding interface, particularly the interaction zone that had developed during fabrication and irradiation. In addition, the microstructure of the foil itself was investigated, along with the morphology of the observed fission gas bubbles. It was observed that a Si-rich interaction layer was present at the U-10Mo foil/cladding interface that exhibited relatively good irradiation behavior, and within the U-10Mo foil the microstructural features differed in some respects from what is typically seen in the U-Mo powders of an irradiated dispersion fuel.

  6. Scalability of the LEU-Modified Cintichem Process: 3-MeV Van de Graaff and 35-MeV Electron Linear Accelerator Studies

    Energy Technology Data Exchange (ETDEWEB)

    Rotsch, David A. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Brossard, Tom [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Roussin, Ethan [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Quigley, Kevin [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Chemerisov, Sergey [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gromov, Roman [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Jonah, Charles [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hafenrichter, Lohman [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Tkac, Peter [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Krebs, John [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

    2016-10-31

    Molybdenum-99, the mother of Tc-99m, can be produced from fission of U-235 in nuclear reactors and purified from fission products by the Cintichem process, later modified for low-enriched uranium (LEU) targets. The key step in this process is the precipitation of Mo with α-benzoin oxime (ABO). The stability of this complex to radiation has been examined. Molybdenum-ABO was irradiated with 3 MeV electrons produced by a Van de Graaff generator and 35 MeV electrons produced by a 50 MeV/25 kW electron linear accelerator. Dose equivalents of 1.7–31.2 kCi of Mo-99 were administered to freshly prepared Mo-ABO. Irradiated samples of Mo-ABO were processed according to the LEU Modified-Cintichem process. The Van de Graaff data indicated good radiation stability of the Mo-ABO complex up to ~15 kCi dose equivalents of Mo-99 and nearly complete destruction at doses >24 kCi Mo-99. The linear accelerator data indicate that even at 6.2 kCi of Mo-99 equivalence of dose, the sample lost ~20% of Mo-99. The 20% loss of Mo-99 at this low dose may be attributed to thermal decomposition of the product from the heat deposited in the sample during irradiation.

  7. 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)

  8. Leu-9 (CD 7) positivity in acute leukemias: a marker of T-cell lineage?

    Science.gov (United States)

    Ben-Ezra, J; Winberg, C D; Wu, A; Rappaport, H

    1987-01-01

    Monoclonal antibody Leu-9 (CD 7) has been reported to be a sensitive and specific marker for T-cell lineage in leukemic processes, since it is positive in patients whose leukemic cells fail to express other T-cell antigens. To test whether Leu-9 is indeed specific for T-cell leukemias, we examined in detail 10 cases of acute leukemia in which reactions were positive for Leu-9 and negative for other T-cell-associated markers including T-11, Leu-1, T-3, and E-rosettes. Morphologically and cytochemically, 2 of these 10 leukemias were classified as lymphoblastic, 4 as myeloblastic, 2 as monoblastic, 1 as megakaryoblastic, and 1 as undifferentiated. The case of acute megakaryoblastic leukemia is the first reported case to be Leu-9 positive. None of the 10 were TdT positive. Of six cases (two monoblastic, one lymphoblastic, one myeloblastic, one megakaryoblastic, and one undifferentiated) in which we evaluated for DNA gene rearrangements, only one, a peroxidase-positive leukemia, showed a novel band on study of the T-cell-receptor beta-chain gene. We therefore conclude that Leu-9 is not a specific marker to T-cell lineage and that, in the absence of other supporting data, Leu-9 positivity should not be used as the sole basis of classifying an acute leukemia as being T-cell derived.

  9. 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.

  10. Preproghrelin Leu72Met polymorphism in patients with type 2 diabetes mellitus.

    Science.gov (United States)

    Ukkola, O; Kesäniemi, Y A

    2003-10-01

    The association between the Leu72Met polymorphism of the preproghrelin gene and diabetic complications was examined in patients with type 2 diabetes mellitus. A total of 258 patients with type 2 diabetes mellitus and 522 control subjects were screened. Genotypes were determined by polymerase chain reaction technique. The diagnosis of coronary heart disease was based on clinical and ECG criteria. Laboratory analyses were carried out in the hospital laboratory. No differences in the genotype distributions and allele frequencies of the preproghrelin Leu72Met polymorphism were found between type 2 diabetes mellitus patients and controls. The polymorphism was not associated with macro- or micro-angiopathy or hypertension. However, Leu72Met polymorphism was associated with serum creatinine (P = 0.006) and lipoprotein(a) [Lp(a)] levels (P = 0.006) with Leu72Leu subjects showing the highest values. This association was observed only amongst diabetic group. The Leu72Met polymorphism of the preproghrelin gene was not related to cardiovascular disease in type 2 diabetes mellitus patients. Leu72Met polymorphism was, however, associated with serum creatinine and Lp(a) levels in diabetic patients. The mechanism might be associated with a possible change in ghrelin product and its somatotropic effect.

  11. Pressure effect on the conformational equilibrium of [Leu]{sup 5}-enkephalin in water

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, A [Department of Environmental Engineering for Symbiosis, Soka University, 1-326 Tangi-cho, Hachioji, Tokyo, 192-8577 (Japan); Takekiyo, T; Yoshimura, Y [Department of Applied Chemistry, National Defence Academy, 1-10-20 Hashirimizu, Yokosuka, Kanagawa, 239-8686 (Japan); Kato, M; Taniguchi, Y, E-mail: shimizu@soka.ac.j, E-mail: take214@nda.ac.j [Department of Applied Chemistry, Ritsumeikan University, 1-1-1, Nojihigashi, Kusatsu, Shiga, 525-8577 (Japan)

    2010-03-01

    The conformational stability of [Leu]{sup 5}-enkephalin,Tyr-Gly-Gly-Phe-Leu, in water have been investigated under high pressure by FTIR spectroscopy. Three peaks at 1638, 1650, and 1680 cm{sup -1} were determined by second derivative FTIR spectra in the amide I' region of [Leu]{sup 5}-enkephalin. The peaks at 1637 and 1680 cm{sup -1} are assigned to the {beta}-strand and turn structures, respectively. These peaks mean that [Leu]{sup 5}-enkephalin takes a {beta}-hairpin-like structure in water. Moreover, the absorbance at 1638 cm{sup -1} increases with increasing pressure, and this change shows a sigmoidal curve. Thus, we concluded that [Leu]{sup 5}-enkephalin has the {beta}-hairpin-like and disordered structures in water. From the FTIR profile at high pressures, the {beta}-hairpin-like structure of [Leu]{sup 5}-enkephalin is stabilized by a high pressures. Our result shows that the folded structures such as {alpha}-helix and {beta}-hairpin structures of short peptide such as [Leu]{sup 5}-enkephalin are stabilized at high pressures.

  12. A neutronic feasibility study for LEU conversion of the High Flux Beam Reactor (HFBR)

    International Nuclear Information System (INIS)

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

    1997-01-01

    A neutronic feasibility study for converting the High Flux Beam Reactor at Brookhaven National Laboratory from HEU to LEU fuel was performed at Argonne National Laboratory. The purpose of this study is to determine what LEU fuel density would be needed to provide fuel lifetime and neutron flux performance similar to the current HEU fuel. The results indicate that it is not possible to convert the HFBR to LEU fuel with the current reactor core configuration. To use LEU fuel, either the core needs to be reconfigured to increase the neutron thermalization or a new LEU reactor design needs to be considered. This paper presents results of reactor calculations for a reference 28-assembly HEU-fuel core configuration and for an alternative 18-assembly LEU-fuel core configuration with increased neutron thermalization. Neutronic studies show that similar in-core and ex-core neutron fluxes, and fuel cycle length can be achieved using high-density LEU fuel with about 6.1 gU/cm 3 in an altered reactor core configuration. However, hydraulic and safety analyses of the altered HFBR core configuration needs to be performed in order to establish the feasibility of this concept. (author)

  13. Oral delivery of [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3, synthetic peptide leptin mimetics: Immunofluorescent localization in the mouse hypothalamus.

    Science.gov (United States)

    Anderson, Brian M; Jacobson, Lauren; Novakovic, Zachary M; Grasso, Patricia

    2017-06-01

    This study describes the localization of [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3, synthetic peptide leptin mimetics, in the hypothalamus of Swiss Webster and C57BL/6J wild-type mice, leptin-deficient ob/ob mice, and leptin-resistant diet-induced obese (DIO) mice. The mice were given [D-Leu-4]-OB3 or MA-[D-Leu-4]-OB3 in 0.3% dodecyl maltoside by oral gavage. Once peak serum concentrations were reached, the mice received a lethal dose of pentobarbital and were subjected to intracardiac perfusion fixation. The brains were excised, post-fixed in paraformaldehyde, and cryo-protected in sucrose. Free-floating frozen coronal sections were cut at 25-µm and processed for imaging by immunofluorescence microscopy. In all four strains of mice, dense staining was concentrated in the area of the median eminence, at the base and/or along the inner wall of the third ventricle, and in the brain parenchyma at the level of the arcuate nucleus. These results indicate that [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3 cross the blood-brain barrier and concentrate in an area of the hypothalamus known to regulate energy balance and glucose homeostasis. Most noteworthy is the localization of [D-Leu-4]-OB3 immunoreactivity within the hypothalamus of DIO mice via a conduit that is closed to leptin in this rodent model, and in most cases of human obesity. Together with our previous studies describing the effects of [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3 on energy balance, glucose regulation, and signal transduction pathway activation, these findings are consistent with a central mechanism of action for these synthetic peptide leptin mimetics, and suggest their potential usefulness in the management of leptin-resistant obesity and type 2 diabetes in humans. Copyright © 2017 Elsevier B.V. All rights reserved.

  14. Poliovirus Polymerase Leu420 Facilitates RNA Recombination and Ribavirin Resistance

    Science.gov (United States)

    Kempf, Brian J.; Peersen, Olve B.

    2016-01-01

    ABSTRACT RNA recombination is important in the formation of picornavirus species groups and the ongoing evolution of viruses within species groups. In this study, we examined the structure and function of poliovirus polymerase, 3Dpol, as it relates to RNA recombination. Recombination occurs when nascent RNA products exchange one viral RNA template for another during RNA replication. Because recombination is a natural aspect of picornavirus replication, we hypothesized that some features of 3Dpol may exist, in part, to facilitate RNA recombination. Furthermore, we reasoned that alanine substitution mutations that disrupt 3Dpol-RNA interactions within the polymerase elongation complex might increase and/or decrease the magnitudes of recombination. We found that an L420A mutation in 3Dpol decreased the frequency of RNA recombination, whereas alanine substitutions at other sites in 3Dpol increased the frequency of recombination. The 3Dpol Leu420 side chain interacts with a ribose in the nascent RNA product 3 nucleotides from the active site of the polymerase. Notably, the L420A mutation that reduced recombination also rendered the virus more susceptible to inhibition by ribavirin, coincident with the accumulation of ribavirin-induced G→A and C→U mutations in viral RNA. We conclude that 3Dpol Leu420 is critically important for RNA recombination and that RNA recombination contributes to ribavirin resistance. IMPORTANCE Recombination contributes to the formation of picornavirus species groups and the emergence of circulating vaccine-derived polioviruses (cVDPVs). The recombinant viruses that arise in nature are occasionally more fit than either parental strain, especially when the two partners in recombination are closely related, i.e., members of characteristic species groups, such as enterovirus species groups A to H or rhinovirus species groups A to C. Our study shows that RNA recombination requires conserved features of the viral polymerase. Furthermore, a

  15. How LeuT shapes our understanding of the mechanisms of sodium-coupled neurotransmitter transporters.

    Science.gov (United States)

    Penmatsa, Aravind; Gouaux, Eric

    2014-03-01

    Neurotransmitter transporters are ion-coupled symporters that drive the uptake of neurotransmitters from neural synapses. In the past decade, the structure of a bacterial amino acid transporter, leucine transporter (LeuT), has given valuable insights into the understanding of architecture and mechanism of mammalian neurotransmitter transporters. Different conformations of LeuT, including a substrate-free state, inward-open state, and competitive and non-competitive inhibitor-bound states, have revealed a mechanistic framework for the transport and transport inhibition of neurotransmitters. The current review integrates our understanding of the mechanistic and pharmacological properties of eukaryotic neurotransmitter transporters obtained through structural snapshots of LeuT.

  16. Nuclear Thermal Rocket Design Using LEU Tungsten Fuel

    International Nuclear Information System (INIS)

    Venneri, Paolo; Kim, Yonghee; Husemeyer, Peter and others

    2013-01-01

    This would then open the possibility for the commercial development and implementation of an NTR. The result was a design for a 114.66 kN thrust rocket engine, with an optimized specific impulse of 801 second, and a thrust-to-weight ratio 5.08. The development and analysis of the reactor was done using an integrated neutronics and thermal hydraulics code that combines MCNP5 using ENDF-B/VI cross sections with a purpose-built thermal hydraulics code. A proof of concept has been proposed for W LEU-NTR design. The current design is built upon traditional NTR design work and implements many of the proven design characteristics and materials from previous designs. Despite the current reactor design being preliminary, it already shows promise in being able to have similar, if not better performance characteristics than current and previous NTR designs. Future work will involve the flattening of radial power profile, optimization of the axial power profile, researching methods to address the full water immersion accident scenario, and further studies regarding the breeding potential in the reactor

  17. Nuclear Thermal Rocket Design Using LEU Tungsten Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, Paolo; Kim, Yonghee; Husemeyer, Peter and others

    2013-10-15

    This would then open the possibility for the commercial development and implementation of an NTR. The result was a design for a 114.66 kN thrust rocket engine, with an optimized specific impulse of 801 second, and a thrust-to-weight ratio 5.08. The development and analysis of the reactor was done using an integrated neutronics and thermal hydraulics code that combines MCNP5 using ENDF-B/VI cross sections with a purpose-built thermal hydraulics code. A proof of concept has been proposed for W LEU-NTR design. The current design is built upon traditional NTR design work and implements many of the proven design characteristics and materials from previous designs. Despite the current reactor design being preliminary, it already shows promise in being able to have similar, if not better performance characteristics than current and previous NTR designs. Future work will involve the flattening of radial power profile, optimization of the axial power profile, researching methods to address the full water immersion accident scenario, and further studies regarding the breeding potential in the reactor.

  18. Progress in safety evaluation for the JMTR core conversion to LEU fuel

    International Nuclear Information System (INIS)

    Sakurai, F.; Komori, Y.; Saito, J.; Komukai, B.; Ando, H.; Nakata, H.; Sakakura, A.; Niiho, S.; Saito, M.; Futamura, Y.

    1991-01-01

    The JMTR (50 MWt) has been in steady operation with MEU fuel since July 1986. The effort is still continued to convert the core from MEU to LEU fuel. The LEU silicide fuel element at 4.8 gU/cm 3 with Cd wires as burnable absorbers has been selected in order to achieve upgraded fuel cycle performance of extended cycle length and reduced control rod movement operation. The neutronic calculation methods (diffusion theory model) developed for the LEU core with Cd wires was benchmarked with a detailed Monte Carlo model and verified experimentally using the critical facility, JMTRC. Hydraulic tests of the LEU silicide fuel element with Cd wires were completed with satisfactory results, and measurements of release/born (R/B) ratios of FPs of silicide fuel at high temperature are in progress. (orig.)

  19. Core management and reactor physics aspects of the conversion of the NRU reactor to LEU

    International Nuclear Information System (INIS)

    Atfield, M.D.

    1985-01-01

    Results of work done to assess the effects of converting the NRU reactor to LEU are presented. The effects are small, and the operational rules and safety analysis, appropriate to the HEU core, will still apply. (author)

  20. Total synthesis of fully tritiated Leu-enkephalin by enzymatic coupling

    Energy Technology Data Exchange (ETDEWEB)

    Hellio, F.; Lecocq, G.; Morgat, J.L.; Gueguen, P. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France). Service de Biochimie)

    1990-09-01

    This paper describes the total enzymatic synthesis of Leu-enkephalin (Tyr-Gly-Gly-Phe-Leu) in which all residues were labelled with tritium. Carboxypeptidase Y from Saccharomyces cerevisiae was the coupling enzyme. ({sup 3}H)-Tyr-NH{sub 2}, ({sup 3}H)-Gly-Oet, ({sup 3}H)-Phe-NH{sub 2} and ({sup 3}H)-Leu-NH{sub 2} were prepared with specific radioactivities ranging between 20 and 60 Ci/mmol (740 to 2220 GBq/mmol). Using a microscale procedure, we obtained a fully tritiated hormone having a specific radioactivity equal to 139 Ci/mmol (5143 GBq/mmol), in agreement with the summation of the specific radioactivities of constituting residue. The radioactive hormone had antigenic properties identical to those of native Leu-enkephalin. It also bound to rat brain opiate receptors like the parental hormone. (author).

  1. Syntheses of deuterated leu-enkephalins and their use as internal standards for the quantification of leu-enkephalin by fast atom bombardment mass spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Benfenati, E. (Istituto di Ricerche Farmacologiche Mario Negri, Bergamo (Italy) Istituto di Ricerche Farmacologiche Mario Negri, Milan (Italy)); Icardi, G.; Chen, S. (Istituto di Ricerche Farmacologiche Mario Negri, Bergamo (Italy)); Fanelli, R. (Istituto di Ricerche Farmacologiche Mario Negri, Milan (Italy))

    1990-04-01

    We have developed a synthetic method for the preparation of di- and pentadeuterated leu-enkephalin (LE), Tyr-Gly-Gly-Phe-Leu, by proton-deuterium exchange using CF[sub 3]COOO[sup 2]H. Four to six deuterium atoms are introduced using a reaction temperature of 120[sup o]C and if 5% of [sup 2]H[sub 2]O is added the di-deuterated LE is obtained. These deuterated compounds are used as internal standards to plot calibration curves of LE using fast atom bombardment mass spectrometry. (author).

  2. 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)

  3. Preproghrelin Leu72Met polymorphism is not associated with type 2 diabetes mellitus.

    Science.gov (United States)

    Kim, Sun-Young; Jo, Dae-Sun; Hwang, Pyoung Han; Park, Ji Hyun; Park, Sung Kwang; Yi, Ho Keun; Lee, Dae-Yeol

    2006-03-01

    Ghrelin is a novel gut-brain peptide, which exerts somatotropic, orexigenic, and adipogenic effects. Genetic variants of ghrelin have been associated with both obesity and insulin metabolism. In this study, we determined a role of preproghrelin Leu72Met polymorphism on type 2 diabetes mellitus and its relationship to variables studied. Genotypes were assessed by polymerase chain reaction. Frequencies of the Leu72Met polymorphism were found to be 35.4% in the type 2 diabetic patients and 32.5% in the normal controls. The Leu72Met polymorphism was not associated with hypertension, macroangiopathy, retinopathy, serum cholesterol, triglyceride, blood urea nitrogen, HbA(1c), lipoprotein (a), fasting insulin, or 24-hour urinary protein levels in the type 2 diabetic group. However, the Leu72Met polymorphism was clearly associated with serum creatinine levels in the diabetic group, as the Met72 carriers exhibited lower serum creatinine levels than the Met72 noncarriers. Our data indicate that the preproghrelin Leu72Met polymorphism is not associated with type 2 diabetes mellitus. However, the Leu72Met polymorphism is associated with serum creatinine levels. These data suggest that Met72 carrier status may be a predictable marker for diabetic nephropathy or renal impairment in type 2 diabetes mellitus.

  4. Association of ghrelin Leu72Met polymorphism with type 2 diabetes mellitus in Chinese population.

    Science.gov (United States)

    Liu, Jing; Liu, Jia; Tian, Li-min; Liu, Ju-xiang; Bing, Ya-jun; Zhang, Ji-ping; Wang, Yun-Fang; Zhang, Lu-yan

    2012-08-10

    Ghrelin, a novel endogenous ligand for the growth hormone secretagogue receptor, is considered to implicate the development of the type 2 diabetes mellitus (T2DM). The Leu72Met (+408C>A) polymorphism of the preproghrelin, has been linked to obesity, insulin resistance and diabetes. To investigate the distribution of ghrelin gene Leu72Met polymorphism and its association with the type 2 diabetes mellitus in Chinese population. We conducted a case-control study on 877 patients with T2DM and 864 controls, which were genotyped by the polymerase chain reaction (PCR) technique, denaturing high performance liquid chromatography (DHPLC) and DNA sequence analysis. Laboratory analyses were carried out in the hospital laboratory. No significant difference in the Leu72Met genotype distributions and allele frequency was observed between type 2 diabetes mellitus and controls (both P>0.05). The polymorphism was not associated with T2DM. However, among the T2DM group, the patients carrying Leu72Leu genotype had significantly increased levels of FPG and serum creatinine compared with variant genotypes (Leu72Met and Met72Met) (Ppolymorphism of the preproghrelin gene was not associated with T2DM in Chinese population. However, it may have some roles in the etiology of insulin resistance. Copyright © 2012 Elsevier B.V. All rights reserved.

  5. Fluxes at experiment facilities in HEU and LEU designs for the FRM-II

    International Nuclear Information System (INIS)

    Hanan, N. A.

    1998-01-01

    An Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm 3 and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime(50 days) and the same neutron flux performance (8 x 10 14 n/cm 2 -s in the reflector). LEU silicide fuel with 4.5 g/cm 3 has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. Several issues that were raised by TUM have been addressed in Refs. 1-3. The conclusions of these analyses are summarized below. This paper addresses four additional issues that have been raised in several forums, including Ref 4: heat generation in the cold neutron source (CNS), the gamma and fast neutron fluxes which are components of the reactor noise in neutron scattering experiments in the experiment hall of the reactor, a fuel cycle length difference, and the reactivity worth of the beam tubes and other experiment facilities. The results show that: (a) for the same thermal neutron flux, the neutron and gamma heating in the CNS is smaller in the LEU design than in the HEU design, and cold neutron fluxes as good or better than those of the HEU design can be obtained with the LEU design; (b) the gamma and fast neutron components of the reactor noise in the experiment hall are about the same in both designs; (c) the fuel cycle length is 50 days for both designs; and (d) the absolute value of the reactivity worth of the beam tubes and other experiment facilities is smaller in the LEU design, allowing its fuel cycle length to be increased to 53 or 54 days. Based on the excellent results for the Alternative LEU Design that were obtained in all analyses, the RERTR Program reiterates its conclusion that there are no major technical

  6. Experimental irradiation of UMo fuel: Pie results and modeling of fuel behaviour

    International Nuclear Information System (INIS)

    Languille, A.; Plancq, D.; Huet, F.; Guigon, B.; Lemoine, P.; Sacristan, P.; Hofman, G.; Snelgrove, J.; Rest, J.; Hayes, S.; Meyer, M.; Vacelet, H.; Leborgne, E.; Dassel, G.

    2002-01-01

    Seven full-sized U Mo plates containing ca. 8 g/cm 3 of uranium in the fuel meat have been irradiated since the beginning of the French U Mo development program. The first three of them with 20% 235 U enrichment were irradiated at maximum surfacic power under 150 W/cm 2 in the OSIRIS reactor up to 50% burn-up and are under examination. Their global behaviour is satisfactory: no failure and a low swelling. The other four plates were irradiated in the HFR Petten at maximum surfacic power between 150 and 250 W/cm 2 with two enrichments 20 and 35%. The experiment was stopped after two cycles due to a fuel failure. The post- irradiation examinations were completed in 2001 in Petten. Examinations showed a correct behaviour of 20% enriched plates and an abnormal behaviour of the two other plates (35%-enriched) with a clad failure on the plate 4. The fuel failure appears to result from a combination of factors that led to high corrosion cladding and high fuel meat temperatures. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-01

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

  8. 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

  9. Standardization of specifications and inspection procedures for LEU plate-type research reactor fuels

    International Nuclear Information System (INIS)

    1988-06-01

    With the transition to high density uranium LEU fuel, fabrication costs of research reactor fuel elements have a tendency to increase because of two reasons. First, the amount of the powder of the uranium compound required increases by more than a factor of five. Second, fabrication requirements are in many cases nearer the fabrication limits. Therefore, it is important that measures be undertaken to eliminate or reduce unnecessary requirements in the specification or inspection procedures of research reactor fuel elements utilizing LEU. An additional stimulus for standardizing specifications and inspection procedures at this time is provided by the fact that most LEU conversions will occur within a short time span, and that nearly all of them will require preparation of new specifications and inspection procedures. In this sense, the LEU conversions offer an opportunity for improving the rationality and efficiency of the fuel fabrication and inspection processes. This report focuses on the standardization of specifications and inspection processes of high uranium density LEU fuels for research reactors. However, in many cases the results can also be extended directly to other research reactor fuels. 15 refs, 1 fig., 3 tabs

  10. Substrate-modulated unwinding of transmembrane helices in the NSS transporter LeuT.

    Science.gov (United States)

    Merkle, Patrick S; Gotfryd, Kamil; Cuendet, Michel A; Leth-Espensen, Katrine Z; Gether, Ulrik; Loland, Claus J; Rand, Kasper D

    2018-05-01

    LeuT, a prokaryotic member of the neurotransmitter:sodium symporter (NSS) family, is an established structural model for mammalian NSS counterparts. We investigate the substrate translocation mechanism of LeuT by measuring the solution-phase structural dynamics of the transporter in distinct functional states by hydrogen/deuterium exchange mass spectrometry (HDX-MS). Our HDX-MS data pinpoint LeuT segments involved in substrate transport and reveal for the first time a comprehensive and detailed view of the dynamics associated with transition of the transporter between outward- and inward-facing configurations in a Na + - and K + -dependent manner. The results suggest that partial unwinding of transmembrane helices 1/5/6/7 drives LeuT from a substrate-bound, outward-facing occluded conformation toward an inward-facing open state. These hitherto unknown, large-scale conformational changes in functionally important transmembrane segments, observed for LeuT in detergent-solubilized form and when embedded in a native-like phospholipid bilayer, could be of physiological relevance for the translocation process.

  11. 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

  12. Preliminary LEU fuel cycle analyses for the Belgian BR2 reactor

    International Nuclear Information System (INIS)

    Deen, J.R.; Snelgrove, J.L.

    1986-01-01

    Fuel cycle calculations have been performed with reference HEU fuel and LEU fuel using Cd wires or boron as burnable absorbers. The 235 U content in the LEU element has increased 20% to 480g compared to the reference HEU element. The number of fuel plates has remained unchanged while the fuel meat thickness has increased to 0.76 mm from 0.51 mm. The LEU meat density is 5.1 Mg U/m 3 . The reference fuel cycle was a 31 element core operating at 56 MW with a 19.8 day cycle length and eight fresh elements loaded per cycle. Comparable fuel cycle characteristics can be achieved using the proposed LEU fuel element with either Cd wires or boron burnable absorbers. The neutron flux for E/sub n/ > 1 eV changes very little (<5%) in LEU relative to HEU cores. Thermal flux reductions are 5 to 10% in non-fueled positions, and 20 to 30% in fuel elements

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  14. Irradiation of full size UMo plates

    International Nuclear Information System (INIS)

    Vacelet, H.; Lavastre, Y.; Grasse, M.; Sacristan, P.; Languille, A.

    1999-01-01

    An important development program for a UMo MTR fuel has been launched in France. The goal of the French working group is to develop a high performing and reprocessable fuel before the end of the US return policy. This paper is focussed on the fabrication of full-sized UMo plates with LEU (Low Enriched Enrichment) and their irradiation in OSIRIS reactor which was started on the 22nd of September. The results of the plates inspection are presented here as well as the irradiation conditions. (author)

  15. Preliminary Multiphysics Analyses of HFIR LEU Fuel Conversion using COMSOL

    Energy Technology Data Exchange (ETDEWEB)

    Freels, James D [ORNL; Bodey, Isaac T [ORNL; Arimilli, Rao V [ORNL; Curtis, Franklin G [ORNL; Ekici, Kivanc [ORNL; Jain, Prashant K [ORNL

    2011-06-01

    4 of this report. The HFIR LEU conversion project has also obtained the services of Dr. Prashant K. Jain of the Reactor & Nuclear Systems Division (RNSD) of ORNL. Prashant has quickly adapted to the COMSOL tools and has been focusing on thermal-structure interaction (TSI) issues and development of alternative 3D model approaches that could yield faster-running solutions. Prashant is the primary contributor to Section 5 of the report. And finally, while incorporating findings from all members of the COMSOL team (i.e., the team) and contributing as the senior COMSOL leader and advocate, Dr. James D. Freels has focused on the 3D model development, cluster deployment, and has contributed primarily to Section 3 and overall integration of this report. The team has migrated to the current release of COMSOL at version 4.1 for all the work described in this report, except where stated otherwise. Just as in the performance of the research, each of the respective sections has been originally authored by the respective authors. Therefore, the reader will observe a contrast in writing style throughout this document.

  16. 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)

  17. 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)

  18. 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

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

    International Nuclear Information System (INIS)

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

    1986-11-01

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

  20. 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

  1. Status of fuel irradiation tests in HANARO

    International Nuclear Information System (INIS)

    Kim, Hark Rho; Lee, Choong Sung; Lee, Kye Hong; Jun, Byung Jin; Lee, Ji Bok

    1999-01-01

    Since 1996 after finishing the long-term operational test, HANARO (High-Flux Advanced Neutron Application Reactor) has been extensively used for material irradiation tests, beam application research, radioisotope production and neutron activation analysis. This paper presents the fuel irradiation test activities which are now conducted or have been finished in HANARO. KAERI developed LEU fuel using an atomization method for the research reactors. Using this LEU, we have set up and conducted three irradiation programs: (1) medium power irradiation test using a short-length mini-assembly made of 3.15 gU/cc U 3 Si, (2) high power irradiation tests using full-length test assemblies made of 3.15 gU/cc U 3 Si, and (3) irradiation test using a short-length mini-plate made of 4.8 gU/cc U 3 Si 2 . DUPIC (Direct Use of spent PWR fuels in CANDU Reactors) simulation fuel pellets, of which compositions are very similar to DUPIC pellets to keep the similarity in the thermo-mechanical property, were developed. Three mini-elements including 5 pellets each were installed in a capsule. This capsule has been irradiated for 2 months and unloaded from the HANARO core at the end of September 1999. Another very important test is the HANARO fuel qualification program at high power, which is required to resolve the licensing issue. This test is imposed on the HANARO operation license due to insufficient test data under high power environment. To resolve this licensing issue, we have been carrying out the required irradiation tests and PIE (Post-irradiation Examination) tests. Through this program, it is believed that the resolution of the licensing issue is achieved. In addition to these programs, several fuel test plans are under way. Through these vigorous activities of fuel irradiation test programs, HANARO is sure to significantly contribute to the national nuclear R and D programs. (author)

  2. Development of Techniques for Small Scale Indigenous 99Mo Production Using LEU Targets at ICN Pitesti-Romania [Country report: Romania

    International Nuclear Information System (INIS)

    2015-01-01

    Initiation of the IAEA Coordinated Research Project (CRP) “Development Techniques for Small Scale Indigenous 99 Mo Production Using LEU Fission or Neutron Activation” during 2005 allowed Member States to participate through their research organization on contractor arrangement to accomplish the CRP objectives. Among these, the participating research organization Institute for Nuclear Research Pitesti Romania (ICN), was the beneficiary of financial support and Argonne National Laboratory assistance provided by US Department of Energy to the CRP for development of techniques for fission 99 Mo production based on LEU modified CINTICHEM process. The Agency’s role in this field was to assist in the transfer and adaptation of existing technology in order to disseminate a technique, which advances international non-proliferation objectives and promotes sustainable development needs, while also contributing to extend the production capacity for addressing supply shortages from the latest years. The Institute for Nuclear Research, considering the existing good conditions of infrastructure of the research reactor with suitable irradiation conditions for radioisotopes, a post irradiation laboratory with direct transfer of irradiated targets from the reactor and handling of high radioactive sources, and simultaneously the existence of an expanding internal market, decided to undertake the necessary steps in order to produce fission molybdenum. The Institute intends to develop the capability to respond to the domestic needs in cooperation with the IFINN–HH from Bucharest, which is able to perform the last step consisting in the loading of fission molybdenum on chromatography generators and dispensing to the final client. The primary scope of the project is the development of the necessary technological steps and chemical processing steps in order to be able to cover the entire process for fission molybdenum production at the required standard of purity

  3. 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

  4. A neutronics study of LEU fuel options for the HFR-Petten

    International Nuclear Information System (INIS)

    Deen, J.R.; Snelgrove, J.L.

    1985-01-01

    The standard HEU fuel cycle characteristics are compared with those of several different LEU fuel cycles in the new vessel configuration. The primary design goals were to provide similar reactivity performance and neutron flux profiles with a minimal increase in 235 U loading. The fuel cycle advantages of Cd burnable absorbers over 10 B are presented. The LEU fuel cycle requirements were calculated also for an extended 32-day cycle and for a reload batch size reduction from six to five standard elements for the standard 26-day cycle. The effects of typical in-core experiments upon neutron flux profiles and fuel loading requirements are also presented. (author)

  5. Neutronic feasibility studies for LEU conversion of the HFR Petten reactor

    International Nuclear Information System (INIS)

    Hanan, N.A.; Deen, J.R.; Matos, J.E.; Hendriks, J.A.; Thijssen, P.J.M.; Wijtsma, F.J.

    2000-01-01

    Design and safety analyses to determine an optimum LEU fuel assembly design using U 3 Si 2 -Al fuel with up to 4.8 g/cm 3 for conversion of the HFR Petten reactor were performed by the RERTR program in cooperation with the Joint Research Centre and NRG. Credibility of the calculational methods and models were established by comparing calculations with recent measurements by NRG for a core configuration set up for this purpose. This model and methodology were then used to study various LEU fissile loading and burnable poison options that would satisfy specific design criteria. (author)

  6. The Pai-associated leuX specific tRNA5(Leu) affects type 1fimbriation in pathogenic Escherichia coli by control of FimB recombinase expression

    DEFF Research Database (Denmark)

    Ritter, A.; Gally, D.; Olsen, Peter Bjarke

    1997-01-01

    The uropathogenic Escherichia coli strain 536 (06:K15:H31) carries two large chromosomalpathogenicity islands (Pais). Both Pais are flanked by tRNA genes. Spontaneous deletion of Pai IIresults in truncation of the leuX tRNA5Leu gene. This tRNA is required for the expression of type 1fimbriae (Fim...

  7. TEM characterization of irradiated U-7Mo/Mg dispersion fuel

    Science.gov (United States)

    Gan, J.; Keiser, D. D.; Miller, B. D.; Jue, J. F.; Robinson, A. B.; Madden, J.

    2017-10-01

    This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 1021 f/cm3, 7.4 × 1014 f/cm3/s and 123 °C, and 5.5 × 1021 f/cm3, 11.0 × 1014 f/cm3/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al3Mg2 and Al12Mg17 along with precipitates of MgO, Mg2Si and FeAl5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rest, J. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)], E-mail: jrest@anl.gov; Hofman, G.L.; Kim, Yeon Soo [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2009-04-15

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

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

    Science.gov (United States)

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

    2009-04-01

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

  10. Uranium briquettes for irradiation target

    Energy Technology Data Exchange (ETDEWEB)

    Saliba-Silva, Adonis Marcelo; Garcia, Rafael Henrique Lazzari; Martins, Ilson Carlos; Carvalho, Elita Fontenele Urano de; Durazzo, Michelangelo, E-mail: saliba@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    Direct irradiation on targets inside nuclear research or multiple purpose reactors is a common route to produce {sup 99}Mo-{sup 99m}Tc radioisotopes. Nevertheless, since the imposed limits to use LEU uranium to prevent nuclear armament production, the amount of uranium loaded in target meats has physically increased and new processes have been proposed for production. Routes using metallic uranium thin film and UAl{sub x} dispersion have been used for this purpose. Both routes have their own issues, either by bringing difficulties to disassemble the aluminum case inside hot cells or by generating great amount of alkaline radioactive liquid rejects. A potential route might be the dispersion of powders of LEU metallic uranium and nickel, which are pressed as a blend inside a die and followed by pulse electroplating of nickel. The electroplating provides more strength to the briquettes and creates a barrier for gas evolution during neutronic disintegration of {sup 235}U. A target briquette platted with nickel encapsulated in an aluminum case to be irradiated may be an alternative possibility to replace other proposed targets. This work uses pulse Ni-electroplating over iron powder briquette to simulate the covering of uranium by nickel. The following parameters were applied 10 times for each sample: 900Hz, -0.84A/square centimeters with duty cycle of 0.1 in Watts Bath. It also presented the optical microscopy analysis of plated microstructure section. (author)

  11. Uranium briquettes for irradiation target

    International Nuclear Information System (INIS)

    Saliba-Silva, Adonis Marcelo; Garcia, Rafael Henrique Lazzari; Martins, Ilson Carlos; Carvalho, Elita Fontenele Urano de; Durazzo, Michelangelo

    2011-01-01

    Direct irradiation on targets inside nuclear research or multiple purpose reactors is a common route to produce 99 Mo- 99m Tc radioisotopes. Nevertheless, since the imposed limits to use LEU uranium to prevent nuclear armament production, the amount of uranium loaded in target meats has physically increased and new processes have been proposed for production. Routes using metallic uranium thin film and UAl x dispersion have been used for this purpose. Both routes have their own issues, either by bringing difficulties to disassemble the aluminum case inside hot cells or by generating great amount of alkaline radioactive liquid rejects. A potential route might be the dispersion of powders of LEU metallic uranium and nickel, which are pressed as a blend inside a die and followed by pulse electroplating of nickel. The electroplating provides more strength to the briquettes and creates a barrier for gas evolution during neutronic disintegration of 235 U. A target briquette platted with nickel encapsulated in an aluminum case to be irradiated may be an alternative possibility to replace other proposed targets. This work uses pulse Ni-electroplating over iron powder briquette to simulate the covering of uranium by nickel. The following parameters were applied 10 times for each sample: 900Hz, -0.84A/square centimeters with duty cycle of 0.1 in Watts Bath. It also presented the optical microscopy analysis of plated microstructure section. (author)

  12. 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.)

  13. Pilot plant production at Riso of LEU silicide fuel for the Danish reactor DR3

    International Nuclear Information System (INIS)

    Toft, P.; Borring, J.; Adolph, E.

    1988-01-01

    A pilot plant for fabricating LEU silicide fuel elements has been established at Riso National Laboratory. Three test elements for the Danish reactor DR3 have been fabricated, based on 19.88% enriched U 3 Si 2 powder that has been purchased elsewhere. The pilot plant has been set up and 3 test elements fabricated without any major difficulties

  14. Neutronic analysis of the JMTR with LEU fuel and burnable poison

    International Nuclear Information System (INIS)

    Nagaoka, Yoshiharu; Oyamada, Rokuro; Matos, J.E.; Woodruff, W.L.

    1985-01-01

    The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed (author)

  15. Neutronic analysis of the JMTR with LEU fuel and burnable poison

    Energy Technology Data Exchange (ETDEWEB)

    Nagaoka, Yoshiharu; Oyamada, Rokuro [Japan Atomic Energy Research Institute, Oarai-machi Ibaraki-ken (Japan); Matos, J E; Woodruff, W L [Argonne National Laboratory, Argonne, IL (United States)

    1985-07-01

    The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed (author)

  16. Infantile presentation of the mtDNA A3243G tRNA(Leu (UUR)) mutation.

    NARCIS (Netherlands)

    Okhuijsen-Kroes, E.J.; Trijbels, J.M.F.; Sengers, R.C.A.; Mariman, E.C.M.; Heuvel, L.P.W.J. van den; Wendel, U.A.H.; Koch, G.; Smeitink, J.A.M.

    2001-01-01

    Mitochondrial DNA (mtDNA) disorders are clinically very heterogeneous, ranging from single organ involvement to severe multisystem disease. One of the most frequently observed mtDNA mutations is the A-to-G transition at position 3243 of the tRNA(Leu (UUR)) gene. This mutation is often related to

  17. 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)

  18. Vibrational absorption spectra, DFT and SCC-DFTB conformational study and analysis of [Leu]enkephalin

    DEFF Research Database (Denmark)

    Abdali, Salim; Niehaus, T.A.; Jalkanen, Karl J.

    2003-01-01

    . Ab initio (DFT at the B3LYP/6-31G* level of theory) and semi-empirical (SCC-DFTB) with and without dispersion correction were applied to simulate the VA spectra of [Leu] enkephalin. In these calculations structures taken from X-ray measurements for different conformers of the molecule were used...

  19. Conformational determination of [Leu]enkephalin based on theoretical and experimental VA and VCD spectral analyses

    DEFF Research Database (Denmark)

    Abdali, Salim; Jalkanen, Karl J.; Cao, X.

    2004-01-01

    Conformational determination of [Leu]enkephalin in DMSO-d6 is carried out using VA and VCD spectral analyses. Conformational energies, vibrational frequencies and VA and VCD intensities are calculated using DFT at B3LYP/6-31G* level of theory. Comparison between the measured spectra...

  20. 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)

  1. Energy levels and quantum states of [Leu]enkephalin conformations based on theoretical and experimental investigations

    DEFF Research Database (Denmark)

    Abdali, Salim; Jensen, Morten Østergaard; Bohr, Henrik

    2003-01-01

    This paper describes a theoretical and experimental study of [Leu]enkephalin conformations with respect to the quantum estates of the atomic structure of the peptide. Results from vibrational absorption measurements and quantum calculations are used to outline a quantum picture and to assign vibr...

  2. Foreign research reactor spent nuclear fuel inventories containing HEU and LEU of US-origin

    International Nuclear Information System (INIS)

    Matos, J.E.

    1995-01-01

    This paper provides estimates of the quantities and types of foreign research reactor spent nuclear fuel containing HEU and LEU of US-origin that are anticipated during the period beginning in January 1996 and extending for 10-15 years

  3. Simple and effective procedure for conformational search of macromolecules. Application to Met- and Leu-Enkephalin

    Energy Technology Data Exchange (ETDEWEB)

    Meirovitch, H.; Meirovitch, E. (Florida State Univ., Tallahassee, FL (United States)); Michel, A.G. (Institut de Recherches Serrier, Suresnes (France)); Vasquez, M. (Protein Design Lab., Mountain View, CA (United States))

    1994-06-23

    A simple and efficient method for searching the conformational space of macromolecules is presented. With this method an initial set of relatively low-energy structures is generated, and their energies are further minimized with a procedure that enables escaping from local energy minima. Illustrative calculations are described for Met- and Leu-enkephalin. 37 refs., 1 tab.

  4. Factor XIII Val34Leu is a genetic factor involved in the etiology of venous thrombosis

    NARCIS (Netherlands)

    Franco, R. F.; Reitsma, P. H.; Lourenço, D.; Maffei, F. H.; Morelli, V.; Tavella, M. H.; Araújo, A. G.; Piccinato, C. E.; Zago, M. A.

    1999-01-01

    A mutation in the factor XIII gene (FXIII Val34Leu) gene was recently reported to confer protection against myocardial infarction, but its relationship with venous thrombosis is unknown. In addition, a mutation in the 5'-untranslated region of the FXII gene (46 C->T) was identified which is

  5. Intercomparison of rod-worth measurement techniques in a LEU-HTR assembly

    International Nuclear Information System (INIS)

    Williams, T.; Chawla, R.

    1994-01-01

    The measurement of absorber-rod worths in the radial reflector of a LEU-HTR pebble bed system is described. Particular emphasis is placed on the choice of complementary measurement techniques to ensure that sensitivities to systematic errors in the calculated parameters used in the analysis are minimised. (author) 3 figs., 3 tabs., 8 refs

  6. 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)

  7. 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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  9. Behaviour of uranium under irradiation; Comportement de l'uranium sous irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Adda, Y; Mustelier, J P; Quere, Y [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-07-01

    The main results obtained in a study of the formation of defects caused in uranium by fission at low temperature are reported. By irradiation at 20 K. it was possible to determine the number of Frenkel pairs produced by one fission. An analysis of the curves giving the variations in electrical resistivity shows the size of the displacement spikes and the mechanism of defect creation due to fission. Irradiations at 77 K gave additional information, showing behaviour differences in the case of recrystallised and of cold worked uranium. The diffusion of rare gases was studied using metal-rare gas alloys obtained by electrical discharge, and samples of irradiated uranium. Simple diffusion is only responsible for the release of the rare gases under vacuum in cases where the rare gas content is very low (very slightly irradiated U). On the other hand when the concentration is higher (samples prepared by electrical discharge) the gas is given off by the formation, growth and coalescence of bubbles; the apparent diffusion coefficient is then quite different from the true coefficient and cannot be used in calculations on swelling. The various factors governing the phenomenon of simple diffusion were examined. It was shown in particular that a small addition of molybdenum could reduce the diffusion coefficient by a factor of 100. The precipitation of gas in uranium (Kr), in silver (Kr) and in Al-Li alloy (He) have been followed by measurement of the crystal parameter and of the electrical resistivity, and by electron microscope examination of thin films. The important part played by dislocations in the generation and growth of bubbles has been demonstrated, and it has been shown also that precipitation of bubbles on the dislocation lattice could block the development of recrystallisation. The results of these studies were compared with observations made on the swelling of uranium and uranium alloys U Mo and U Nb strongly irradiated between 400 and 700 C. In the case of Cubic

  10. [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3, small molecule synthetic peptide leptin mimetics, improve glycemic control in diet-induced obese (DIO) mice.

    Science.gov (United States)

    Wang, Anke; Anderson, Brian M; Novakovic, Zachary M; Grasso, Patricia

    2018-03-01

    We have previously shown that following oral delivery in dodecyl maltoside (DDM), [D-Leu-4]-OB3 and its myristic acid conjugate, MA-[D-Leu-4]-OB3, improved energy balance and glucose homeostasis in genetically obese/diabetic mouse models. More recently, we have provided immunohistochemical evidence indicating that these synthetic peptide leptin mimetics cross the blood-brain barrier and concentrate in the area of the arcuate nucleus of the hypothalamus in normal C57BL/6J and Swiss Webster mice, in genetically obese ob/ob mice, and in diet-induced obese (DIO) mice. In the present study, we describe the effects of oral delivery of [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3 on glycemic control in diet-induced (DIO) mice, a non-genetic rodent model of obesity and its associated insulin resistance, which more closely recapitulates common obesity and diabetes in humans. Male C57BL/6J and DIO mice, 17, 20, and 28 weeks of age, were maintained on a low-fat or high-fat diet and given vehicle (DDM) alone or [D-Leu-4]-OB3 or MA-[D-Leu-4]-OB3 in DDM by oral gavage for 12 or 14 days. Body weight gain, food and water intake, fasting blood glucose, oral glucose tolerance, and serum insulin levels were measured. Our data indicate that (1) [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3 restore glucose tolerance in male DIO mice maintained on a high-fat diet to levels comparable to those of non-obese C57BL/6J wild-type mice of the same age and sex maintained on a low-fat diet; and (2) the influence of [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3 on glycemic control appears to be independent of their effects on energy balance. These results suggest that [D-Leu-4]-OB3 and/or MA-[D-Leu-4]-OB3 may have application to the management of the majority of cases of common obesity in humans, a state characterized at least in part, by leptin resistance resulting from a defect in leptin transport across the blood-brain barrier. They further suggest that these small molecule synthetic peptide leptin mimetics, through their

  11. Effects of ionizing radiation and pretreatment with [D-Leu6,des-Gly10] luteinizing hormone-releasing hormone ethylamide on developing rat ovarian follicles

    International Nuclear Information System (INIS)

    Jarrell, J.; YoungLai, E.V.; McMahon, A.; Barr, R.; O'Connell, G.; Belbeck, L.

    1987-01-01

    To assess the effects of a gonadotropin-releasing hormone agonist, [D-Leu6,des-Gly10] luteinizing hormone-releasing hormone ethylamide, in ameliorating the damage caused by ionizing radiation, gonadotropin-releasing hormone agonist was administered to rats from day 22 to 37 of age in doses of 0.1, 0.4, and 1.0 microgram/day or vehicle and the rats were sacrificed on day 44 of age. There were no effects on estradiol, progesterone, luteinizing, or follicle-stimulating hormone, nor an effect on ovarian follicle numbers or development. In separate experiments, rats treated with gonadotropin-releasing hormone agonist in doses of 0.04, 0.1, 0.4, or 1.0 microgram/day were either irradiated or sham irradiated on day 30 and all groups sacrificed on day 44 of age. Irradiation produced a reduction in ovarian weight and an increase in ovarian follicular atresia. Pretreatment with the agonist prevented the reduction in ovarian weight and numbers of primordial and preantral follicles but not healthy or atretic antral follicles. Such putative radioprotection should be tested on actual reproductive performance

  12. Characterization of an irradiated RERTR-7 fuel plate using transmission electron microscopy

    International Nuclear Information System (INIS)

    Gan, J.; Keiser, D.D. Jr.; Miller, B.D.; Robinson, A.B.; Medvedev, P.

    2010-01-01

    Transmission electron microscopy (TEM) has been used to characterize an irradiated fuel plate with Al-2Si matrix from the Reduced Enrichment Research and Test Reactor RERTR-7 experiment that was irradiated under moderate reactor conditions. The results of this work showed the presence of a bubble superlattice within the U-7Mo grains that accommodated fission gases (e.g., Xe). The presence of this structure helps the U-7Mo exhibit a stable swelling behaviour during irradiation. Furthermore, TEM analysis showed that the Si-rich interaction layers that develop around the fuel particles at the U-7Mo/matrix interface during fuel plate fabrication and irradiation become amorphous during irradiation. An important question that remains to be answered about the irradiation behaviour of U-Mo dispersion fuels is how do more aggressive irradiation conditions affect the behaviour of fission gases within the U-7Mo fuel particles and in the amorphous interaction layers on the microstructural scale that can be characterized using TEM? This paper will discuss the results of TEM analysis that was performed on a sample taken from an irradiated RERTR-7 fuel plate with Al-2Si matrix. This plate was exposed to more aggressive irradiation conditions than the RERTR-6 plate. The microstructural features present within the U-7Mo and the amorphous interaction layers will be discussed. The results of this analysis will be compared to what was observed in the earlier RERTR-6 fuel plate characterization. (author)

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  14. A comparison of the radiological consequences of a HEU and LEU fueled research reactor

    International Nuclear Information System (INIS)

    Kollas, J.G.

    1985-01-01

    An analysis of the design basis accident radiological consequences of the HEU and LEU fueled Greek Research Reactor is presented. Doses and individual cancer risk from exposure to the passing radioactive cloud are estimated up to a distance of 20 km from the reactor site. Collective exposure and latent health effects are estimated for the total Athens area of 3081000 inhabitants. The results indicate that the plutonium isotopes buildup in the LEU fuel does not increase appreciably the consequences in respect to the HEU fueled reactor. The plutonium impact concerns mainly bone effects and secondly lung and whole body effects. The contribution to the limiting thyroid dose and the corresponding thyroid effects is insignificant. (author)

  15. Study on Al-alloy or silicide LEU for DR3 in Denmark

    Energy Technology Data Exchange (ETDEWEB)

    Haack, Karsten [Riso National Laboratory, DK 4000 Roskilde (Germany)

    1985-07-01

    The 10 MW D{sub 2}0-moderated and -cooled research reactor DR3 has at present HEU fuel available for continued operation till early 19. This report presents the status of a feasibility study prepared for selection of the best suited candidate LEU fuel type for DR3 at a potential conversion in 1988. At the moment two alternatives are evaluated: UAl-alloy with modified geometry and U{sub 3}Si{sub 2} with unchanged geometry. A decision on the type selected for further investigation is expected late 1984. The investigation should comprise development, in- and out-of-pile--testing and licensing activities on the potential LEU option. (author)

  16. Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel

    International Nuclear Information System (INIS)

    Deen, J.R.; Snelgrove, J.L.; Papastergiou, C.

    1992-01-01

    The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed

  17. 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.

  18. 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)

  19. 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)

  20. Alternating access mechanisms of LeuT-fold transporters: trailblazing towards the promised energy landscapes.

    Science.gov (United States)

    Kazmier, Kelli; Claxton, Derek P; Mchaourab, Hassane S

    2017-08-01

    Secondary active transporters couple the uphill translocation of substrates to electrochemical ion gradients. Transporter conformational motion, generically referred to as alternating access, enables a central ligand binding site to change its orientation relative to the membrane. Here we review themes of alternating access and the transduction of ion gradient energy to power this process in the LeuT-fold class of transporters where crystallographic, computational and spectroscopic approaches have converged to yield detailed models of transport cycles. Specifically, we compare findings for the Na + -coupled amino acid transporter LeuT and the Na + -coupled hydantoin transporter Mhp1. Although these studies have illuminated multiple aspects of transporter structures and dynamics, a number of questions remain unresolved that so far hinder understanding transport mechanisms in an energy landscape perspective. Copyright © 2016 Elsevier Ltd. All rights reserved.

  1. 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

  2. Further data of silicide fuel for the LEU conversion of JMTR

    International Nuclear Information System (INIS)

    Saito, M.; Futamura, Y.; Nakata, H.; Ando, H.; Sakurai, F.; Ooka, N.; Sakakura, A.; Ugajin, M.; Shirai, E.

    1990-01-01

    Silicide fuel data for the safety assessment of the JMTR LEU fuel conversion are being measured. The data include fission product release, thermal properties, behaviour under accident conditions, and metallurgical characteristics. The methods used in the experiments are discussed. Results of fission products release at high temperature are described. The release of iodine from the silicide fuel is considerably lower than for U-Al alloy fuel

  3. Natural convection cooling of LEU cores for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Khan, L.A.; Bokhari, I.H.; Akhtar, K.M.

    1991-08-01

    The first high power and equilibrium LEU cores of PARR-1 have been analysed to assess the maximum operating power based on natural convection cooling, need for forced cooling to remove the decay heat and to estimate safety margins that commensurate with the predetermined power limit. Computer code NATCON and standard correlations have been used for the analysis. The parameters studied includes coolant velocity, temperature distribution in the core, heat fluxes at onset of nucleate boiling, pulsed boiling and burnup. (author)

  4. Simultaneous extraction of. beta. -endorphin and leu- and met-enkephalins from human and rat plasma

    Energy Technology Data Exchange (ETDEWEB)

    Bhathena, S.J.; Smith, P.M.; Kennedy, B.W. (Dept. of Agriculture, Beltsville, MD (USA)); Voyles, N.R.; Recant, L. (Diabetes Research Laboratory, Washington, DC (USA))

    1989-01-01

    A simple, rapid and reliable procedure is described to simultaneously concentrated and purify {beta}-endorphin, leu-and met-enkephalins from small volumes of human and rat plasma before radioimmunoassay is performed. It uses C{sub 18} Sep-Pak reverse phase cartridges. The effectiveness of different protease inhibitors in preventing degradation of opiates by plasma and different solvent systems for eluting opiates is also evaluated.

  5. Techno-economic study on conversion of SAFARI-1 to LEU silicide fuel

    International Nuclear Information System (INIS)

    Ball, G.; Malherbe, F.J.

    2004-01-01

    This paper marks the conclusion of the techno-economic study into the conversion of SAFARI-1 reactor in South Africa to LEU silicide fuel. Several different fuel types were studied and their characteristics compared to the current HEU fuel. The technical feasibility of operating SAFARI-1 with the different fuels as well as the overall economic impact of the fuels is discussed and conclusions drawn.(author)

  6. The beginning of the LEU fuel elements manufacturing in the Chilean Commission of Nuclear Energy

    International Nuclear Information System (INIS)

    Contreras, H.; Chavez, J.C.; Marin, J.; Lisboa, J.; Olivares, L.; Jimenez, O.

    1998-01-01

    The U 3 Si 2 LEU fuel fabrication program at CCHEN has started with the assembly of four leaders fuel elements for the RECH-1 reactor. This activity has involved a stage of fuel plates qualification, to evaluate fabrication procedures and quality controls and quality assurance. The qualification extent was 50% of the fuel plates, equivalent to the number of plates required for the assembly of two fuel elements. (author)

  7. Neutronic analysis of the JMTR with LEU fuel and burnable poison

    International Nuclear Information System (INIS)

    Nagaoka, Yoshiharu; Oyamada, Rokuro; Matos, J.E.; Woodruff, W.L.

    1984-01-01

    The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed. 2 refs., 10 figs., 5 tabs

  8. 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)

  9. The Leu72Met polymorphism of the ghrelin gene is significantly associated with binge eating disorder.

    Science.gov (United States)

    Monteleone, Palmiero; Tortorella, Alfonso; Castaldo, Eloisa; Di Filippo, Carmela; Maj, Mario

    2007-02-01

    The pathophysiological mechanisms underlying binge eating disorder are poorly understood. Evidence exists for the fact that abnormalities in peptides involved in the regulation of appetite, including ghrelin, may play a role in binge eating behavior. Genes involved in the ghrelin physiology may therefore contribute to the biological vulnerability to binge eating disorder. We examined whether two polymorphisms of the ghrelin gene, the G152A (Arg51Gln) and C214A (Leu72Met), were associated with binge eating disorder. Ninety obese or nonobese women with binge eating disorder and 119 normal weight women were genotyped at the ghrelin gene. Statistical analyses showed that the Leu72Met ghrelin gene variant was significantly more frequent in binge eating disorder patients (chi2=5.940; d.f.=1, P=0.01) and was associated with a moderate, but significant risk to develop binge eating disorder (odds ratio=2.725, 95% confidence interval: 1.168-6.350). Although these data should be regarded as preliminary because of the small sample size, they suggest that the Leu72Met ghrelin gene variant may contribute to the genetic susceptibility to binge eating disorder.

  10. Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    El-Sahlamy, Neama M. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt)

    2017-11-15

    In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1{sub X}-A1, U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1{sub X}-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U{sub 3}Si{sub 2}-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

  11. Dysregulated autophagy in restrictive cardiomyopathy due to Pro209Leu mutation in BAG3.

    Science.gov (United States)

    Schänzer, A; Rupp, S; Gräf, S; Zengeler, D; Jux, C; Akintürk, H; Gulatz, L; Mazhari, N; Acker, T; Van Coster, R; Garvalov, B K; Hahn, A

    2018-03-01

    Myofibrillary myopathies (MFM) are hereditary myopathies histologically characterized by degeneration of myofibrils and aggregation of proteins in striated muscle. Cardiomyopathy is common in MFM but the pathophysiological mechanisms are not well understood. The BAG3-Pro209Leu mutation is associated with early onset MFM and severe restrictive cardiomyopathy (RCM), often necessitating heart transplantation during childhood. We report on a young male patient with a BAG3-Pro209Leu mutation who underwent heart transplantation at eight years of age. Detailed morphological analyses of the explanted heart tissue showed intracytoplasmic inclusions, aggregation of BAG3 and desmin, disintegration of myofibers and Z-disk alterations. The presence of undegraded autophagosomes, seen by electron microscopy, as well as increased levels of p62, LC3-I and WIPI1, detected by immunohistochemistry and western blot analyses, indicated a dysregulation of autophagy. Parkin and PINK1, proteins involved in mitophagy, were slightly increased whereas mitochondrial OXPHOS activities were not altered. These findings indicate that altered autophagy plays a role in the pathogenesis and rapid progression of RCM in MFM caused by the BAG3-Pro209Leu mutation, which could have implications for future therapeutic strategies. Copyright © 2018 Elsevier Inc. All rights reserved.

  12. 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

  13. Calculation of mixed HEU-LEU cores for the HOR research reactor with the scale code system

    International Nuclear Information System (INIS)

    Leege, P.F.A. de; Gibcus, H.P.M.; Hoogenboom, J.E.; Vries, J.W. de

    1997-01-01

    The HOR reactor of Interfaculty Reactor Institute (IRI), Delft, The Netherlands, will be converted to use low enriched fuel (LEU) assemblies. As there are still many usable high enriched (HEU) fuel assemblies present, there will be a considerable reactor operation time with mixed cores with both HEU and LEU fuel assemblies. At IRI a comprehensive reactor physics code system and evaluated nuclear data is implemented for detailed core calculations. One of the backbones of the IRI code system is the well-known SCALE code system package. Full core calculations are performed with the diffusion theory code BOLD VENTURE, the nodal code SILWER, and the Monte Carlo code KENO Va. Results are displayed of a strategy from a HEU core to a mixed HEU-LEU core and eventually a LEU core. (author)

  14. Development of new ORIGEN2 data library sets for research reactors with light water cooled oxide and silicide LEU (20 w/o) fuels based on JENDL-3.3 nuclear data

    International Nuclear Information System (INIS)

    Liem, Peng Hong; Sembiring, Tagor Malem

    2013-01-01

    Highlights: • We developed new ORIGEN2 data library sets for research reactors based on JENDL-3.3. • The sets cover oxide and silicide LEU fuels with meat density up to 4.74 g U/cm 3 . • Two kinds of data library sets are available: fuel region and non-fuel regions. • We verified the new data library sets with other codes. • We validated the new data library against a non-destructive test. -- Abstract: New sets of ORIGEN2 data library dedicated to research/testing reactors with light water cooled oxide and silicide LEU fuel plates based on JENDL-3.3 nuclear data were developed, verified and validated. The new sets are considered to be an extension of the most recent release of ORIGEN2.2UPJ code, i.e. the ORLIBJ33 library sets. The newly generated ORIGEN2 data library sets cover both oxide and silicide LEU fuels with fuel meat density range from 2.96 to 4.74 g U/cm 3 used in the present and future operation of the Indonesian 30 MWth RSG GAS research reactor. The new sets are expected applicable also for other research/testing reactors which utilize similar fuels or have similar neutron spectral indices. In addition to the traditional ORIGEN2 library sets for fuel depletion analyses in fuel regions, in the new data library sets, new ORIGEN2 library sets for irradiation/activation analyses were also prepared which cover all representative non-fuel regions of RSG GAS such as reflector elements, irradiation facilities, etc. whose neutron spectra are significantly softer than fuel regions. Verification with other codes as well as validation with a non-destructive test result showed promising results where a good agreement was confirmed

  15. Progress of the RERTR program in 2001

    International Nuclear Information System (INIS)

    Travelli, A.

    2002-01-01

    This paper describes the 2001 progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners. Postirradiation examinations of microplates have continued to reveal excellent irradiation behavior of U-Mo dispersion fuels in a variety of compositions and irradiating conditions. Irradiation of two new batches of mini plates of greater sizes was completed in the ATR to investigate the swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g/cm 3 range. Qualification of the U-Mo dispersion fuels has been declared by a patent issue involving KAERI. Test fuel elements with uranium density of 6 g/cm 3 are being fabricated by BWXT and are expected to begin undergoing irradiation in the HFR-Petten reactor around March 2003, with a goal of qualifying this fuel by mid 2005. U-Mo fuel with uranium density of 8-9 g/cm 3 is expected to be qualified by mid 2007. Final irradiation tests of LEU 99 Mo targets in the RAS-GAS reactor at BATAN, in Indonesia, had to be postponed because of the 9/11 attacks, but the results collected to date indicate that these targets will soon be ready for commercial production. Excellent cooperation is also in progress with the CNEA in Argentina, MDSN/ AECL in Canada, and ANSTO in Australia. Irradiation testing of five WWR-M2 tube-type fuel assemblies fabricated by the NZChK and containing LEU UO 2 dispersion fuel was successfully completed within the Russian RERTR program. A new LEU U-Mo pin-type fuel that could be used to convert most Russian-designed research reactors has been developed by VNIJNM and is ready for testing. Four additional shipments containing 822 spent fuel assemblies from foreign research reactors were accepted by the U.S. by September 30, 2001. Altogether, 4'562 spent fuel assemblies from foreign research

  16. Progress of the RERTR program in 2001

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A. [Technology Development Division Argonne National Laboratory, 9700 South Cass Avenue, Argonne, Illinois 60439-4841 (United States)

    2002-07-01

    This paper describes the 2001 progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners. Postirradiation examinations of microplates have continued to reveal excellent irradiation behavior of U-Mo dispersion fuels in a variety of compositions and irradiating conditions. Irradiation of two new batches of mini plates of greater sizes was completed in the ATR to investigate the swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g/cm{sup 3} range. Qualification of the U-Mo dispersion fuels has been declared by a patent issue involving KAERI. Test fuel elements with uranium density of 6 g/cm{sup 3} are being fabricated by BWXT and are expected to begin undergoing irradiation in the HFR-Petten reactor around March 2003, with a goal of qualifying this fuel by mid 2005. U-Mo fuel with uranium density of 8-9 g/cm{sup 3} is expected to be qualified by mid 2007. Final irradiation tests of LEU {sup 99}Mo targets in the RAS-GAS reactor at BATAN, in Indonesia, had to be postponed because of the 9/11 attacks, but the results collected to date indicate that these targets will soon be ready for commercial production. Excellent cooperation is also in progress with the CNEA in Argentina, MDSN/ AECL in Canada, and ANSTO in Australia. Irradiation testing of five WWR-M2 tube-type fuel assemblies fabricated by the NZChK and containing LEU UO{sub 2} dispersion fuel was successfully completed within the Russian RERTR program. A new LEU U-Mo pin-type fuel that could be used to convert most Russian-designed research reactors has been developed by VNIJNM and is ready for testing. Four additional shipments containing 822 spent fuel assemblies from foreign research reactors were accepted by the U.S. by September 30, 2001. Altogether, 4'562 spent fuel

  17. What the difference to use LEU and HEU fuel elements separately or together in a research reactor

    International Nuclear Information System (INIS)

    Kaya, S.; Uestuen, G.

    2005-01-01

    Concerning of nuclear material safety, most of the research reactors are advised to shift from HEU (high enriched-%93 U-235) to LEU (low enriched-%20 U-235) fuel elements. When LEU and HEU fuel elements are to be used together in a research reactor, some design and safety problems are encountered. According to use of the reactor, some research reactors such as MTR type may not show any considerable difference for HEU or LEU fuel elements, but the efficiency of radioisotope production generated by thermal neutron interaction may decrease about twenty-thirty percent when LEU fuel elements are used. Here, fine mesh-sized 3D neutronic analysis of TR-2 research reactor is presented to indicate the arising problem when LEU end HEU fuel elements are used together in a research reactor. Partial thermohydraulic analysis of the reactor is also given to show the betterness of the LEU fuel element design. However, there might be some points that should be noticed for safer operation of plate type fuelled research reactors. (author)

  18. Evaluation of alkaline dissolution of Al 6061 and Al 1050 for the production of Mo-99 from LEU targets

    International Nuclear Information System (INIS)

    Mindrisz, Ana C.; Camilo, Ruth L.; Araujo, Izilda C.; Forbicini, Christina A.L.G. de O.

    2013-01-01

    Since 2008, due to the global crisis in the production of radioisotope 99 Mo, which product of decay, 99m Tc, is the tracer element most often used in nuclear medicine and accounts for about 80% of all diagnostic procedures in vivo. Studies on the alkaline dissolution to obtain 9 9M o from irradiated UAl x -Al LEU targets are under development. Processing time should be minimized, considering the short half-life of 99 Mo and 99m Tc, about 66 h and 6 h, respectively. This makes dissolution time a significant factor in the development of the process. This paper presents the results of alkaline dissolution of 'scraps' of Al 6061 and 1050, used to simulate the dissolution process of UAl x -Al targets. Dissolution time and gas releasing were evaluated using the following alkaline solutions: a) NaOH 3 mol.L -1 and NaNO 3 2 mol.L -1 , b) NaOH 3 mol.L -1 and NaNO 3 4 mol.L -1 . The initial temperature of dissolution was 85 deg C in all cases. Al 6061 showed values of dissolution time greater than that for Al 1050, 25% for NaNO 3 2 mol.L -1 and 104.55% for NaNO 3 4 mol.L -1 . The dissolution with NaNO 3 2 mol.L -1 showed that the gas releasing for Al 6061 was 2.7% greater than for Al 1050. However Al 1050 showed that gas releasing 9.92% greater than for Al 6061 during the dissolution with NaNO 3 4 mol.L -1 . The decision about what type of alloy has to be used, Al 1050 or Al 6061, it will be upto the group that will manufacture the targets for the RMB. (author)

  19. Suppression of pokeweed mitogen-stimulated immunoglobulin production in patients with rheumatoid arthritis after treatment with total lymphoid irradiation

    International Nuclear Information System (INIS)

    Kotzin, B.L.; Strober, S.; Kansas, G.S.; Terrell, C.P.; Engleman, E.G.

    1984-01-01

    Patients with intractable rheumatoid arthritis (RA) were treated with total lymphoid irradiation (TLI, 200 rad). The authors previously reported long-lasting clinical improvement in this group associated with a persistent decrease in circulating Leu-3 (helper subset) T cells and marked impairment of in vitro lymphocyte function. In the present experiments, they studied the mechanisms underlying the decrease in pokeweed mitogen stimulated immunoglobulin (Ig) secretion observed after TLI. Peripheral blood mononuclear cells (PBL) from TLI-treated patients produced 10-fold less Ig (both IgM and IgG) in response to pokeweed mitogen than before radiotherapy. This decrease in Ig production was associated with the presence of suppressor cells in co-culture studies. By using responder cells obtained from normal individuals (allogeneic system), PBL from eight of 12 patients after TLI suppressed Ig synthesis by more than 50%. In contrast, PBL from the same patients before TLI failed to suppress Ig synthesis. PBL with suppressive activity contained suppressor T cells, and the latter cells bore the Leu-2 surface antigen. In 50% of the patients studied suppressor cells were also found in the non-T fraction and were adherent to plastic. Interestingly, the Leu-2 + cells from TLI-treated patients were no more potent on a cell per cell basis than purified Leu-2 + cells obtained before TLI. Additional experiments suggested that the suppression mediated by T cells after TLI is related to the increased ratio of Leu-2 to Leu-3 cells observed after radiotherapy

  20. RERTR-7 Irradiation Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    D. M. Perez; M. A. Lillo; G. S. Chang; G. A. Roth; N. E. Woolstenhulme; D. M. Wachs

    2011-12-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-7A, was designed to test several modified fuel designs to target fission densities representative of a peak low enriched uranium (LEU) burnup in excess of 90% U-235 at peak experiment power sufficient to generate a peak surface heat flux of approximately 300 W/cm2. The RERTR-7B experiment was designed as a high power test of 'second generation' dispersion fuels at peak experiment power sufficient to generate a surface heat flux on the order of 230 W/cm2.1 The following report summarizes the life of the RERTR-7A and RERTR-7B experiments through end of irradiation, including as-run neutronic analyses, thermal analyses and hydraulic testing results.

  1. Status and progress of the RERTR program in the year 2004

    International Nuclear Information System (INIS)

    Travelli, A.

    2005-01-01

    The overall status of the RERTR program at the time of the last RERTR meeting is reviewed and the progress achieved since that meeting is described. In the fuel area, unexpected failures of LEU U-Mo dispersion plates and tubes under irradiation testing have prompted a revision of the plans to qualify these fuels. While potential solutions to the difficulties with U-Mo dispersion fuels are being explored in collaboration with our international partners, greater emphasis has been placed on accelerating development of monolithic LEU U-Mo fuel. The feasibility of converting several Russian-designed research reactors to LEU fuels has been addressed, and progress has been made in the development of LEU based 99 Mo production processes. The Russian RERTR program has made significant advances. A very important event of 2004 was the US DOE establishment of the Global Threat Reduction Initiative (GTRI). This new program accelerates and combines under the same US DOE management several programs, including RERTR, which aim to secure, remove, or dispose of, nuclear and other radioactive materials throughout the world that are vulnerable to theft by terrorists. (author)

  2. Microstructure of RERTR DU-alloys irradiated with krypton ions up to 100 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Gan, J. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Keiser, D.D., E-mail: Dennis.Keiser@inl.gov [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Miller, B.D.; Wachs, D.M. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Allen, T.R. [Department of Engineering Physics, University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Kirk, M.; Rest, J. [Materials Science Division, Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States)

    2011-04-15

    The radiation stability of the interaction product formed at the fuel-matrix interface of research reactor dispersion fuels, under fission-product bombardment, has a strong impact on fuel performance. Three depleted uranium alloys were cast that consisted of the following five phases to be investigated: U(Si, Al){sub 3}, (U, Mo)(Si, Al){sub 3}, UMo{sub 2}Al{sub 20}, U{sub 6}Mo{sub 4}Al{sub 43}, and UAl{sub 4}. Irradiation of transmission electron microscopy (TEM) disc samples with 500-keV Kr ions at 200 deg. C to doses up to {approx}100 displacements per atom (dpa) were conducted using a 300-keV electron microscope equipped with an ion accelerator. TEM results show that the U(Si, Al){sub 3} and UAl{sub 4} phases remain crystalline at 100 dpa without forming voids. The (U, Mo)(Si, Al){sub 3} and UMo{sub 2}Al{sub 20} phases become amorphous at 1 and {approx}2 dpa, respectively, and show no evidence of voids at 100 dpa. The U{sub 6}Mo{sub 4}Al{sub 43} phase goes to amorphous at less than 1 dpa and reveals high density voids at 100 dpa.

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

    Science.gov (United States)

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

    2017-08-01

    Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

  4. Microstructure of RERTR DU-alloys irradiated with krypton ions up to 100 dpa

    Science.gov (United States)

    Gan, J.; Keiser, D. D., Jr.; Miller, B. D.; Wachs, D. M.; Allen, T. R.; Kirk, M.; Rest, J.

    2011-04-01

    The radiation stability of the interaction product formed at the fuel-matrix interface of research reactor dispersion fuels, under fission-product bombardment, has a strong impact on fuel performance. Three depleted uranium alloys were cast that consisted of the following five phases to be investigated: U(Si, Al) 3, (U, Mo)(Si, Al) 3, UMo 2Al 20, U 6Mo 4Al 43, and UAl 4. Irradiation of transmission electron microscopy (TEM) disc samples with 500-keV Kr ions at 200 °C to doses up to ˜100 displacements per atom (dpa) were conducted using a 300-keV electron microscope equipped with an ion accelerator. TEM results show that the U(Si, Al) 3 and UAl 4 phases remain crystalline at 100 dpa without forming voids. The (U, Mo)(Si, Al) 3 and UMo 2Al 20 phases become amorphous at 1 and ˜2 dpa, respectively, and show no evidence of voids at 100 dpa. The U 6Mo 4Al 43 phase goes to amorphous at less than 1 dpa and reveals high density voids at 100 dpa.

  5. 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)

  6. 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

  7. Role of Annular Lipids in the Functional Properties of Leucine Transporter LeuT Proteomicelles.

    Science.gov (United States)

    LeVine, Michael V; Khelashvili, George; Shi, Lei; Quick, Matthias; Javitch, Jonathan A; Weinstein, Harel

    2016-02-16

    Recent work has shown that the choice of the type and concentration of detergent used for the solubilization of membrane proteins can strongly influence the results of functional experiments. In particular, the amino acid transporter LeuT can bind two substrate molecules in low concentrations of n-dodecyl β-d-maltopyranoside (DDM), whereas high concentrations reduce the molar binding stoichiometry to 1:1. Subsequent molecular dynamics (MD) simulations of LeuT in DDM proteomicelles revealed that DDM can penetrate to the extracellular vestibule and make stable contacts in the functionally important secondary substrate binding site (S2), suggesting a potential competitive mechanism for the reduction in binding stoichiometry. Because annular lipids can be retained during solubilization, we performed MD simulations of LeuT proteomicelles at various stages of the solubilization process. We find that at low DDM concentrations, lipids are retained around the protein and penetration of detergent into the S2 site does not occur, whereas at high concentrations, lipids are displaced and the probability of DDM binding in the S2 site is increased. This behavior is dependent on the type of detergent, however, as we find in the simulations that the detergent lauryl maltose-neopentyl glycol, which is approximately twice the size of DDM and structurally more closely resembles lipids, does not penetrate the protein even at very high concentrations. We present functional studies that confirm the computational findings, emphasizing the need for careful consideration of experimental conditions, and for cautious interpretation of data in gathering mechanistic information about membrane proteins.

  8. Microcystis aeruginos strain [D-Leu1] Mcyst-LR producer, from Buenos Aires province, Argentina

    Directory of Open Access Journals (Sweden)

    Lorena Rosso

    2014-04-01

    Full Text Available Objective: To show the toxicological and phylogenetic characterization of a native Microcystis aeruginosa (M. aeruginosa strain (named CAAT 2005-3 isolated from a water body of Buenos Aires province, Argentine. Methods: A M. aeruginosa strain was isolated from the drainage canal of the sewage treatment in the town of Pila, Buenos Aires province, Argentina and acclimated to laboratory conditions. The amplification of cpcBA-IGS Phcocyanin (PC, intergenic spacer and flanking regions was carried out in order to build a phylogenetic tree. An exactive/orbitrap mass spectrometer equipped with an electrospray ionization source (Thermo Fisher Scientific, Bremen, Germany was used for the LC/ESI-HRMS microcystins analysis. The number of cell/mL and [D-Leu1] Mcyst-LR production obtained as a function of time was modelled using the Gompertz equation. Results: The phylogenetic analysis showed that the sequence clustered with others M. aeruginosa sequences obtained from NCBI. The first Argentinian strain of M. aeruginosa (CAAT 2005-3 growing under culture conditions maintains the typical colonial architecture of M. aeruginosa with profuse mucilage. M. aeruginosa CAAT 2005-3 expresses a toxin variant, that was identified by LC-HRMS/Orbitrapas as [D-Leu1] microcystin-LR ([M+H]+=1 037.8 m/z. Conclusions: [D-Leu1] microcystin-LR has been also detected in M. aeruginosa samples from Canada, Brazil and Argentina. This work provides the basis for technological development and production of analytical standards of toxins present in our region.

  9. Documentation Experiences for Jamaican SLOWPOKE-2 Conversion from HEU to LEU

    International Nuclear Information System (INIS)

    Warner, T.-A.; Dennis, H.; Antoine, J.

    2015-01-01

    The Jamaican SLOWPOKE–2 (JM–1) is a 20 kW research reactor manufactured by Atomic Energy of Canada Limited and has been operating since March 1984, in the department of the International Centre for Environmental and Nuclear Sciences (ICENS), at the University of the West Indies, Mona Campus in Kingston, Jamaica. The pool type reactor has been primarily used for Neutron Activation Analysis in environmental, agricultural, geochemical, health-related studies and mineral exploration. The University, assisted by the IAEA under the GTRI/RERTR program, is currently in the process of converting from HEU to LEU. Extensive documentation on policies, general requirements, elements of the conversion quality assurance (QA) system and conversion QA administrative procedures is required for the conversion. The core conversion activities are being carried out in accordance with current international standards and regulatory guidelines of the newly established Jamaican Radiation Safety Authority (RSA) with agreement between the RSA and IAEA or DOE related to Nuclear Safety and Control. The documentation structure has taken into consideration nuclear safety and licensing, LEU fuel design and conversion analysis, LEU fuel procurement and fabrication, removal of HEU fuel and reactor maintenance and conversion and commissioning, with the conversion QA manual at the apex of the structure. To a large extent, the documentation format will adhere to that of the IAEA applicable regulatory standards and guidance documents. The major challenge of the conversion activities, it is envisioned, will come from the absence of any previous regulatory framework in Jamaica; however, a timeline for the process, which includes training and equipping of regulators, will guide operation. (author)

  10. Food irradiation

    International Nuclear Information System (INIS)

    Soothill, R.

    1987-01-01

    The issue of food irradiation has become important in Australia and overseas. This article discusses the results of the Australian Consumers' Association's (ACA) Inquiry into food irradiation, commissioned by the Federal Government. Issues discussed include: what is food irradiation; why irradiate food; how much food is consumer rights; and national regulations

  11. Performance of Estimation of distribution algorithm for initial core loading optimization of AHWR-LEU

    International Nuclear Information System (INIS)

    Thakur, Amit; Singh, Baltej; Gupta, Anurag; Duggal, Vibhuti; Bhatt, Kislay; Krishnani, P.D.

    2016-01-01

    Highlights: • EDA has been applied to optimize initial core of AHWR-LEU. • Suitable value of weighing factor ‘α’ and population size in EDA was estimated. • The effect of varying initial distribution function on optimized solution was studied. • For comparison, Genetic algorithm was also applied. - Abstract: Population based evolutionary algorithms now form an integral part of fuel management in nuclear reactors and are frequently being used for fuel loading pattern optimization (LPO) problems. In this paper we have applied Estimation of distribution algorithm (EDA) to optimize initial core loading pattern (LP) of AHWR-LEU. In EDA, new solutions are generated by sampling the probability distribution model estimated from the selected best candidate solutions. The weighing factor ‘α’ decides the fraction of current best solution for updating the probability distribution function after each generation. A wider use of EDA warrants a comprehensive study on parameters like population size, weighing factor ‘α’ and initial probability distribution function. In the present study, we have done an extensive analysis on these parameters (population size, weighing factor ‘α’ and initial probability distribution function) in EDA. It is observed that choosing a very small value of ‘α’ may limit the search of optimized solutions in the near vicinity of initial probability distribution function and better loading patterns which are away from initial distribution function may not be considered with due weightage. It is also observed that increasing the population size improves the optimized loading pattern, however the algorithm still fails if the initial distribution function is not close to the expected optimized solution. We have tried to find out the suitable values for ‘α’ and population size to be considered for AHWR-LEU initial core loading pattern optimization problem. For sake of comparison and completeness, we have also addressed the

  12. Present status of the use of LEU in aqueous reactors to produce Mo-99

    International Nuclear Information System (INIS)

    Ball, Russell M.; Pavshook, V.A.; Khvostionov, V.Ye.

    1998-01-01

    An operating aqueous homogeneous reactor, the ARGUS at Kurchatov Institute, has been used to produce fission product molybdenum-99 (Mo-99), widely used in nuclear medicine to produce technetium-99m (Tc-99m). The Mo-99 has been extracted from the sulfate solution using an organic sorbent after operation at 1 kW/liter. after purification, the material has been assayed and the result is well within required specification of the USPharmacopaeia. Operation calculation are presented to show the sources and quantity of alpha activity when LEU is used. (author)

  13. Experience in producing LEU fuel elements for the RSG-GAS

    International Nuclear Information System (INIS)

    Suripto, A.; Soentono, S.

    1991-01-01

    To achieve a self-reliance in the operation of the 30 MW Multipurpose Research Reactor at Serpong (the RSG-GAS), a fuel element production facility has been constructed nearby. The main task of the facility is to produce MTR type fuel and control elements containing U 3 O 8 -Al dispersion LEU fuel for the RSG-GAS. The hot commissioning activity has started in early 1988 after completion of the cold commissioning using depleted uranium in 1987, marking the beginning of the real production activity. This paper briefly describes the main features of the fuel production facility, the production experience gained so far, and its current production activity. (orig.)

  14. 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

  15. Food irradiation

    International Nuclear Information System (INIS)

    Lindqvist, H.

    1996-01-01

    This paper is a review of food irradiation and lists plants for food irradiation in the world. Possible applications for irradiation are discussed, and changes induced in food from radiation, nutritional as well as organoleptic, are reviewed. Possible toxicological risks with irradiated food and risks from alternative methods for treatment are also brought up. Ways to analyze weather food has been irradiated or not are presented. 8 refs

  16. Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania's 14-MW TRIGA reactor

    International Nuclear Information System (INIS)

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

    1992-01-01

    The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, each containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations

  17. Preproghrelin Leu72Met polymorphism predicts a lower rate of developing renal dysfunction in type 2 diabetic nephropathy.

    Science.gov (United States)

    Lee, Dae-Yeol; Kim, Sun-Young; Jo, Dae-Sun; Hwang, Pyoung Han; Kang, Kyung Pyo; Lee, Sik; Kim, Won; Park, Sung Kwang

    2006-07-01

    Ghrelin is a novel peptide hormone, which exerts somatotropic, orexigenic and adipogenic effects. Recent studies have shown that the preproghrelin Leu72Met polymorphism is associated with serum creatinine (Scr) concentration in type 2 diabetes; 72Met carriers exhibited lower Scr levels as compared with the 72Met non-carriers. We hypothesized that the preproghrelin Leu72Met polymorphism is associated with a lower rate of developing renal dysfunction in patients with type 2 diabetic nephropathy. The preproghrelin Leu72Met polymorphism was investigated using PCR techniques in 138 patients with diabetic nephropathy divided into two groups, one with normal renal function and the other with renal dysfunction. Determination of the frequency of the preproghrelin Leu72Met polymorphism was the main outcome measure. The frequency of the Leu72Met polymorphism in diabetic nephropathy was significantly lower in patients with renal dysfunction (15.9%, P polymorphism was also associated with serum total cholesterol levels in diabetic nephropathy patients with renal dysfunction; the 72Met carriers had lower total cholesterol levels than the 72Met non-carriers (P < 0.05). These data suggest that 72Met carrier status may be used as a marker predicting a lower chance of developing renal dysfunction in diabetic nephropathy.

  18. Development of a chromosomally integrated metabolite-inducible Leu3p-alpha-IPM "off-on" gene switch.

    Directory of Open Access Journals (Sweden)

    Maria Poulou

    2010-08-01

    Full Text Available Present technology uses mostly chimeric proteins as regulators and hormones or antibiotics as signals to induce spatial and temporal gene expression.Here, we show that a chromosomally integrated yeast 'Leu3p-alpha-IotaRhoMu' system constitutes a ligand-inducible regulatory "off-on" genetic switch with an extensively dynamic action area. We find that Leu3p acts as an active transcriptional repressor in the absence and as an activator in the presence of alpha-isopropylmalate (alpha-IotaRhoMu in primary fibroblasts isolated from double transgenic mouse embryos bearing ubiquitously expressing Leu3p and a Leu3p regulated GFP reporter. In the absence of the branched amino acid biosynthetic pathway in animals, metabolically stable alpha-IPM presents an EC(50 equal to 0.8837 mM and fast "OFF-ON" kinetics (t(50ON = 43 min, t(50OFF = 2.18 h, it enters the cells via passive diffusion, while it is non-toxic to mammalian cells and to fertilized mouse eggs cultured ex vivo.Our results demonstrate that the 'Leu3p-alpha-IotaRhoMu' constitutes a simpler and safer system for inducible gene expression in biomedical applications.

  19. Comparison of thermohydraulic and nuclear aspects in a standard HEU core and a typical LEU core for the HFR Petten. A case study

    International Nuclear Information System (INIS)

    Pruimboom, H.; Tas, A.

    1985-01-01

    Within the framework of the RERTR program various HEU-LEU core calculations have been performed by ANL in a cooperative effort with ECN and JRC Petten. The main purpose of this work has been to gain competence in analysing HEU-LEU core conversion for high power Materials Testing Reactors and to assist in a possible HEU-LEU conversion of the HFR Petten. For reference purposes the present HFR standard core (HEU) in the 'old' vessel geometry was calculated at first. As a next step the new vessel geometry and the increased fuel weights were taken into account. Subsequently various LEU HFR core options have been analysed. Main parameters in the LEU study were the uranium loading in the meat, the fuel type, the thickness of the meat, the number of fuel plates per element and the type of burnable poison applied. Though the study has not yet been completed, one of its striking preliminary results concerns the increased power peaking in the LEU fuel elements as compared with the HEU situation. A preliminary analysis of the thermal characteristics of a typical LEU core as compared with a standard HEU core has been made and is presented in the paper. A short survey of the various HEU and LEU calculations is given. The thermal safety analysis procedure for the HFR, as based on the flow instability criterion, is clarified. Finally, the thermal comparison HEU versus LEU and the resulting conclusions are presented. (author)

  20. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, New York (United States)

    2007-07-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat.

  1. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan; Kim, Yeon Soo

    2007-01-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

  2. Agonist properties of a stable hexapeptide analog of neurotensin, N alpha MeArg-Lys-Pro-Trp-tLeu-Leu (NT1).

    Science.gov (United States)

    Akunne, H C; Demattos, S B; Whetzel, S Z; Wustrow, D J; Davis, D M; Wise, L D; Cody, W L; Pugsley, T A; Heffner, T G

    1995-04-18

    The major signal transduction pathway for neurotensin (NT) receptors is the G-protein-dependent stimulation of phospholipase C, leading to the mobilization of intracellular free Ca2+ ([Ca2+]i) and the stimulation of cyclic GMP. We investigated the functional actions of an analog of NT(8-13), N alpha MeArg-Lys-Pro-Trp-tLeu-Leu (NT1), and other NT related analogs by quantitative measurement of the cytosolic free Ca2+ concentration in HT-29 (human colonic adenocarcinoma) cells using the Ca(2+)-sensitive dye fura-2/AM and by effects on cyclic GMP levels in rat cerebellar slices. The NT receptor binding affinities for these analogs to HT-29 cell membranes and newborn (10-day-old) mouse brain membranes were also investigated. Data obtained from HT-29 cell and mouse brain membrane preparations showed saturable single high-affinity sites and binding densities (Bmax) of 130.2 and 87.5 fmol/mg protein, respectively. The respective KD values were 0.47 and 0.39 nM, and the Hill coefficients were 0.99 and 0.92. The low-affinity levocabastine-sensitive site was not present (K1 > 10,000) in either membrane preparation. Although the correlation of binding between HT-29 cell membranes and mouse brain membranes was quite significant (r = 0.92), some of the reference agents had lower binding affinities in the HT-29 cell membranes. The metabolically stable compound NT1 plus other NT analogs and related peptides [NT, NT(8-13), xenopsin, neuromedin N, NT(9-13), kinetensin and (D-Trp11)-NT] increased intracellular Ca2+ levels in HT-29 cells, indicating NT receptor agonist properties. The effect of NT1 in mobilizing [Ca2+]i blocked by SR 48692, a non-peptide NT antagonist. Receptor binding affinities of NT analogs to HT-29 cell membranes were positively correlated with potencies for mobilizing intracellular calcium in the same cells. In addition, NT1 increased cyclic GMP levels in rat cerebellar slices, confirming the latter findings of its NT agonist action. These results substantiate

  3. The tetrapeptide Arg-Leu-Tyr-Glu inhibits VEGF-induced angiogenesis

    International Nuclear Information System (INIS)

    Baek, Yi-Yong; Lee, Dong-Keon; So, Ju-Hoon; Kim, Cheol-Hee; Jeoung, Dooil; Lee, Hansoo; Choe, Jongseon; Won, Moo-Ho; Ha, Kwon-Soo; Kwon, Young-Guen; Kim, Young-Myeong

    2015-01-01

    Kringle 5, derived from plasminogen, is highly capable of inhibiting angiogenesis. Here, we have designed and synthesized 10 tetrapeptides, based on the amino acid properties of the core tetrapeptide Lys-Leu-Tyr-Asp (KLYD) originating from anti-angiogenic kringle 5 of human plasminogen. Of these, Arg-Leu-Tyr-Glu (RLYE) effectively inhibited vascular endothelial growth factor (VEGF)-induced endothelial cell proliferation, migration and tube formation, with an IC 50 of 0.06–0.08 nM, which was about ten-fold lower than that of the control peptide KLYD (0.79 nM), as well as suppressed developmental angiogenesis in a zebrafish model. Furthermore, this peptide effectively inhibited the cellular events that precede angiogenesis, such as ERK and eNOS phosphorylation and nitric oxide production, in endothelial cells stimulated with VEGF. Collectively, these data demonstrate that RLYE is a potent anti-angiogenic peptide that targets the VEGF signaling pathway. - Highlights: • The tetrapeptide RLYE inhibited VEGF-induced angiogenesis in vitro. • RLYE also suppressed neovascularization in a zebrafish model. • Its effect was correlated with inhibition of VEGF-induced ERK and eNOS activation. • RLYE may be used as a therapeutic drug for angiogenesis-related diseases

  4. Core instrumentation and pre-operational procedures for core conversion HEU to LEU

    International Nuclear Information System (INIS)

    1984-02-01

    This report is intended for the reactor operator, to be used as a manual or checklist for general guidance on pre-startup activities that need to be addressed in preparation for conversion to Low Enriched Fuel (LEU). All nuclear, thermodynamic and safety calculations should have been performed prior to this stage of the core conversion process. During these calculations and certainly before ordering the new LEU fuel elements the reactor operator needs to very carefully consider additional important factors concerning the new fuel: fuel reliability, reliability of fuel fabricator, reprocessing contract or fuel element storage and disposal, economics of the new fuel cycle. At this stage, too, a preoperational experimental programme has to be developed and presented to the regulatory authorities for approval. This experimental programme could lead to additional requirements on: in-core instrumentation, out-of-core instrumentation or additional experimental devices. Detailed instructions on specific tests and measurements are not provided in this report since much information on the subject is available in the open literature

  5. The progesterone receptor Val660→Leu polymorphism and breast cancer risk

    International Nuclear Information System (INIS)

    De Vivo, Immaculata; Hankinson, Susan E; Colditz, Graham A; Hunter, David J

    2004-01-01

    Recent evidence suggests a role for progesterone in breast cancer development and tumorigenesis. Progesterone exerts its effect on target cells by interacting with its receptor; thus, genetic variations, which might cause alterations in the biological function in the progesterone receptor (PGR), can potentially contribute to an individual's susceptibility to breast cancer. It has been reported that the PROGINS allele, which is in complete linkage disequilibrium with a missense substitution in exon 4 (G/T, valine→leucine, at codon 660), is associated with a decreased risk for breast cancer. Using a nested case-control study design within the Nurses' Health Study cohort, we genotyped 1252 cases and 1660 matched controls with the use of the Taqman assay. We did not observe any association of breast cancer risk with carrying the G/T (Val660→Leu) polymorphism (odds ratio 1.10, 95% confidence interval 0.93–1.30). In addition, we did not observe an interaction between this allele and menopausal status and family history of breast cancer as reported previously. Overall, our study does not support an association between the Val660→Leu PROGINS polymorphism and breast cancer risk

  6. Association between ghrelin gene (Leu72Met) polymorphism and ghrelin serum level with coronary artery diseases.

    Science.gov (United States)

    Hedayatizadeh-Omran, Akbar; Rafiei, Alireza; Khajavi, Rezvan; Alizadeh-Navaei, Reza; Mokhberi, Vahid; Moradzadeh, Kambiz

    2014-02-01

    Research shows that ghrelin gene polymorphism has some association with coronary artery diseases (CAD). Due to genetic differences among nations and the high prevalence of CAD, we conducted this study to examine the possible association between the polymorphism of ghrelin gene Leu72Met and CAD among an Iranian population. This case-control study was undertaken with patients who were referred to referral heart center, in 2011, with chest pain or a positive exercise test. Patients with risk factors for heart disease or who were surgery candidates, who underwent angiography and echocardiography, were also included. DNA extractions were performed using a modified salting out method, and the ghrelin region was amplified using polymerase chain reaction. The presence of the Leu72Met polymorphism and the serum levels of ghrelin were determined using the restriction fragment length polymorphism method and the enzyme-linked immunosorbent assay, respectively. The results indicated that in CAD patients, the incidence of heart failure was significantly different between the groups with genotypes CC or AA+CA (p=0.041). Mean serum level of ghrelin in the CAD group was significantly higher than that in the control group (pghrelin genotypes and serum levels of ghrelin in both the CAD and control groups (ppolymorphism, as well as an increase in serum levels of ghrelin associated with genotype distribution such that ghrelin levels have an inverse relationship with the frequency of the CC genotype.

  7. The tetrapeptide Arg-Leu-Tyr-Glu inhibits VEGF-induced angiogenesis

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Yi-Yong; Lee, Dong-Keon [Department of Molecular and Cellular Biochemistry, School of Medicine, Kangwon National University, Chuncheon, Gangwon-do, 200-702 (Korea, Republic of); So, Ju-Hoon; Kim, Cheol-Hee [Department of Biology, Chungnam National University, Daejeon, 305-764 (Korea, Republic of); Jeoung, Dooil [Department of Biochemistry, College of Natural Sciences, Kangwon National University, Chuncheon, Gangwon-do, 200-702 (Korea, Republic of); Lee, Hansoo [Department of Life Sciences, College of Natural Sciences, Kangwon National University, Chuncheon, Gangwon-do, 200-702 (Korea, Republic of); Choe, Jongseon [Department of Immunology, School of Medicine, Kangwon National University, Chuncheon, Gangwon-do, 200-702 (Korea, Republic of); Won, Moo-Ho [Department of Neurobiology, School of Medicine, Kangwon National University, Chuncheon, Gangwon-do, 200-702 (Korea, Republic of); Ha, Kwon-Soo [Department of Molecular and Cellular Biochemistry, School of Medicine, Kangwon National University, Chuncheon, Gangwon-do, 200-702 (Korea, Republic of); Kwon, Young-Guen [Department of Biochemistry, College of Life Science and Biotechnology, Yonsei University, Seoul, 120-752 (Korea, Republic of); Kim, Young-Myeong, E-mail: ymkim@kangwon.ac.kr [Department of Molecular and Cellular Biochemistry, School of Medicine, Kangwon National University, Chuncheon, Gangwon-do, 200-702 (Korea, Republic of)

    2015-08-07

    Kringle 5, derived from plasminogen, is highly capable of inhibiting angiogenesis. Here, we have designed and synthesized 10 tetrapeptides, based on the amino acid properties of the core tetrapeptide Lys-Leu-Tyr-Asp (KLYD) originating from anti-angiogenic kringle 5 of human plasminogen. Of these, Arg-Leu-Tyr-Glu (RLYE) effectively inhibited vascular endothelial growth factor (VEGF)-induced endothelial cell proliferation, migration and tube formation, with an IC{sub 50} of 0.06–0.08 nM, which was about ten-fold lower than that of the control peptide KLYD (0.79 nM), as well as suppressed developmental angiogenesis in a zebrafish model. Furthermore, this peptide effectively inhibited the cellular events that precede angiogenesis, such as ERK and eNOS phosphorylation and nitric oxide production, in endothelial cells stimulated with VEGF. Collectively, these data demonstrate that RLYE is a potent anti-angiogenic peptide that targets the VEGF signaling pathway. - Highlights: • The tetrapeptide RLYE inhibited VEGF-induced angiogenesis in vitro. • RLYE also suppressed neovascularization in a zebrafish model. • Its effect was correlated with inhibition of VEGF-induced ERK and eNOS activation. • RLYE may be used as a therapeutic drug for angiogenesis-related diseases.

  8. Antidepressant Specificity of Serotonin Transporter Suggested by Three LeuT-SSRI Structures

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Z.; Zhen, J; Karpowich, N; Law, C; Reith, M; Wang, D

    2009-01-01

    Sertraline and fluoxetine are selective serotonin re-uptake inhibitors (SSRIs) that are widely prescribed to treat depression. They exert their effects by inhibiting the presynaptic plasma membrane serotonin transporter (SERT). All SSRIs possess halogen atoms at specific positions, which are key determinants for the drugs' specificity for SERT. For the SERT protein, however, the structural basis of its specificity for SSRIs is poorly understood. Here we report the crystal structures of LeuT, a bacterial SERT homolog, in complex with sertraline, R-fluoxetine or S-fluoxetine. The SSRI halogens all bind to exactly the same pocket within LeuT. Mutation at this halogen-binding pocket (HBP) in SERT markedly reduces the transporter's affinity for SSRIs but not for tricyclic antidepressants. Conversely, when the only nonconserved HBP residue in both norepinephrine and dopamine transporters is mutated into that found in SERT, their affinities for all the three SSRIs increase uniformly. Thus, the specificity of SERT for SSRIs is dependent largely on interaction of the drug halogens with the protein's HBP.

  9. 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)

  10. Stimulation of Escherichia coli F-18Col- Type-1 fimbriae synthesis by leuX

    DEFF Research Database (Denmark)

    Newman, Joseph V.; Burghoff, Robert L.; Pallesen, Lars

    1994-01-01

    Escherichia coli F-18, a normal human fecal isolate, is an excellent colonizer of the streptomycin-treated mouse large intestine. E. coli F-18Col-, a derivative of E. coli F-18 which no longer makes the E. coli F-18 colicin, colonizes the large intestine as well as E. coli F-18 when fed to mice...... alone but is eliminated when fed together with E. coli F-18. Recently we randomly cloned E. coli F-18 DNA into E. coli F-18Col- and let the mouse intestine select the best colonizer. In this way, we isolated a 6.5-kb E. coli F-18 DNA sequence that simultaneously stimulated synthesis of type 1 fimbriae...... and enhanced E. coli F-18Col- colonizing ability. In the present investigation we show that the gene responsible for stimulation of type 1 fimbriae synthesis appears to be leuX, which encodes a tRNA specific for the rare leucine codon UUG. Moreover, it appears that expression of leuX may be regulated by two...

  11. 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)

  12. Transition from HEU to LEU fuel in Romania's 14-MW TRIGA reactor

    International Nuclear Information System (INIS)

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

    1995-01-01

    The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 x 5 square array of HEU U (10 wt% - ZrH - Er 2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incoloy. With a total inventory of 35 HEU fuel clusters, burnup, considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each of the original 29 fuel clusters had an average 235 U burnup in the range from 50 to 62%. Because of the U.S. policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% 235 U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations. (author)

  13. Food irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Gruenewald, T

    1985-01-01

    Food irradiation has become a matter of topical interest also in the Federal Republic of Germany following applications for exemptions concerning irradiation tests of spices. After risks to human health by irradiation doses up to a level sufficient for product pasteurization were excluded, irradiation now offers a method suitable primarily for the disinfestation of fruit and decontamination of frozen and dried food. Codex Alimentarius standards which refer also to supervision and dosimetry have been established; they should be adopted as national law. However, in the majority of cases where individual countries including EC member-countries so far permitted food irradiation, these standards were not yet used. Approved irradiation technique for industrial use is available. Several industrial food irradiation plants, partly working also on a contractual basis, are already in operation in various countries. Consumer response still is largely unknown; since irradiated food is labelled, consumption of irradiated food will be decided upon by consumers.

  14. SEM characterization of an irradiated monolithic U-10Mo fuel plate

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B.; Finlay, M.R.

    2010-01-01

    Results of scanning electron microscopy (SEM) characterization of irradiated U-7Mo dispersion fuel plates with differing amounts of matrix Si have been reported. However, to date, no results of SEM analysis of irradiated U-Mo monolithic fuel plates have been reported. This paper describes the first SEM characterization results for an irradiated monolithic U-10Mo fuel plate. Two samples from this fuel plate were characterized. One sample was produced from the low-flux side of the fuel plate, and another was produced at the high-flux side of the fuel plate. This characterization focused on the microstructural features present at the U-10Mo foil/AA6061 cladding interface, particularly the interaction zone that had developed during fabrication and any continued development during irradiation. In addition, the microstructure of the foil itself was investigated, along with the morphology of the observed fission gas bubbles. It was observed that a Si-rich interaction layer was present at the U-10Mo foil/cladding interface that exhibited relatively good irradiation behavior, and within the U-10Mo foil the microstructural features differed in some respects from what is typically seen in the U-7Mo powders of an irradiated dispersion fuel. (author)

  15. Key considerations in the conversion to LEU of a Mo-99 commercially producing reactor: SAFARI-1 of South Africa

    International Nuclear Information System (INIS)

    Stumpf, W.E.; Vermaak, A.P.; Ball, G.

    2000-01-01

    Apart from the technological demands and considerations associated with the conversion of a Mo-99 commercially producing reactor to LEU, a number of commercial challenges also need to be addressed. This is particularly the case when the reactor is primarily used as a source for the production, on an uninterrupted basis, of significant quantities of Mo-99 to satisfy long term commitments to a range of global customers. This paper highlights key business considerations which are applicable in the conversion process of firstly, reactor fuel to LEU and secondly target plates for Mo-99, also to LEU, using the SAFARI-1 reactor in South Africa as a typical example of such a commercially utilized reactor. (author)

  16. Key considerations in the conversion to LEU of a Mo-99 commercially producing reactor: SAFARI-1 of South Africa

    Energy Technology Data Exchange (ETDEWEB)

    Stumpf, W E; Vermaak, A P; Ball, G [NECSA, PO Box 582, Pretoria (South Africa)

    2000-10-01

    Apart from the technological demands and considerations associated with the conversion of a Mo-99 commercially producing reactor to LEU, a number of commercial challenges also need to be addressed. This is particularly the case when the reactor is primarily used as a source for the production, on an uninterrupted basis, of significant quantities of Mo-99 to satisfy long term commitments to a range of global customers. This paper highlights key business considerations which are applicable in the conversion process of firstly, reactor fuel to LEU and secondly target plates for Mo-99, also to LEU, using the SAFARI-1 reactor in South Africa as a typical example of such a commercially utilized reactor. (author)

  17. Fuel cycle flexibility in Advanced Heavy Water Reactor (AHWR) with the use of Th-LEU fuel

    International Nuclear Information System (INIS)

    Thakur, A.; Singh, B.; Pushpam, N.P.; Bharti, V.; Kannan, U.; Krishnani, P.D.; Sinha, R.K.

    2011-01-01

    The Advanced Heavy Water Reactor (AHWR) is being designed for large scale commercial utilization of thorium (Th) and integrated technological demonstration of the thorium cycle in India. The AHWR is a 920 MW(th), vertical pressure tube type cooled by boiling light water and moderated by heavy water. Heat removal through natural circulation and on-line fuelling are some of the salient features of AHWR design. The physics design of AHWR offers considerable flexibility to accommodate different kinds of fuel cycles. Our recent efforts have been directed towards a case study for the use of Th-LEU fuel cycle in a once-through mode. The discharged Uranium from Th-LEU cycle has proliferation resistant characteristics. This paper gives the initial core, fuel cycle characteristics and online refueling strategy of Th-LEU fuel in AHWR. (author)

  18. 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 information to analysts evaluating spent fuel subcriticality. The original purpose of the subcritical measurements was to validate computer model predictions that spent N Reactor fuel of a particular, typical exposure (2740 MWd/t) had a critical mass equal to twice that of unexposed fuel of the same type. The motivation for performing this work was driven by the need to increase spent fuel storage limits. These subcritical measurements confirmed the computer model predictions

  19. A Conserved Leucine Occupies the Empty Substrate Site of LeuT in the Na+-free Return State

    DEFF Research Database (Denmark)

    Malinauskaite, Lina; Said, Saida; Sahin, Caglanur

    2016-01-01

    Bacterial members of the neurotransmitter:sodium symporter (NSS) family perform Na+-dependent amino-acid uptake and extrude H+ in return. Previous NSS structures represent intermediates of Na+/substrate binding or intracellular release, but not the inward-to-outward return transition. Here we...... report crystal structures of Aquifex aeolicus LeuT in an outward-oriented, Na+- and substrate-free state likely to be H+-occluded. We find a remarkable rotation of the conserved Leu25 into the empty substrate-binding pocket and rearrangements of the empty Na+ sites. Mutational studies of the equivalent...

  20. Food irradiation

    International Nuclear Information System (INIS)

    Sato, Tomotaro; Aoki, Shohei

    1976-01-01

    Definition and significance of food irradiation were described. The details of its development and present state were also described. The effect of the irradiation on Irish potatoes, onions, wiener sausages, kamaboko (boiled fish-paste), and mandarin oranges was evaluated; and healthiness of food irradiation was discussed. Studies of the irradiation equipment for Irish potatoes in a large-sized container, and the silo-typed irradiation equipment for rice and wheat were mentioned. Shihoro RI center in Hokkaido which was put to practical use for the irradiation of Irish potatoes was introduced. The state of permission of food irradiation in foreign countries in 1975 was introduced. As a view of the food irradiation in the future, its utilization for the prevention of epidemics due to imported foods was mentioned. (Serizawa, K.)

  1. Gamma irradiator

    International Nuclear Information System (INIS)

    Simonet, G.

    1986-09-01

    Fiability of devices set around reactors depends on material resistance under irradiation noticeably joints, insulators, which belongs to composition of technical, safety or physical incasurement devices. The irradiated fuel elements, during their desactivation in a pool, are an interesting gamma irradiation device to simulate damages created in a nuclear environment. The existing facility at Osiris allows to generate an homogeneous rate dose in an important volume. The control of the element distances to irradiation box allows to control this dose rate [fr

  2. Food irradiation

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    The article explains what radiation does to food to preserve it. Food irradiation is of economic importance to Canada because Atomic Energy of Canada Limited is the leading world supplier of industrial irradiators. Progress is being made towards changing regulations which have restricted the irradiation of food in the United States and Canada. Examples are given of applications in other countries. Opposition to food irradiation by antinuclear groups is addressed

  3. TRIGA Research Reactor Conversion to LEU and Modernization of Safety Related Systems

    Energy Technology Data Exchange (ETDEWEB)

    Sanda, R. M. [Institute for Nuclear Research Piteşti (SCN-Piteşti), Piteşti (Romania)

    2014-08-15

    The USA and IAEA proposed an international programme to reduce the enrichment of uranium in research reactors by converting nuclear fuel containing HEU into fuel containing 20% enriched uranium. The Government of Romania joined the programme and actively supported political, scientific, technical and economic actions that led to the conversion of the active area of the 14 MW TRIGA reactor at the Institute for Nuclear Research in Piteşti in May 2006. This confirmed the continuity of the Romanian Government’s non-proliferation policy and their active support of international cooperation. Conversion of the Piteşti research reactor was made possible by completion of milestones in the Research Agreement for Reactor Conversion, a contract signed with the US Department of Energy and Argonne National Laboratory. This agreement provided scientific and technical support and the possibility of delivery of all HEU TRIGA fuel to the United States. Additionally, about 65% of the fresh LEU fuel needed to start the conversion was delivered in the period 1992–1994. Furthermore, conversion was promoted through IAEA Technical Cooperation project ROM/4/024 project funded primarily by the United States that supported technical and scientific efforts and the delivery of the remaining required LEU nuclear fuel to complete the conversion. Nuclear fuel to complete the conversion was made by the French company CERCA with a tripartite contract among the IAEA, CERCA and Romania. The contract was funded by the US Department of Energy with a voluntary contribution by the Romanian Government. The contract stipulated manufacturing and delivery of LEU fuel by CERCA with compliance measures for quality, delivery schedule and safety requirements set by IAEA standards and Romanian legislation. The project was supported by the ongoing technical cooperation, safeguards, legal and procurement assistance of the IAEA, in particular its Department of Nuclear Safety. For Romanian research, the

  4. Food irradiation

    International Nuclear Information System (INIS)

    Beyers, M.

    1977-01-01

    The objectives of food irradiation are outlined. The interaction of irradiation with matter is then discussed with special reference to the major constituents of foods. The application of chemical analysis in the evaluation of the wholesomeness of irradiated foods is summarized [af

  5. Integrin beta3 Leu33Pro polymorphism and risk of hip fracture: 25 years follow-up of 9233 adults from the general population

    DEFF Research Database (Denmark)

    Tofteng, Charlotte L; Bach-Mortensen, Pernille; Bojesen, Stig E

    2007-01-01

    for the integrin beta3 Leu33Pro polymorphism have a two-fold risk of hip fracture, mainly confined to postmenopausal women. Integrin beta3 Leu33Pro homozygosity could prove a useful marker for risk of future hip fracture and may contribute to pharmacogenetic variation in effects of integrin alphavbeta3 antagonists....

  6. IRT‑Sofia, HEU to LEU conversion: regulatory approval tasks solution overview

    International Nuclear Information System (INIS)

    Mitev, Mladen; Belousov, Sergey; Dimitrov, Dobromir

    2014-01-01

    The HEU to LEU conversion of the IRT–Sofia research reactor of the Institute for Nuclear Research and Nuclear Energy of the Bulgarian Academy of Sciences was jointly studied with the Argonne National Laboratory as a part of the RERTR Programme. The main purpose of the collaboration consisted in accomplishment of safety analyses and preparation of documents used for regulatory approval tasks solution. The main steps and results which are fundamental for the preparation of IRT–Sofia Safety Analyses Report including Operating Limits and Conditions are presented in this paper. The documents prepared by INRNE in accordance with the European nuclear safety requirements and IAEA recommendations were submitted for approval to the Bulgarian Nuclear Regulatory Agency at the end of 2010. Key words: research reactor, safety analyses report, Nuclear Regulatory Agency

  7. Analysis of the loss of coolant accident for LEU cores of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Khan, L.A.; Bokhari, I.H.; Raza, S.H.

    1993-12-01

    Response of LEU cores for PARR-1 to a Loss of Coolant Accident (LOCA) has been studied. It has been assumed that pool water drains out to double ended rupture of primary coolant pipe or complete shearing of an experimental beam tube. Results show that for an operating power level of 10 MW, both the first high power and equilibrium cores would enter into melting conditions if the pool drain time is less than 22 h and 11 h respectively. However, an Emergency Core Cooling System (ECCS) capable of spraying the core at flow rate of 8.3 m/sup 3/h, for the above mentioned duration, would keep the peak core temperature much below the critical value. Maximum operating power levels below which melting would not occur have been assessed to 3.4 MW and 4.8 MW, respectively, for the first high power and equilibrium cores. (author) 5 figs

  8. Analysis of the critical and first full power operating cores for PARR using leu oxide fuel

    International Nuclear Information System (INIS)

    Khan, L.A.; Qazi, M.K.; Bokhari, I.H.; Fazal, R.

    1989-10-01

    This paper explains the analysis for determining the first full power operating core for PARR using LEU oxide fuel. The core configuration selected for this first full power operation contains about 6.13 kg of U-235 distributed in 19 standard and five control fuel elements. The neutron flux level is doubled when core is shifted from 5MW to 10 MW. Total nuclear power peaking factor of the core is 2.03. The analysis shows that the core can be operated safely at 5 MW with a flow rate of 520 meter cube per hour and at 10 MW with a flow rate of 900 meter cube per hour. (A.B.). 10 figs

  9. RHF RELAP5 model and preliminary loss-of-offsite-power simulation results for LEU conversion

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Bergeron, A. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Dionne, B. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Thomas, F. [Institut Laue-Langevin (ILL), Grenoble (Switzerland). RHF Reactor Dept.

    2014-08-01

    The purpose of this document is to describe the current state of the RELAP5 model for the Institut Laue-Langevin High Flux Reactor (RHF) located in Grenoble, France, and provide an update to the key information required to complete, for example, simulations for a loss of offsite power (LOOP) accident. A previous status report identified a list of 22 items to be resolved in order to complete the RELAP5 model. Most of these items have been resolved by ANL and the RHF team. Enough information was available to perform preliminary safety analyses and define the key items that are still required. Section 2 of this document describes the RELAP5 model of RHF. The final part of this section briefly summarizes previous model issues and resolutions. Section 3 of this document describes preliminary LOOP simulations for both HEU and LEU fuel at beginning of cycle conditions.

  10. Core physics analysis in support of the FNR HEU-LEU demonstration experiment

    International Nuclear Information System (INIS)

    Losey, David C.; Brown, Forrest B.; Martin, William R.; Lee, John C.

    1983-01-01

    A core neutronics analysis has been undertaken to assess the impact of low-enrichment fuel on the performance and utilization of the FNR As part of this analytic effort a computer code system has been assembled which will be of general use in analyzing research reactors with MTR-type fuel. The code system has been extensively tested and verified in calculations for the present high enrichment core. The analysis presented here compares the high-and-low enrichment fuels in batch and equilibrium core configurations which model the actual FNR operating conditions. The two fuels are compared for cycle length, fuel burnup, and flux and power distributions, as well as for the reactivity effects which are important in assessing the impact of LEU fuel on reactor shutdown margin. (author)

  11. Core physics analysis in support of the FNR HEU-LEU demonstration experiment

    Energy Technology Data Exchange (ETDEWEB)

    Losey, David C; Brown, Forrest B; Martin, William R; Lee, John C [Department of Nuclear Engineering, University of Michigan (United States)

    1983-08-01

    A core neutronics analysis has been undertaken to assess the impact of low-enrichment fuel on the performance and utilization of the FNR As part of this analytic effort a computer code system has been assembled which will be of general use in analyzing research reactors with MTR-type fuel. The code system has been extensively tested and verified in calculations for the present high enrichment core. The analysis presented here compares the high-and-low enrichment fuels in batch and equilibrium core configurations which model the actual FNR operating conditions. The two fuels are compared for cycle length, fuel burnup, and flux and power distributions, as well as for the reactivity effects which are important in assessing the impact of LEU fuel on reactor shutdown margin. (author)

  12. A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor

    International Nuclear Information System (INIS)

    Bretscher, M. M.

    1998-01-01

    The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% 235 U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm 3 and 3.8 gU/cm 3 are required to match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively

  13. RERTR program activities related to the development and application of new LEU fuels

    International Nuclear Information System (INIS)

    Travelli, A.

    1983-01-01

    The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U 3 Si 2 -Al and U 3 Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm 3 each year, from the current 1.7 g U/cm 3 to the 7.0 g U/cm 3 which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years

  14. Food irradiation

    International Nuclear Information System (INIS)

    Macklin, M.

    1987-01-01

    The Queensland Government has given its support the establishment of a food irradiation plant in Queensland. The decision to press ahead with a food irradiation plant is astonishing given that there are two independent inquiries being carried out into food irradiation - a Parliamentary Committee inquiry and an inquiry by the Australian Consumers Association, both of which have still to table their Reports. It is fair to assume from the Queensland Government's response to date, therefore, that the Government will proceed with its food irradiation proposals regardless of the outcomes of the various federal inquiries. The reasons for the Australian Democrats' opposition to food irradiation which are also those of concerned citizens are outlined

  15. Food irradiation

    International Nuclear Information System (INIS)

    Duchacek, V.

    1989-01-01

    The ranges of doses used for food irradiation and their effect on the processed foods are outlined. The wholesomeness of irradiated foods is discussed. The present food irradiation technology development in the world is described. A review of the irradiated foods permitted for public consumption, the purposes of food irradiaton, the doses used and a review of the commercial-scale food irradiators are tabulated. The history and the present state of food processing in Czechoslovakia are described. (author). 1 fig., 3 tabs., 13 refs

  16. Irradiated foods

    International Nuclear Information System (INIS)

    Darrington, Hugh

    1988-06-01

    This special edition of 'Food Manufacture' presents papers on the following aspects of the use of irradiation in the food industry:- 1) an outline view of current technology and its potential. 2) Safety and wholesomeness of irradiated and non-irradiated foods. 3) A review of the known effects of irradiation on packaging. 4) The problems of regulating the use of irradiation and consumer protection against abuse. 5) The detection problem - current procedures. 6) Description of the Gammaster BV plant in Holland. 7) World outline review. 8) Current and future commercial activities in Europe. (U.K.)

  17. Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showing agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm2 and temporary heat flux limit of 600 W/cm2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.

  18. 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.

  19. Structures of LeuT in bicelles define conformation and substrate binding in a membrane-like context

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Hui; Elferich, Johannes; Gouaux, Eric (Oregon HSU)

    2012-02-13

    Neurotransmitter sodium symporters (NSSs) catalyze the uptake of neurotransmitters into cells, terminating neurotransmission at chemical synapses. Consistent with the role of NSSs in the central nervous system, they are implicated in multiple diseases and disorders. LeuT, from Aquifex aeolicus, is a prokaryotic ortholog of the NSS family and has contributed to our understanding of the structure, mechanism and pharmacology of NSSs. At present, however, the functional state of LeuT in crystals grown in the presence of n-octyl-{beta}-D-glucopyranoside ({beta}-OG) and the number of substrate binding sites are controversial issues. Here we present crystal structures of LeuT grown in DMPC-CHAPSO bicelles and demonstrate that the conformations of LeuT-substrate complexes in lipid bicelles and in {beta}-OG detergent micelles are nearly identical. Furthermore, using crystals grown in bicelles and the substrate leucine or the substrate analog selenomethionine, we find only a single substrate molecule in the primary binding site.

  20. Direct assessment of substrate binding to the Neurotransmitter:Sodium Symporter LeuT by solid state NMR

    DEFF Research Database (Denmark)

    Erlendsson, Simon; Gotfryd, Kamil; Larsen, Flemming Hofmann

    2017-01-01

    The Neurotransmitter:Sodium Symporters (NSSs) represent an important class of proteins mediating sodium-dependent uptake of neurotransmitters from the extracellular space. The substrate binding stoichiometry of the bacterial NSS protein, LeuT, and thus the principal transport mechanism, has been...

  1. The Leu72Met polymorphism of the GHRL gene prevents the development of diabetic nephropathy in Chinese patients with type 2 diabetes mellitus.

    Science.gov (United States)

    Zhuang, Langen; Li, Ming; Yu, Changhua; Li, Can; Zhao, Mingming; Lu, Ming; Zheng, Taishan; Zhang, Rong; Zhao, Weijing; Bao, Yuqian; Xiang, Kunsan; Jia, Weiping; Wang, Niansong; Liu, Limei

    2014-02-01

    The preproghrelin (GHRL) Leu72Met polymorphism (rs 696217) is associated with obesity, reduced glucose-induced insulin secretion in healthy or diabetic subjects, and reduced serum creatinine (Scr) levels in type 2 diabetes. We evaluated the association of the Leu72Met polymorphism with measures of insulin sensitivity in non-diabetic control individuals and type 2 diabetics, and whether this variation contributes to the development of diabetic nephropathy (DN) in type 2 diabetes. A case-control study was performed of 291 non-diabetic control subjects and 466 patients with type 2 diabetes, of whom 238 had DN with overt albuminuria (DN group; albuminuric excretion rate [AER] ≥ 300 mg/24 h) and 228 did not have DN, but had diabetes for more than 10 years (non-DN group). Genotyping was performed using a TaqMan PCR assay. The Leu/Leu, Leu/Met, and Met/Met genotype frequencies were significantly different between the non-DN and DN groups (p = 0.011). The frequency of the variant genotypes (Leu/Met, Met/Met) was significantly lower in the DN group than the non-DN group (23.5 vs. 36.0 %, p = 0.003). Met/Met non-diabetic control subjects had lower BMI and Scr levels and higher eGFR level than Leu/Leu or Leu/Met individuals (p GHRL Leu72Met polymorphism may help to maintain normal renal function and may protect against the development of DN by reducing albuminuria and improving renal function in Chinese patients with type 2 diabetes.

  2. 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...

  3. The Leu72Met polymorphism of the ghrelin gene is associated with a decreased risk for type 2 diabetes.

    Science.gov (United States)

    Berthold, Heiner K; Giannakidou, Eleni; Krone, Wilhelm; Mantzoros, Christos S; Gouni-Berthold, Ioanna

    2009-01-01

    Ghrelin is involved in several metabolic and cardiovascular processes. The Leu72Met polymorphism of its gene was associated with an increased risk of type 2 diabetes (DM2) in some, but not all studies. Its association with atherosclerosis is not known. We investigated 420 Caucasian subjects with DM2 and 430 controls without diabetes (56.6% male, age 62+/-10 years). The Leu72Leu genotype frequencies were 89.76/84.65%, the Leu72Met 9.52/15.12% and the Met72Met 0.71/0.23% (P=0.029) in the DM2 and controls groups, respectively. In subjects with Met72+ genotypes the risk of DM2 was significantly decreased (univariate OR 0.63, 95% CI 0.42-0.95, P=0.026). In a logistic regression model, body mass index, hypertension and a positive family history for diabetes were predictors of diabetes while the polymorphism remained negatively associated with the disease (OR 0.62, 95% CI 0.40-0.97, P=0.036). After adjusting for known risk factors for atherosclerosis, the Met72+ variant was not associated with atherosclerotic disease (OR 1.41, 95% CI 0.78-2.54, P=0.25). Ghrelin concentrations were not associated with the polymorphism, DM2 or atherosclerotic disease. The Leu72Met polymorphism of the ghrelin gene is associated with a decreased risk for DM2. There is no association between the variant and atherosclerotic disease or ghrelin concentrations.

  4. Kr ion irradiation study of the depleted-uranium alloys

    Science.gov (United States)

    Gan, J.; Keiser, D. D.; Miller, B. D.; Kirk, M. A.; Rest, J.; Allen, T. R.; Wachs, D. M.

    2010-12-01

    Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si) 3, (U, Mo)(Al, Si) 3, UMo 2Al 20, U 6Mo 4Al 43 and UAl 4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10 19 ions/m 2 (˜10 dpa) with an Kr ion flux of 10 16 ions/m 2/s (˜4.0 × 10 -3 dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

  5. Kr ion irradiation study of the depleted-uranium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Gan, J., E-mail: Jian.Gan@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Keiser, D.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Miller, B.D. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Kirk, M.A.; Rest, J. [Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 (United States); Allen, T.R. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Wachs, D.M. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

    2010-12-01

    Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si){sub 3}, (U, Mo)(Al, Si){sub 3}, UMo{sub 2}Al{sub 20}, U{sub 6}Mo{sub 4}Al{sub 43} and UAl{sub 4}. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 {sup o}C to ion doses up to 2.5 x 10{sup 19} ions/m{sup 2} ({approx}10 dpa) with an Kr ion flux of 10{sup 16} ions/m{sup 2}/s ({approx}4.0 x 10{sup -3} dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

  6. Progress in the development of very high density research and test reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    Wachs, D.M. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, Idaho 83415 (United States)

    2009-06-15

    New nuclear fuels are being developed to enable many of the most important research and test reactors worldwide to convert from high enriched uranium (HEU) fuels to low enriched uranium (LEU) fuels without significant loss in performance. The last decade of work has focused on the development of uranium-molybdenum alloy (U-Mo) based fuels and is an international effort that includes the active participation of more than ten national programs. The US RERTR program, under the NNSA's Global Threat Reduction Initiative (GTRI), is in the process of developing both dispersion and monolithic U-Mo fuel designs. While the U-Mo fuel alloy has behaved extremely well under irradiation, initial testing (circa 2003) revealed that the U-Mo fuels dispersed in aluminum had an unexpected tendency toward unstable swelling (pillowing) under high-power conditions. Technical investigations were initiated worldwide at this time by the partner programs to understand this behavior as well as to develop and test remedies. The behavior was corrected by modifying the chemistry of the U-Mo/Al interfaces in both fuel designs. In the dispersion fuel design, this was accomplished by the addition of small amounts of silicon to the aluminum matrix material. Two methods are under development for the monolithic fuel design, which include the application of a thin layer of silicon or a thin zirconium based diffusion barrier at the fuel/clad interface. This paper gives an overview of the current status of U-Mo fuel development, including basic research results, manufacturing aspects, results of the latest irradiations and post irradiation examinations, the approach to fuel performance qualification, and the scale-up and commercialization of fabrication technology. (authors)

  7. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N. R. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Brown, N. R. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Baek, J. S [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Cheng, L. Y. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.

    2014-04-30

    It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-Enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size-Plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

  8. Foodstuff irradiation

    International Nuclear Information System (INIS)

    1982-01-01

    Report written on behalf of the Danish Food Institute summarizes national and international rules and developments within food irradiation technology, chemical changes in irradiated foodstuffs, microbiological and health-related aspects of irradiation and finally technological prospects of this conservation form. Food irradiatin has not been hitherto applied in Denmark. Radiation sources and secondary radiation doses in processed food are characterized. Chemical changes due to irradiation are compared to those due to p.ex. food heating. Toxicological and microbiological tests and their results give no unequivocal answer to the problem whether a foodstuff has been irradiated. The most likely application fields in Denmark are for low radiation dosis inhibition of germination, riping delay and insecticide. Medium dosis (1-10 kGy) can reduce bacteria number while high dosis (10-50 kGy) will enable total elimination of microorganisms and viruses. Food irradiation can be acceptable as technological possibility with reservation, that further studies follow. (EG)

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

    International Nuclear Information System (INIS)

    Taboada, Horacio; Gonzalez, Alfredo G.

    2005-01-01

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

  10. Transformation of Saccharomyces cerevisiae with UV-irradiated single-stranded plasmid.

    Science.gov (United States)

    Zgaga, Z

    1991-08-01

    UV-irradiated single-stranded replicative plasmids were used to transform different yeast strains. The low doses of UV used in this study (10-75 J/m2) caused a significant decrease in the transforming efficiency of plasmid DNA in the Rad+ strain, while they had no effect on transformation with double-stranded plasmids of comparable size. Neither the rev3 mutation, nor the rad18 or rad52 mutations influenced the efficiency of transformation with irradiated single-stranded plasmid. However, it was found to be decreased in the double rev3 rad52 mutant. Extracellular irradiation of plasmid that contains both URA3 and LEU2 genes (psLU) gave rise to up to 5% Leu- transformants among selected Ura+ ones in the repair-proficient strain. Induction of Leu- transformants was dose-dependent and only partially depressed in the rev3 mutant. These results suggest that both mutagenic and recombinational repair processes operate on UV-damaged single-stranded DNA in yeast.

  11. LAPTM4b recruits the LAT1-4F2hc Leu transporter to lysosomes and promotes mTORC1 activation.

    Science.gov (United States)

    Milkereit, Ruth; Persaud, Avinash; Vanoaica, Liviu; Guetg, Adriano; Verrey, Francois; Rotin, Daniela

    2015-05-22

    Mammalian target of rapamycin 1 (mTORC1), a master regulator of cellular growth, is activated downstream of growth factors, energy signalling and intracellular essential amino acids (EAAs) such as Leu. mTORC1 activation occurs at the lysosomal membrane, and involves V-ATPase stimulation by intra-lysosomal EAA (inside-out activation), leading to activation of the Ragulator, RagA/B-GTP and mTORC1 via Rheb-GTP. How Leu enters the lysosomes is unknown. Here we identified the lysosomal protein LAPTM4b as a binding partner for the Leu transporter, LAT1-4F2hc (SLC7A5-SLAC3A2). We show that LAPTM4b recruits LAT1-4F2hc to lysosomes, leading to uptake of Leu into lysosomes, and is required for mTORC1 activation via V-ATPase following EAA or Leu stimulation. These results demonstrate a functional Leu transporter at the lysosome, and help explain the inside-out lysosomal activation of mTORC1 by Leu/EAA.

  12. Hemibody irradiation

    International Nuclear Information System (INIS)

    Schen, B.C.; Mella, O.; Dahl, O.

    1992-01-01

    In a large number of cancer patients, extensive skeletal metastases or myelomatosis induce vast suffering, such as intolerable pain and local complications of neoplastic bone destruction. Analgetic drugs frequently do not yield sufficient palliation. Irradiation of local fields often has to be repeated, because of tumour growth outside previously irradiated volumes. Wide field irradiation of the lower or upper half of the body causes significant relief of pain in most patients. Adequate pretreatment handling of patients, method of irradiation, and follow-up are of importance to reduce side effects, and are described as they are carried out at the Department of Oncology, Haukeland Hospital, Norway. 16 refs., 2 figs

  13. Irradiation behavior of uranium oxide - Aluminum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products and as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show that, with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 g U/cm 3 ) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼63% 235 U burnup). (author)

  14. Irradiation behavior of uranium oxide-aluminum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, G.L.; Rest, J.; Snelgrove, J.L.

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products, as well as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show, that with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 gm/cc) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼ 63% 235 U burnup)

  15. Associations of Leu72Met Polymorphism of Preproghrelin with Ratios of Plasma Lipids Are Diversified by a High-Carbohydrate Diet in Healthy Chinese Adolescents.

    Science.gov (United States)

    Su, Mi; Qiu, Li; Wang, Qian; Jiang, Zhen; Liu, Xiao Juan; Lin, Jia; Fang, Ding Zhi

    2015-01-01

    The association of preproghrelin Leu72Met polymorphism with plasma lipids profile was inconsistently reported and needs more studies to be confirmed. Our study was to investigate the changes of plasma lipids ratios after a high-carbohydrate (high-CHO) diet in healthy Chinese adolescents with different genotypes of this polymorphism. Fifty-three healthy university students were given a washout diet of 54.1% carbohydrate for 7 days, followed by a high-CHO diet of 70.1% carbohydrate for 6 days. The anthropometric and biological parameters were analyzed at baseline and before and after the high-CHO diet. When compared with those before the high-CHO diet, body mass index (BMI) decreased in the male and female Met72 allele carriers. Decreased low-/high-density lipoprotein cholesterol (LDL-C/HDL-C) was observed in all participants except the female subjects with the Leu72Leu genotype. TG/HDL-C and log (TG/HDL-C) were increased only in the female subjects with the Leu72Leu genotype. These results suggest that the Met72 allele of preproghrelin Leu72Met polymorphism may be associated with decreased BMI induced by the high-CHO diet in male and female adolescents, while the Leu72 allele with increased TG/HDL-C and log (TG/HDL-C) in the female adolescents only. Furthermore, the decreasing effect of the high-CHO diet on LDL/HDL-C may be eliminated in the female Leu72Leu homozygotes. © 2015 S. Karger AG, Basel.

  16. Treatment of intractable rheumatoid arthritis with total lymphoid irradiation

    International Nuclear Information System (INIS)

    Kotzin, B.L.; Strober, S.; Engleman, E.G.; Calin, A.; Hoppe, R.T.; Kansas, G.S.; Terrell, C.P.; Kaplan, H.S.

    1981-01-01

    Eleven patients with intractable rheumatoid arthritis were treated with total lymphoid irradiation (total dose, 2000 rad) in an uncontrolled feasibility study, as an alternative to long-term therapy with cytotoxic drugs such as cyclophosphamide and azathioprine. During a follow-up period of five to 18 months after total lymphoid irradiation, there was a profound and sustained suppression of the absolute lymphocyte count and in vitro lymphocyte function, as well as an increase in the ratio of Leu-2 (suppressor/cytotoxic) to Leu-3 (helper) T cells in the blood. Persistent circulating suppressor cells of the mixed leukocyte response and of pokeweed mitogen-induced immunoglobulin secretion developed in most patients. In nine of the 11 patients, these changes in immune status were associated with relief of joint tenderness and swelling and with improvement in function scores. Maximum improvement occurred approximately six months after irradiation and continued for the remainder of the observation period. Few severe or chronic side effects were associated with the radiotherapy

  17. A comparative and predictive study of the annual fuel cycle costs for HEU and LEU fuels in the High Flux Reactor, Petten, 1985-1993

    Energy Technology Data Exchange (ETDEWEB)

    Moss, R L; May, P

    1985-07-01

    The internationally agreed constraint on availability of supply of HEU fuels to Research and Test Reactors has necessitated that a cost analysis be carried out to determine the financial effect of converting the core of the HFR from HEU to LEU fuels. A computer program, written at Petten and based on information extracted from studies in Europe and the USA, identifies the major cost variables to be manufacturing, uranium, reprocessing and transport costs. Comparison between HEU and LEU cores have been carried out and includes the effects of inflation and exchange rate fluctuations. Conversion of the HFR core to LEU fuels is shown to be financially disadvantageous. (author)

  18. A comparative and predictive study of the annual fuel cycle costs for HEU and LEU fuels in the High Flux Reactor, Petten, 1985-1993

    International Nuclear Information System (INIS)

    Moss, R.L.; May, P.

    1985-01-01

    The internationally agreed constraint on availability of supply of HEU fuels to Research and Test Reactors has necessitated that a cost analysis be carried out to determine the financial effect of converting the core of the HFR from HEU to LEU fuels. A computer program, written at Petten and based on information extracted from studies in Europe and the USA, identifies the major cost variables to be manufacturing, uranium, reprocessing and transport costs. Comparison between HEU and LEU cores have been carried out and includes the effects of inflation and exchange rate fluctuations. Conversion of the HFR core to LEU fuels is shown to be financially disadvantageous. (author)

  19. Food irradiation

    International Nuclear Information System (INIS)

    Mercader, J.P.; Emily Leong

    1985-01-01

    The paper discusses the need for effective and efficient technologies in improving the food handling system. It defines the basic premises for the development of food handling. The application of food irradiation technology is briefly discussed. The paper points out key considerations for the adoption of food irradiation technology in the ASEAN region (author)

  20. Food irradiation

    International Nuclear Information System (INIS)

    Matsuyama, Akira

    1990-01-01

    This paper reviews researches, commentaries, and conference and public records of food irradiation, published mainly during the period 1987-1989, focusing on the current conditions of food irradiation that may pose not only scientific or technologic problems but also political issues or consumerism. Approximately 50 kinds of food, although not enough to fill economic benefit, are now permitted for food irradiation in the world. Consumerism is pointed out as the major factor that precludes the feasibility of food irradiation in the world. In the United States, irradiation is feasible only for spices. Food irradiation has already been feasible in France, Hollands, Belgium, and the Soviet Union; has under consideration in the Great Britain, and has been rejected in the West Germany. Although the feasibility of food irradiation is projected to increase gradually in the future, commercial success or failure depends on the final selection of consumers. In this respect, the role of education and public information are stressed. Meat radicidation and recent progress in the method for detecting irradiated food are referred to. (N.K.) 128 refs

  1. Irradiation proctitis

    International Nuclear Information System (INIS)

    Minami, Akira

    1977-01-01

    Literatures on late rectal injuries are discussed, referring to two patients with uterine cervical cancer in whom irradiation proctitis occurred after telecobalt irradiation following uterine extirpation. To one patients, a total of 5000 rads was irradiated, dividing into 250 rads at one time, and after 3 months, irradiation with a total of 2000 rads, dividing into 200 rads at one time, was further given. In another one patient, two parallel opposing portal irradiation with a total of 6000 rads was given. About a year after the irradiation, rectal injuries and cystitis, accompanying with hemorrhage, were found in both of the patients. Rectal amputation and proctotoreusis were performed. Cystitis was treated by cystic irradiation in the urological department. Pathohistological studies of the rectal specimen revealed atrophic mucosa, and dilatation of the blood vessels and edema in the colonic submucosa. Incidence of this disease, term when the disease occurs, irradiation dose, type of the disease, treatment and prevention are described on the basis of the literatures. (Kanao, N.)

  2. Irradiation proctitis

    Energy Technology Data Exchange (ETDEWEB)

    Minami, A [Osaka Kita Tsishin Hospital (Japan)

    1977-06-01

    Literatures on late rectal injuries are discussed, referring to two patients with uterine cervical cancer in whom irradiation proctitis occurred after telecobalt irradiation following uterine extirpation. To one patients, a total of 5000 rads was irradiated, dividing into 250 rads at one time, and after 3 months, irradiation with a total of 2000 rads, dividing into 200 rads at one time, was further given. In another one patient, two parallel opposing portal irradiation with a total of 6000 rads was given. About a year after the irradiation, rectal injuries and cystitis, accompanying with hemorrhage, were found in both of the patients. Rectal amputation and proctotoreusis were performed. Cystitis was treated by cystic irradiation in the urological department. Pathohistological studies of the rectal specimen revealed atrophic mucosa, and dilatation of the blood vessels and edema in the colonic submucosa. Incidence of this disease, term when the disease occurs, irradiation dose, type of the disease, treatment and prevention are described on the basis of the literatures.

  3. Food irradiation

    International Nuclear Information System (INIS)

    Kobayashi, Yasuhiko; Kikuchi, Masahiro

    2009-01-01

    Food irradiation can have a number of beneficial effects, including prevention of sprouting; control of insects, parasites, pathogenic and spoilage bacteria, moulds and yeasts; and sterilization, which enables commodities to be stored for long periods. It is most unlikely that all these potential applications will prove commercially acceptable; the extend to which such acceptance is eventually achieved will be determined by practical and economic considerations. A review of the available scientific literature indicates that food irradiation is a thoroughly tested food technology. Safety studies have so far shown no deleterious effects. Irradiation will help to ensure a safer and more plentiful food supply by extending shelf-life and by inactivating pests and pathogens. As long as requirement for good manufacturing practice are implemented, food irradiation is safe and effective. Possible risks of food irradiation are not basically different from those resulting from misuse of other processing methods, such as canning, freezing and pasteurization. (author)

  4. Irradiation damage

    Energy Technology Data Exchange (ETDEWEB)

    Howe, L.M

    2000-07-01

    There is considerable interest in irradiation effects in intermetallic compounds from both the applied and fundamental aspects. Initially, this interest was associated mainly with nuclear reactor programs but it now extends to the fields of ion-beam modification of metals, behaviour of amorphous materials, ion-beam processing of electronic materials, and ion-beam simulations of various kinds. The field of irradiation damage in intermetallic compounds is rapidly expanding, and no attempt will be made in this chapter to cover all of the various aspects. Instead, attention will be focused on some specific areas and, hopefully, through these, some insight will be given into the physical processes involved, the present state of our knowledge, and the challenge of obtaining more comprehensive understanding in the future. The specific areas that will be covered are: point defects in intermetallic compounds; irradiation-enhanced ordering and irradiation-induced disordering of ordered alloys; irradiation-induced amorphization.

  5. Irradiation damage

    International Nuclear Information System (INIS)

    Howe, L.M.

    2000-01-01

    There is considerable interest in irradiation effects in intermetallic compounds from both the applied and fundamental aspects. Initially, this interest was associated mainly with nuclear reactor programs but it now extends to the fields of ion-beam modification of metals, behaviour of amorphous materials, ion-beam processing of electronic materials, and ion-beam simulations of various kinds. The field of irradiation damage in intermetallic compounds is rapidly expanding, and no attempt will be made in this chapter to cover all of the various aspects. Instead, attention will be focused on some specific areas and, hopefully, through these, some insight will be given into the physical processes involved, the present state of our knowledge, and the challenge of obtaining more comprehensive understanding in the future. The specific areas that will be covered are: point defects in intermetallic compounds; irradiation-enhanced ordering and irradiation-induced disordering of ordered alloys; irradiation-induced amorphization

  6. High plasma ghrelin protects from coronary heart disease and Leu72Leu polymorphism of ghrelin gene from cancer in healthy adults during the 19 years follow-up study.

    Science.gov (United States)

    Laurila, M; Santaniemi, M; Kesäniemi, Y A; Ukkola, O

    2014-11-01

    The aim of our investigation was to find out if ghrelin concentrations or polymorphisms predict the future risk for cardiovascular diseases and cancer in a population-based cohort initiated in 1991 (491 hypertensive and 513 control subjects). Total mortality and hospital events were followed up for 19 years. Fasting total ghrelin concentrations were determined and Arg51Gln, Leu72Met and -501 A > C polymorphisms identified. Cox regression analysis was performed. The mean value in the control cohort was 674 pg/ml whereas in the hypertensive cohort it was 661 pg/ml. The associations found suggest that in the controls the highest ghrelin quartile protected from CHD (coronary heart disease). The results were significant without or with adjustments for age, sex, smoking, systolic blood pressure and LDL cholesterol, BMI, type 2 diabetes or QUICK index. C/C variant of the promoter associated with the prevention of IHD (ischemic heart disease) in the hypertensive group (pghrelin concentration was related to protection from CHD and Leu72Leu genotype to prevention of cancer in healthy adults during the 19 years follow-up. C/C promoter protects from IHD in the hypertensive subjects. Copyright © 2014 Elsevier Inc. All rights reserved.

  7. The status of HEU to LEU core conversion activities at the Jamaica SLOWPOKE

    Energy Technology Data Exchange (ETDEWEB)

    Preston, J.; Grant, C., E-mail: john.preston@uwimona.edu.jm [Univ. of the West Indies, Mona Campus, International Centre for Environmental and Nuclear Sciences, Mona (Jamaica)

    2012-12-15

    The SLOWPOKE reactor in Jamaica has been operated by the International Centre for Environmental and Nuclear Sciences, University of the West Indies since 1984, mainly for the purpose of Neutron Activation Analysis. The HEU core with current utilization has another 14 years of operation, before the addition of a large beryllium annulus would be required to further extend the life-time by 15 years. However, in keeping with the spirit of the Reduced Enrichment for Research and Test Reactors (RERTR) program, the decision was taken in 2003 to convert the core from HEU to LEU, in line with those at the Ecole Polytechnic and RMC SLOWPOKE facilities. This paper reports on the current status of the conversion activities, including key fuel manufacture and regulatory issues, which have seen substantial progress during the last year. A timetable for the complete process is given, and provided that the fuel fabrication can be completed in the estimated 18 months, the core conversion should be accomplished by the end of 2014. (author)

  8. The status of HEU and LEU core conversion activities at the Jamaica SLOWPOKE

    Energy Technology Data Exchange (ETDEWEB)

    Preston, J.; Grant, C., E-mail: john.preston@uwimona.edu.jm [Univ. of the West Indies, Mona Campus, International Centre for Environmental and Nuclear Sciences, Kingston (Jamaica)

    2013-07-01

    The SLOWPOKE reactor in Jamaica has been operated by the International Centre for Environmental and Nuclear Sciences, University of the West Indies since 1984, mainly for the purpose of Neutron Activation Analysis. The HEU core with current utilization has another 14 years of operation, before the addition of a large beryllium annulus would be required to further extend the life-time by 15 years. However, in keeping with the spirit of the Reduced Enrichment for Research and Test Reactors (RERTR) program, the decision was taken in 2003 to convert the core from HEU to LEU, inline with those at the Ecole Polytechnic and RMC SLOWPOKE facilities. This paper reports on the current status of the conversion activities, including key fuel manufacture and regulatory issues, which have seen substantial progress during the last year. A timetable for the complete process is given, and provided that the fuel fabrication can be completed in the estimated 18 months, the core conversion should be accomplished by the end of 2014. (author)

  9. Preliminary radiological consequence estimates for a reference LEU core for PARR

    International Nuclear Information System (INIS)

    Ali Khan, L.

    1990-01-01

    Radiological consequence analysis of a reference LEU core for Pakistan Research Reactor (PARR) has been carried out using mathematical models. It was assumed that 20% of the fuel, having an average burn-up of 50% achieved by continuously operating the reactor for 300 days at 10 MW, fails. It was further assumed that 100% of the noble gases and a fraction of iodine are released. Three modes of leakage from reactor building have been considered. These are exhaust through the normal ventilation system, through emergency ventilation system and leakage from the building. The whole body and thyroid doses have been calculated for 2 hours and 30 days at the boundaries of the exclusion zone at 450m and the low population zone at 1000m. For the releases at stack height through normal and emergency ventilation system, doses at both the boundaries remain within emergency dose limits of 300 rem for thyroid and 25 rem for the whole body. However, in the case of direct release from the containment building, the limiting thyroid dose of 300 rem, at 1000m, for 30 days exposure is achieved for a leak rate of 27% per day under Pasquill condition E. The results presented in this report are only preliminary estimates. A more accurate detailed analysis for various burnups will be carried out using standard computer codes

  10. Preliminary radiological consequence estimate for a reference LEU core for PARR

    International Nuclear Information System (INIS)

    Younus, M.; Khan, L.A.; Akhtar, K.M.; Pervez, S.

    1988-07-01

    Radiological consequence analysis of a reference LEU core for Pakistan Research Reactor (PARR) has been carried out using mathematical models. Three modes of leakage from reactor building have been considered. These are exhaust through the normal ventilation system, through emergency ventilation system and leakage from the building. The whole body and thyroid does have been calculated for the duration of 2 hours and 30 days at the boundaries of exclusion zone at 450m and low population zone at 100m. For the releases at stack height through normal and emergency ventilation systems, does at both the boundaries remain within relevant emergency does limits of 300 rem for thyroid and 25 rem for whole body. However, in case of direct release from the containment building, limiting thyroid does of 300 rem, at 1000m, for 30 days exposure is achieved for a leak rate of 27% per day under Pasquill condition E. The results presented in this report are only preliminary estimates. A more accurate detailed analysis, for various burnups, will be carried using standard computer codes. (orig./A.B)

  11. Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

    Energy Technology Data Exchange (ETDEWEB)

    Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

    1985-07-01

    This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

  12. Implementation of a Gadolinium Burnable Absorber in the Carbide LEU-NTR

    International Nuclear Information System (INIS)

    Venneria, Paolo; Kim, Yonghee

    2015-01-01

    Among the most crucial are the rapid reactivity depletion during full-power operation and the positive reactivity insertion during the full-submersion criticality accident. In previous work, it has been suggested that both challenges can be mitigated through the successful implementation of a burnable absorber in the active core. Of the poisons previously surveyed, one of the most promising is Gadolinium in the form of Gadolina (Gd2O4). This paper explores the possibility of different methods by which the Gadolinia can be implemented in the core and makes a preliminary study of its effect on the full submersion criticality accident and the reactivity depletion during operation. The application of a Gadolinium neutron absorber in the active core region of the LEU-NTR has been shown to be neutronically feasible. It can be introduced into the core in various locations without resulting in core performance loss. The utility of the poison in terms of mitigating the full-submersion reactivity accident and the rapid change in reactivity during full-power operation have been preliminarily shown and the first steps towards eventual implementation made. Future work will consist of determining the maximum poison content in the core and tailoring the self-shielding effect in order to determine a specific Gd depletion rate

  13. TOLL-LIKE RECEPTOR 7 GENE Gln11Leu MISSENSEMUTATION AND SUSCEPTIBILITY TO PSORIASIS

    Directory of Open Access Journals (Sweden)

    E. S. Galimova

    2017-01-01

    Full Text Available Toll-like receptor (TLR are responsible for recognizing various molecular patterns associated with pathogens. Their expression have been detected in skin cells such as keratinocytes and melanocytes. Numerous experimental studies demonstrate the key role of TLRs in the pathogenesis of immune diseases, including psoriasis. The objective of this study is to analyze the associations of polymorphisms in TLR7 gene and the risk of psoriasis development. DNA samples were collected from 138 patients with psoriasis and 317 healthy controls. Genotyping of rs179003, rs179008, rs179020, rs850632, rs12013728 polymorphic loci in TLR7 gene was performed using the SNPlex™method (AB, USA. SNP in the TLR7 gene rs179008 (Gln11Leu was associated with psoriasis in entire psoriasis, late onset and sporadic subgroups (Рс = 0.0065, OR = 1.95; Рс = 0.0004, OR = 2.50; Рс = 0.0078, OR = 2.2, respectively. In conclusion, this study is the first to identify genetic variants of the TLR7 gene significantly associated with psoriasis. 

  14. An investigative approach to explore optimum assembly process design for annular targets carrying LEU foil

    Science.gov (United States)

    Hoyer, Annemarie

    Technetium-99m is the most widely used nuclear isotope in the medical field, with nearly 80 to 85% of all diagnostic imaging procedures. The daughter isotope of molybdenum-99 is currently produced using weapons-grade uranium. A suggested design for aluminum targets carrying low-enriched uranium (LEU) foil is presented for the fulfillment of eliminating highly enriched uranium (HEU) for medical isotope production. The assembly process that this research focuses on is the conventional draw-plug process which is currently used and lastly the sealing process. The research is unique in that it is a systematic approach to explore the optimal target assembly process to produce those targets with the required quality and integrity. Conducting 9 parametric experiments, aluminum tubes with a nickel foil fission-barrier and a surrogate stainless steel foil are assembled, welded and then examined to find defects, to determine residual stresses, and to find the best cost-effective target dimensions. The experimental design consists of 9 assembly combinations that were found through orthogonal arrays in order to explore the significance of each factor. Using probabilistic modeling, the parametric study is investigated using the Taguchi method of robust analysis. Depending on the situation, optimal conditions may be a nominal, a minimized or occasionally a maximized condition. The results will provide the best target design and will give optimal quality with little or no assembly defects.

  15. Leu452His mutation in lipoprotein lipase gene transfer associated with hypertriglyceridemia in mice in vivo.

    Directory of Open Access Journals (Sweden)

    Kaiyue Sun

    Full Text Available Mutated mouse lipoprotein lipase (LPL containing a leucine (L to histidine (H substitution at position 452 was transferred into mouse liver by hydrodynamics-based gene delivery (HD. Mutated-LPL (MLPL gene transfer significantly increased the concentrations of plasma MLPL and triglyceride (TG but significantly decreased the activity of plasma LPL. Moreover, the gene transfer caused adiposis hepatica and significantly increased TG content in mouse liver. To understand the effects of MLPL gene transfer on energy metabolism, we investigated the expression of key functional genes related to energy metabolism in the liver, epididymal fat, and leg muscles. The mRNA contents of hormone-sensitive lipase (HSL, adipose triglyceride lipase (ATGL, fatty acid-binding protein (FABP, and uncoupling protein (UCP were found to be significantly reduced. Furthermore, we investigated the mechanism by which MLPL gene transfer affected fat deposition in the liver, fat tissue, and muscle. The gene expression and protein levels of forkhead Box O3 (FOXO3, AMP-activated protein kinase (AMPK, and peroxisome proliferator-activated receptor-gamma coactivator 1 alpha (PGC-1α were found to be remarkably decreased in the liver, fat and muscle. These results suggest that the Leu452His mutation caused LPL dysfunction and gene transfer of MLPL in vivo produced resistance to the AMPK/PGC-1α signaling pathway in mice.

  16. TOLL-LIKE RECEPTOR 7 GENE Gln11Leu MISSENSEMUTATION AND SUSCEPTIBILITY TO PSORIASIS

    Directory of Open Access Journals (Sweden)

    E. S. Galimova

    2017-01-01

    Full Text Available Toll-like receptor (TLR are responsible for recognizing various molecular patterns associated with pathogens. Their expression have been detected in skin cells such as keratinocytes and melanocytes. Numerous experimental studies demonstrate the key role of TLRs in the pathogenesis of immune diseases, including psoriasis. The objective of this study is to analyze the associations of polymorphisms in TLR7 gene and the risk of psoriasis development. DNA samples were collected from 138 patients with psoriasis and 317 healthy controls. Genotyping of rs179003, rs179008, rs179020, rs850632, rs12013728 polymorphic loci in TLR7 gene was performed using the SNPlex™method (AB, USA. SNP in the TLR7 gene rs179008 (Gln11Leu was associated with psoriasis in entire psoriasis, late onset and sporadic subgroups (Рс = 0.0065, OR = 1.95; Рс = 0.0004, OR = 2.50; Рс = 0.0078, OR = 2.2, respectively. In conclusion, this study is the first to identify genetic variants of the TLR7 gene significantly associated with psoriasis.

  17. GPX1 Pro(198)Leu polymorphism, erythrocyte GPX activity, interaction with alcohol consumption and smoking, and risk of colorectal cancer

    DEFF Research Database (Denmark)

    Hansen, Rikke Dalgaard; Krath, Britta N.; Frederiksen, Kirsten

    2009-01-01

    polymorphism and several lifestyle factors predict GPX activity in erythrocytes. The present study was nested within the prospective “Diet, Cancer and Health” study of 57,053 Danes including 375 colorectal cancer cases and a comparison group of 779 individuals matched on gender. Biomaterial was sampled...... and information on lifestyle factors was obtained from questionnaires filled in at enrolment in 1993–1997. GPX1 Pro198Leu, hOGG1 Ser326Cys and erythrocyte GPX enzyme activity were not associated with risk of colorectal cancer. We observed a higher risk associated with alcohol consumption and smoking among......198Leu genotype, gender, smoking intensity, and intake of fruits and vegetables. Our results indicate that lifestyle-related oxidative stress may be a risk factor for colorectal cancer among subjects with a lowered defence....

  18. The use of LeuT as a model in elucidating binding sites for substrates and inhibitors in neurotransmitter transporters

    DEFF Research Database (Denmark)

    Løland, Claus Juul

    2015-01-01

    Background: The mammalian neurotransmitter transporters are complex proteins playing a central role in synaptic transmission between neurons by rapid reuptake of neurotransmitters. The proteins which transport dopamine, noradrenaline and serotonin belong to the Neurotransmitter:Sodium Symporters...... (NSS). Due to their important role, dysfunctions are associated with several psychiatric and neurological diseases and they also serve as targets for a wide range of therapeutic and illicit drugs. Despite the central physiological and pharmacological importance, direct evidence on structure......–function relationships on mammalian NSS proteins has so far been unsuccessful. The crystal structure of the bacterial NSS protein, LeuT, has been a turning point in structural investigations. Scope of review: To provide an update on what is known about the binding sites for substrates and inhibitors in the Leu...

  19. 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)

  20. Neutronic calculations in core conversion of the IAN-R1 research reactor from MTR HEU to TRIGA LEU fuel

    International Nuclear Information System (INIS)

    Sarta Fuentes, Jose A.; Castiblanco, L.A.

    2003-01-01

    With cooperation of the International Atomic Energy Agency (IAEA), neutronic calculations were carried out for conversion of the Ian-R1 Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to establish a staff for neutronic calculation at the Instituto de Cancan's Nucleares y Energia s Alternatives (INEA) a program was established. This program included training, acquisition of hardware, software and calculation for the core with MTR-HEU fuel , enriched nominally to 93% and calculation for several arrangements with the TRIGA-LEU fuel, enriched to 19.7%. The results were verified and compared with several groups of calculation at the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, and General Atomics (GA) in United States. As a result of this program, several technical reports have been wrote. (author)