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)

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)

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)

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)

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.

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.

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)

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)

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

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)

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)

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

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

LEU{sub b}ased Fission Mo-99 Process with Reduced Solid Wastes

Energy Technology Data Exchange (ETDEWEB)

Lee, Seungkon; Lee, Suseung; Jung, Sunghee; Hong, Soonbog; Jang, Kyungduk; Choi, Sang Mu; Lee, Jun Sig; Lim, Incheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

{sup 99m} Tc emits 140 keV of very low gamma-ray radiation energy, as low as conventional diagnostic X-ray, and has short half-life of 6.0058 hours. Therefore, as radioactive tracer, {sup 99m} Tc provides high quality diagnostic images but keeps total patient radiation exposure low. Depending on the tagging pharmaceuticals and procedures, {sup 99m} Tc can be applied for the diagnostics of various target organs and diseases: brain, myocardium, thyroid, lungs, liver, gallbladder, kidneys, skeleton, blood and tumors. More than 95% of {sup 99}Mo is produced through fission of {sup 235}U worldwide because, {sup 99m}o generated from the fission (fission {sup 99}Mo) exhibits very high specific activity (<100 Ci/g). Over 90% of fission {sup 99}Mo producers have been used highly enriched uranium (HEU) targets so far. However, the IAEA recommends the use of low enriched uranium (LEU) to the {sup 99}Mo producers for nonproliferation reason. These days, worldwide {sup 99}Mo supply is not only insufficient but also unstable. Because, most of the main {sup 99}Mo production reactors are about 50 years old and suffered from frequent and unscheduled shutdown. Planned weekly productivity of 2000 Ci fission {sup 99}Mo, in a 6-day reference, will cover 100% domestic demand of Korea, as well as 20% of international market. It is expected to replace 4.3 million USD ($800/Ci) of {sup 99}Mo import for domestic market while exporting 82.8 million USD for world market, annually.

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.

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

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.

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.

Comment on the contribution of S.C. Mo, N.A. Hanan and J.E. Matos: 'Comparison of the FRM-II HEU design with an alternative LEU design'

International Nuclear Information System (INIS)

Boening, K.

2004-01-01

The results of the reference paper, which came to our attention for the first time during this RERTR Meeting, are more or less consistent with neutronic data we have obtained earlier within the FRM-II project (i.e. with own calculations and extrapolations). However, a realistic comparison of the HEU design of the FR.M-II (HEU = highly enriched uranium, 93 % U-235) with an alternative LEU design (LEU = low enriched uranium, 20 % U-235) is only possible on the basis of identical assumptions on the input parameters and has to consider more than neutronic data only. Serious scientists and experts should not confuse the politicians with academic studies touching some aspects of the full story only. The comparison has shown that the performance and reliability of the FRM-II design, which uses HEU fuel, is so advantageous that it can not - not even approximately - be met by an alternative design using LEU fuel. A change of the FRM-II design from HEU to LEU fuel with the results as shown above - i.e. less performance, higher costs, more nuclear waste and higher risk potential, and all of this with a delay of at least 5 years this could never be justified. If a future development of more advanced fuels should allow us to achieve our scientific goals at the conditions as identified above also with uranium of reduced enrichment - there would be no objection to a corresponding later conversion. Activities to realize a new neutron source in Germany go back to the late 70's with the project of a new middle flux beam reactor (MSR), which was abandoned shortly later in favour of an ambitious new spallation neutron source (SNQ). After this project also having been terminated around 1985 because of too high costs and technological risks, the hopes of the German community of neutron scientists focussed on the FRM-II. If non-technical pressure would damage this project this would equally provide irreversible damage to the large and still prospering field of neutron research in Germany

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)

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)

Development of Commercial-scale Fission Mo-99 Production System

Energy Technology Data Exchange (ETDEWEB)

Lee, Seung-Kon; Lee, Suseung; Hong, Soon-Bog; Jang, Kyung-Duk; Park, Ul Jael; Lee, Jun Sig [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

These days, worldwide {sup 99} Mo supply is not only insufficient but also unstable. Because, most of the main {sup 99}Mo production reactors are more than years old and suffered from frequent and unscheduled shutdown. Therefore, movement to replace old reactors to keep stable supply is now active. Under these conditions, 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. 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. For the fission Mo production process development, hot experiments with irradiated LEU targets will be done in 4th quarter of 2016. Then, verification of the production process with quality control will be followed until the commercial production of fission {sup 99}Mo scheduled in 2019.

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

International Nuclear Information System (INIS)

1995-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

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

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.

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

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)

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.

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)

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.

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

Alternative energy resources for MoDOT

Science.gov (United States)

2011-02-01

This research investigates environmentally friendly alternative energy sources that could be used by MoDOT in various areas, and develops applicable and sustainable strategies to implement those energy sources.

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)

Development of fission Mo-99 production technology

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Ho; Choung, W. M.; Lee, K. I. and others

2001-05-01

This R and D project is planed to supply domestic demands of Mo-99 through fission route, and consequently this project will be expected to rise up utilization of HANARO and KAERI's capability for marketing extension into domestic and oversea radiopharmaceutical market. HEU and LEU target types are decided and designed for fission Mo-99 production in domestic. Experimental study of target fabrication technology was performed and developed processing equipments. And conceptual design of target loading/unloading in/from HANARO device are performed. Tracer test of Mo-99 separation and purification process was performed, test results reach to Mo-99 recovery yield above 80% and decontamination factor above 1600. Combined Mo-99 separation and purification process was decided for hot test scheduled from next year, and performance test was performed. Conceptual design for modification of existing hot cell for fission Mo-99 production facility was performed and will be used for detail design. Assumption for the comparison of LEU and HEU target in fission Mo-99 production process were suggested and compared of merits and demerits in view of fabrication technology and economy feasibility.

Development of fission Mo-99 production technology

International Nuclear Information System (INIS)

Park, Jin Ho; Choung, W. M.; Lee, K. I. and others

2001-05-01

This R and D project is planed to supply domestic demands of Mo-99 through fission route, and consequently this project will be expected to rise up utilization of HANARO and KAERI's capability for marketing extension into domestic and oversea radiopharmaceutical market. HEU and LEU target types are decided and designed for fission Mo-99 production in domestic. Experimental study of target fabrication technology was performed and developed processing equipments. And conceptual design of target loading/unloading in/from HANARO device are performed. Tracer test of Mo-99 separation and purification process was performed, test results reach to Mo-99 recovery yield above 80% and decontamination factor above 1600. Combined Mo-99 separation and purification process was decided for hot test scheduled from next year, and performance test was performed. Conceptual design for modification of existing hot cell for fission Mo-99 production facility was performed and will be used for detail design. Assumption for the comparison of LEU and HEU target in fission Mo-99 production process were suggested and compared of merits and demerits in view of fabrication technology and economy feasibility

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

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

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

International Nuclear Information System (INIS)

1995-09-01

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

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

Separation of fission 99Mo by alpha-benzoin oxime precipitation in nitric medium

International Nuclear Information System (INIS)

Yamaura, Mitiko; Freitas, Antonio A.; Egute, Nayara dos S.; Camilo, Ruth L.; Araujo, Izilda C.; Forbicini, Christina A.L.G. de O.

2011-01-01

Since 2009, the production of generators 99 Mo/ 99 mTc suffers a crisis of global supply due to technical problems of the two reactors which account for 64% of world production of fission 99 Mo. By the project of Brazilian Multipurpose Reactor (RMB), the Brazilian government invests in the construction of the first multipurpose reactor suitable for the domestic production of 99 Mo from LEU targets in order to supply of fission 99 Mo in the coming decades. The IPEN started the research of the technology and production of fission 99 Mo from acid and alkaline dissolutions of Low Enriched Uranium (LEU) targets as well as other used radioisotopes in nuclear medicine. This work is part of the research of the technology of the fission 99 Mo from acid dissolution of the LEU targets that is being developed at the IPEN. In this study the separation of the Mo by precipitation with alpha-benzoin oxime in nitric medium and the recovery by dissolution were investigated. The precipitation studies were performed by batch assays with nitric solution of Mo(VI), containing 99 Mo tracer, and uranyl ions. Influence of concentration of permanganate from 0.03 to 2.5%, dissolution temperature at 30 deg C and 150 deg C and the uranium concentration from 74 g.L -1 to 115 g.L -1 was studied. Results indicated that the precipitation of Mo with alpha-benzoin oxime from nitric medium is highly efficient, and its recovery by dissolution with basic solution of H 2 O 2 gave a high yield. (author)

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)

Development of the Mo-99 process at CRNL

International Nuclear Information System (INIS)

Burrill, K.A.; Harrison, R.J.

1987-11-01

Highly enriched uranium (HEU) is used for Mo-99 production at CRNL. Dissolution of the targets and loading of the solution onto AI 2 0 3 columns is discussed. Development work continues to reduce processing time and overall product cost. A process for treating the fission product waste has been selected and a facility for processing is being designed. Low enriched uranium (LEU) is planned for targets eventually. Our experience with Si-based fuel for targets is poor, and alternatives are being sought

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

Development of annular targets for 99Mo production

International Nuclear Information System (INIS)

Conner, C.; Lewandowski, E.F.; Snelgrove, J.L.; Liberatore, M.W.; Walker, D.E.; Wiencek, T.C.; McGann, D.J.; Hofman, G.L.; Vandegrift, G.F.

1999-01-01

During 1999, significant progress was made in the development of a low-enriched uranium (LEU) target for production of 99 Mo. Successful conversion requires an inexpensive, reliable target. To keep the target geometry the same when changing from high-enriched uranium (HEU) to LEU targets, a denser form of uranium is required in order to increase the amount of uranium per target by a factor of approximately five. Targets containing LEU in the form of a metal foil are being developed for producing 99 Mo from the fissioning of 235 U. A new annular target was developed this year, and seven targets were irradiated in the Indonesian RSG-GAS reactor. Results of development of this annular target and its performance during irradiation are described. (author)

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

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)

Using molybdenum depleted in 95Mo in UMo fuel

International Nuclear Information System (INIS)

Bakker, K.; Wijtsma, F.; Bos, A.; Mol, C.; Rakhorst, H.; Bretscher, M.; Hofman, G.; Snelgrove, J.

2002-01-01

In recent years significant interest was gained in UMo fuel to be used in Material Test Reactors. This interest was induced by the fact that UMo fuel is mechanically stable, even at high uranium concentrations and high U-burnup. These properties are required in order to use Low Enriched Uranium (LEU) and still be able to achieve high flux and burnup values and, thus, to facilitate the conversion from High Enriched Uranium (HEU) to LEU. Neutronics computations have shown that, although the Mo concentration in UMo fuel is not very high (about 5 - 10w%), the neutron absorption cross sections of natural Mo are sufficiently high to have a considerable negative impact on the reactivity of this UMo fuel. In the present research the neutron absorption cross sections of natural Mo are discussed and the option to reduce the cross section of molybdenum by depleting the Mo in 95 Mo is described. Finally the economic consequences of using Mo depleted in 95 Mo are briefly discussed

Production of Mo-99 using low-enriched uranium silicide

International Nuclear Information System (INIS)

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

1994-01-01

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

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

Directory of Open Access Journals (Sweden)

Seung-Kon Lee

2016-06-01

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

Low enrichment Mo-99 target development program at ANSTO

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-15

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

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.

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

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)'

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Neutronic 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)'.

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

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.

Feasibility study on mass production of (n,γ)99Mo

International Nuclear Information System (INIS)

Jun, Byung Jin; Tanimoto, Masataka; Kimura, Akihiro; Hori, Naohiko; Izumo, Hironobu; Tsuchiya, Kunihiko

2011-01-01

The world is currently suffering from a severe shortage of 99 Mo and various efforts have been given for its availability. The (n,γ) method is one of candidates for the alternative supply of 99 Mo. The only but critical shortage of (n,γ) 99 Mo is its extremely low specific activity, which gives inconveniency in the extraction of 99m Tc and is consequently converted to additional cost. Potential technologies which make the (n,γ) 99 Mo competitive by reducing the additional cost are already available. It is expected that verification of such technologies is much easy and cost effective compared to any other options known for the alternative 99 Mo production. Because Japan and Korea import all 99 Mo from long distance, the cost benefit of local (n,γ) 99 Mo production in these countries is especially large. If five high flux reactors in China, Japan and Korea are utilized for the cross backup supply of (n,γ) 99 Mo, stable availability of 99 Mo in the region can be secured. Therefore, it is necessary to evaluate its feasibility on (n,γ) 99 Mo production in the Asia region. In this report, we studied feasibility of the mass (n,γ) 99 Mo production from viewpoints of global and regional status of 99 Mo demand and supply, competitiveness with other production methods, requirements and flow of the 99 Mo, production capability, cost, convenience in usage, and alternative technologies to overcome its shortage. (author)

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)

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.

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.

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

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)

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

AMORE Mo-99 Spike Test Results

Energy Technology Data Exchange (ETDEWEB)

Youker, Amanda J. [Argonne National Lab. (ANL), Argonne, IL (United States); Krebs, John F. [Argonne National Lab. (ANL), Argonne, IL (United States); Quigley, Kevin J. [Argonne National Lab. (ANL), Argonne, IL (United States); Byrnes, James P. [Argonne National Lab. (ANL), Argonne, IL (United States); Rotsch, David A [Argonne National Lab. (ANL), Argonne, IL (United States); Brossard, Thomas [Argonne National Lab. (ANL), Argonne, IL (United States); Wesolowski, Kenneth [Argonne National Lab. (ANL), Argonne, IL (United States); Alford, Kurt [Argonne National Lab. (ANL), Argonne, IL (United States); Chemerisov, Sergey [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, George F. [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-09-27

With funding from the National Nuclear Security Administrations Material Management and Minimization Office, Argonne National Laboratory (Argonne) is providing technical assistance to help accelerate the U.S. production of Mo-99 using a non-highly enriched uranium (non-HEU) source. A potential Mo-99 production pathway is by accelerator-initiated fissioning in a subcritical uranyl sulfate solution containing low enriched uranium (LEU). As part of the Argonne development effort, we are undertaking the AMORE (Argonne Molybdenum Research Experiment) project, which is essentially a pilot facility for all phases of Mo-99 production, recovery, and purification. Production of Mo-99 and other fission products in the subcritical target solution is initiated by putting an electron beam on a depleted uranium (DU) target; the fast neutrons produced in the DU target are thermalized and lead to fissioning of U-235. At the end of irradiation, Mo is recovered from the target solution and separated from uranium and most of the fission products by using a titania column. The Mo is stripped from the column with an alkaline solution. After acidification of the Mo product solution from the recovery column, the Mo is concentrated (and further purified) in a second titania column. The strip solution from the concentration column is then purified with the LEU Modified Cintichem process. A full description of the process can be found elsewhere [1–3]. The initial commissioning steps for the AMORE project include performing a Mo-99 spike test with pH 1 sulfuric acid in the target vessel without a beam on the target to demonstrate the initial Mo separation-and-recovery process, followed by the concentration column process. All glovebox operations were tested with cold solutions prior to performing the Mo-99 spike tests. Two Mo-99 spike tests with pH 1 sulfuric acid have been performed to date. Figure 1 shows the flow diagram for the remotely operated Mo-recovery system for the AMORE project

An alternative route for the preparation of the medical isotope 99Mo from the 238U(γ, f) and 100Mo(γ, n) reactions

International Nuclear Information System (INIS)

Naik, H.; Goswami, A.; Suryanarayana, S.V.; Jagadeesan, K.C.; Thakare, S.V.; Joshi, P.V.; Nimje, V.T.; Mittal, K.C.; Venugopal, V.; Kailas, S.

2013-01-01

The radionuclide 99 Mo, which has a half-life of 65.94 h was produced from 238 U(γ, f) and 100 Mo(γ, n) reactions using a 10 MeV electron linac at EBC, Kharghar Navi-Mumbai, India. This has been investigated since the daughter product 99m Tc is very important from a medical point of view and can be produced in a generator from the parent 99 Mo. The activity of 99 Mo was analyzed by a γ-ray spectrometric technique using a HPGe detector. From the detected γ-rays activity of 140.5 and 739.8 keV, the amount of 99 Mo produced was determined. For comparison, the amount of 99 Mo from 238 U(γ, f) and 100 Mo(γ, n) reactions was also estimated using the experimental photon flux from 197 Au(γ, n) 196 Au reaction. The amount of 99 Mo from the detected γ-lines is in agreement with the estimated value for 238 U(γ, f) and 100 Mo(γ, n) reactions. The production of 99 Mo activity from 238 U(γ, f) and 100 Mo(γ, n) reactions is a relevant and novel approach, which provides alternative routes to 235,238 U(n, f) and 98 Mo(n, γ) reactions, circumventing the need for a reactor. The viability and practicality of the 99 Mo production from the 238 U(γ, f) and 100 Mo(γ, n) reactions alternative to 235,238 U(n, f) and 98 Mo(n, γ) reactions has been emphasize. An estimate has been also arrived based on the experimental data of present work to fulfill the requirement of DOE. (author)

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

Studies on molybdenum elution study in dowex 1x8 resin applied on purification process of fission 99Mo

International Nuclear Information System (INIS)

Damasceno, M.O.; Yamaura, M.; Santos, J.L. dos; Forbicini, C.A.L.G. de O.

2013-01-01

Molybdenum-99 is the most widely employed radioisotope in nuclear medicine, due to its decay product, Technetium-99, which is used in radio-pharmaceutical marking molecules for diagnostic examinations tumor dis-eases. Today Brazil imports 99 Mo from some countries, so the National Commission of Nuclear Energy (CNEN) is implementing a new research reactor RMB, currently in the conceptual design phase. The process of separation of fission 99 Mo begins with the dissolution of uranium targets after irradiation in reactor; the resulting solution goes through a series of chromatographic columns that allows a gradual decontamination of other components, yielding the 99 Mo with high radio-chemical and chemical purity for use in nuclear medicine as a generator of 99 mTc. This work is part of the RMB research project to separate and purify the fission 99 Mo by chromatographic columns from alkaline dissolution of LEU UAl x -Al targets. In the present study Mo removal by batch assays and glass column was investigated using anionic exchanger Dowex 1x8. Different salts and its concentration, cations and temperature were evaluated on elution of molybdenum and iodine (contaminant) retained on resin Dowex 1x8, aiming at their use in the process of separation and purification in chromatography columns on Brazilian project. Results showed high recovery of Mo and low-level contamination by iodine using NaHCO 3 hot solution. (author)

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)

Development of 99Mo isotope production targets employing uranium metal foils

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

Kuperman, A. J.

2005-01-01

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

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)

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)

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)

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.

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)

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)

Thermal compatibility of U-2wt.%Mo and U-10wt.%Mo fuel prepared by centrifugal atomization for high density research reactor fuels

International Nuclear Information System (INIS)

Kim Ki Hwan; Lee Don Bae; Kim Chang Kyu; Kuk Il Hyun; Hofman, G.E.

1997-01-01

Research on the intermetallic compounds of uranium was revived in 1978 with the decision by the international research reactor community to develop proliferation-resistant fuels. The reduction of 93% 235 U (HEU) to 20% 235 U (LEU) necessitates the use of higher U-loading fuels to accommodate the addition 238 U in the LEU fuels. While the vast majority of reactors can be satisfied with U 3 Si 2 -Al dispersion fuel, several high performance reactors require high loadings of up to 8-9 g U cm -3 . Consequently, in the renewed fuel development program of the Reduced Enrichment for Research and Test Reactors (RERTR) Program, attention has shifted to high density uranium alloys. Early irradiation experiments with uranium alloys showed promise of acceptable irradiation behavior, if these alloys can be maintained in their cubic γ-U crystal structure. It has been reported that high density atomized U-Mo powders prepared by rapid cooling have metastable isotropic γ-U phase saturated with molybdenum, and good γ-U phase stability, especially in U-10wt.%Mo alloy fuel. If the alloy has good thermal compatibility with aluminium, and this metastable gamma phase can be maintained during irradiation, U-Mo alloy would be a prime candidate for dispersion fuel for research reactors. In this paper, U-2w.%Mo and U-10w.%Mo alloy powder which have high density (above 15 g-U/cm 3 ), are prepared by centrifugal atomization. The U-Mo alloy fuel meats are made into rods extruding the atomized powders. The characteristics related to the thermal compatibility of U-2w.%Mo and U-10w.%Mo alloy fuel meat at 400 o C for time up to 2000 hours are examined. (author)

Development of fission Mo-99 production technology

International Nuclear Information System (INIS)

Park, Jin Ho; Choung, W. M.; Lee, K. I. and others

2000-05-01

Fission Mo-99 is the only parent nuclide of Tc-99m, an extremely useful tool for mdeical diagnosis, with an estimated usage of greater than 80% of nuclear medicine applicatons. HEU and LEU targets to optimize in HANARO irradiation condition suggested and designed for domestic production of fission Mo-99. The optimum process conditions are established in each unit process to meet quality requirements of fission Mo-99 products, and the results of performance test in combined process show Mo separation and purification yield of the above 97%. The concept of Tc generator production process is established, and the result of performance test show Tc production yield of 98.4% in Tc generator procuction process. The drafts is prepared for cooperation of technical cooperation and business investment with foreign country. Evaluation on economic feasibility is accompanied for fission Mo-99 and Tc-99m generator production

Development of fission Mo-99 production technology

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Ho; Choung, W. M.; Lee, K. I. and others

2000-05-01

Fission Mo-99 is the only parent nuclide of Tc-99m, an extremely useful tool for mdeical diagnosis, with an estimated usage of greater than 80% of nuclear medicine applicatons. HEU and LEU targets to optimize in HANARO irradiation condition suggested and designed for domestic production of fission Mo-99. The optimum process conditions are established in each unit process to meet quality requirements of fission Mo-99 products, and the results of performance test in combined process show Mo separation and purification yield of the above 97%. The concept of Tc generator production process is established, and the result of performance test show Tc production yield of 98.4% in Tc generator procuction process. The drafts is prepared for cooperation of technical cooperation and business investment with foreign country. Evaluation on economic feasibility is accompanied for fission Mo-99 and Tc-99m generator production.

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)

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)

Progress on the Development of (n,γ)99Mo/99mTc generator, an Alternative of Conventional Generator

International Nuclear Information System (INIS)

Lee, Jun Sig; Choi, Seung Rag; Nam, Seong Su; Park, Ul Jae; Son, Kwang Jae; Choi, Kang Hyuk; Choi, Sun Ju

2010-01-01

Even though different types of generators have been developed to extract 99m Tc, most of the generators uses 99 Mo from the fission products as the mother radionuclide of 99m Tc. Recently, the crisis of 99 Mo production becomes one of the international issues as 99m Tc is a dominant diagnostic radionuclide. The shortage of 99m Tc has been predicted in the society for more than 10 years. However, actions to prevent such crisis were slow as the initial investment to construct a new research reactor for the production of 99 Mo is high. Currently, it is expected the shortage of 99m Tc will last at least for more than 5 years. As an alternative to minimize such crisis, a new approach is proposed and studied. In this approach, the mother source of 99m Tc comes not from fission products but from the neutron irradiation of molybdenum oxide. Hence, most of the research reactors, which do not have capability to produce fission molybdenum, (n,f) 99 Mo can produce 99 Mo. The key issue in this approach is the uptake capacity of the generator column for (n, γ) 99 Mo, which has extremely low specific activity compared to (n,f) 99 Mo. Currently, the results from the research activities at Korea Atomic Energy Research Institute have shown such approach has enough potential as an alternative of the conventional generator. Hence, the progress of the research is reported in this paper

Recent global crisis in 99Mo production and supplies: analysis and lessons

International Nuclear Information System (INIS)

Ramamoorthy, Natesan; Ed Bradley; Adelfang, Pablo; )

2011-01-01

The radioisotope 99m Tc is a vital need for diagnostic functional imaging of organs and physiologic processes in nuclear medicine centres and hospitals all over the world. 99m Tc (6 h) is obtained from the radioactive (beta) decay of 99 Mo (66 h) and is generally availed at the user end in hospitals or radiopharmacies from a 99 Mo- 99m Tc generator, and therefore reliable weekly availability of 99 Mo of very high specific activity is essential for sustainable supplies of 99m Tc. Over the past several decades, the commercial availability of 99 Mo (mostly fission-produced) from corporate sources has been satisfactory. However, from around the end of 2007, the production and supplies of 99 Mo were severely and repeatedly affected causing worldwide concerns and led to calls for international efforts to address the security of supplies of 99 Mo and 99m Tc. Various initiatives and analytical reviews during the past two years, including at the IAEA and OECD-NEA, have revealed a number of issues and lessons learnt. Research reactor managers and their governments serving 99 Mo- 99m Tc industry have faced several challenges, technological, regulatory and economical among others, apart from the fact that these reactors are not dedicated to isotopes production. Furthermore, towards long term security of 99 Mo supplies, switch over to using LEU targets and establishing some additional processing capacity need to be addressed. Several alternative non-HEU technologies for producing 99 Mo and 99m Tc are also under consideration, including through non-fission and non-reactor based methods. The article will draw analytical conclusions and recommendations based on the available professional knowledge and experience in the field as well as from the findings reported in the recent documents and in meetings including that of the IAEA and OECD-NEA. (author)

Development of fission Mo-99 production technology - A nuclear feasibility study on UN target for Mo-99 production in HANARO

Energy Technology Data Exchange (ETDEWEB)

Kim, Myung Hyun; Kim, Woo Sik [Kyunghee University, Seoul (Korea)

2000-03-01

Nuclear target design satisfying all the constraints for fission moly production in HANARO was proposed in this project. The 'MCNP-ORIGEN' code system which was previously proposed for a design tool, was evaluated by the comparison with through the 'MCNP-Analytic Eq.' system. A characteristics of each chemical processing step were analysed and material balance was set up to evaluate the overall yield ratio of Mo-99 recovery. A parametric study was done for the optimum HEU target design. Tested parameters were target thickness, recoil-loss rate to the fuel thickness, target radius, cladding materials, thickness of irradiation guide tube, and barrier materials. Optimized HEU target design was proposed which satisfying the constraints and having high production yield. For a LEU target design using 19.7 w/o UN powder fuel, a parametric study was also done for the optimization of fuel thickness, powder packing density, mixture material volume ratio. 24 refs., 35 figs., 57 tabs. (Author)

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.

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.

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.

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...

[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...

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

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)

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.

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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)

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

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)

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)

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)

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)

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.

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)

Progress in converting 99Mo production from high-to-low-enriched uranium - 1999

International Nuclear Information System (INIS)

Snelgrove, J.L.; Vandegrift, G.F.; Conner, C.; Wiencek, T.C.; Hofman, G.L.

1999-01-01

Over this past year, extraordinary progress has been made in executing our charter to assist in converting Mo-99 production worldwide from HEU to LEU. Building on the successful development of the experimental LEU-foil target, we have designed a new, economical irradiation target. We have also successfully demonstrated, in collaboration with BATAN in Indonesia, that LEU can be substituted for HEU in the Cintichem target without loss of product yield or purity; in fact, conversion may make economic sense. We are interacting with a number of commercial producers - we have begun active collaborations with the CNEA and ANSTO; we are working to define the scope of collaborations with MDS Nordion and Mallinckrodt; and IRE has offered its services to irradiate and test a target at the appropriate time. Conversion of the CNEA process is on schedule. Other papers presented at this meeting will present specific results on the demonstration of the LEU-modified Cintichem process, the development of the new target, and progress in converting the CNEA process. (author)

Study of the dynamics of the MoO2-Mo2C system for catalytic partial oxidation reactions

Science.gov (United States)

Cuba Torres, Christian Martin

On a global scale, the energy demand is largely supplied by the combustion of non-renewable fossil fuels. However, their rapid depletion coupled with environmental and sustainability concerns are the main drivers to seek for alternative energetic strategies. To this end, the sustainable generation of hydrogen from renewable resources such as biodiesel would represent an attractive alternative solution to fossil fuels. Furthermore, hydrogen's lower environmental impact and greater independence from foreign control make it a strong contender for solving this global problem. Among a wide variety of methods for hydrogen production, the catalytic partial oxidation offers numerous advantages for compact and mobile fuel processing systems. For this reaction, the present work explores the versatility of the Mo--O--C catalytic system under different synthesis methods and reforming conditions using methyl oleate as a surrogate biodiesel. MoO2 exhibits good catalytic activity and exhibits high coke-resistance even under reforming conditions where long-chain oxygenated compounds are prone to form coke. Moreover, the lattice oxygen present in MoO2 promotes the Mars-Van Krevelen mechanism. Also, it is introduced a novel beta-Mo2C synthesis by the in-situ formation method that does not utilize external H2 inputs. Herein, the MoO 2/Mo2C system maintains high catalytic activity for partial oxidation while the lattice oxygen serves as a carbon buffer for preventing coke formation. This unique feature allows for longer operation reforming times despite slightly lower catalytic activity compared to the catalysts prepared by the traditional temperature-programmed reaction method. Moreover, it is demonstrated by a pulse reaction technique that during the phase transformation of MoO2 to beta-Mo2C, the formation of Mo metal as an intermediate is not responsible for the sintering of the material wrongly assumed by the temperature-programmed method.

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.

Optimum nuclear design of target fuel rod for Mo-99 production in HANARO

Energy Technology Data Exchange (ETDEWEB)

Kim, Myung Hyun [Kyung Hee University, Seoul (Korea)

1998-04-01

Nuclear target design for Mo-99 production in HANARO was performed, KAERI proposed target design was analyzed and its feasibility was shown. Three commercial target designs of Cintichem, ANL and KAERI were tested for the HANARO irradiation an d they all satisfied with design specification. A parametric study was done for target design options and Mo-99 yields ratio and surface heat flux were compared. Tested parameters were target fuel thickness, irradiation location, target axial length, packing density of powder fuel, size of target radius, target geometry, fuel enrichment, fuel composition, and cladding material. Optimized target fuel was designed for both LEU and HEU options. (author). 17 refs., 33 figs., 42 tabs.

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)

U-8 wt %Mo and 7 wt %Mo alloys powder obtained by an hydride-de hydride process; Obtencion de polvo de aleaciones U-8% Mo y U-7% Mo (en peso) mediante hidruracion

Energy Technology Data Exchange (ETDEWEB)

Balart, Silvia N; Bruzzoni, Pablo; Granovsky, Marta S; Gribaudo, Luis M.J.; Hermida, Jorge D; Ovejero, Jose; Rubiolo, Gerardo H; Vicente, Eduardo E [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Materiales

2000-07-01

Uranium-molybdenum alloys are been tested as a component in high-density LEU dispersion fuels with very good performances. These alloys need to be transformed to powder due to the manufacturing requirements of the fuels. One method to convert ductile alloys into powder is the hydride-de hydride process, which takes advantage of the ability of the U-{alpha} phase to transform to UH{sub 3}: a brittle and relatively low-density compound. U-Mo alloys around 7 and 8 wt % Mo were melted and heat treated at different temperature ranges in order to partially convert {gamma} -phase to {alpha} -phase. Subsequent hydriding transforms this {alpha} -phase to UH{sub 3}. The volume change associated to the hydride formation embrittled the material which ends up in a powdered alloy. Results of the optical metallography, scanning electron microscopy, X-ray diffraction during different steps of the process are shown. (author)

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)

Photo-transmutation of {sup 100}Mo to {sup 99}Mo with Laser-Compton Scattering Gamma-ray

Energy Technology Data Exchange (ETDEWEB)

Lee, Jiyoung; Rehman, Haseeb ur; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

2016-10-15

This paper presents a photonuclear transmutation method using laser Compton scattering (LCS) gamma-ray beam. Potential production rate (reaction rate) of 99Mo using the photonuclear (γ,n) reaction is evaluated. Rigorous optimization of the LCS spectrum has also been performed to maximize production of the 99Mo. Cyclotron proton accelerators are used worldwide to produce many short-living medical isotopes. However, few are capable of producing Mo-99 and none are suitable for producing more than a small fraction of the required amounts. More than 90% of the world's demand of 99Mo is sourced from five nuclear reactors. Two of these reactors have already been decommissioned and the rest are more than 45 years old. Relatively short half-life of the parent 99Mo requires continuous re-supply to meet the requirements of medical industry. Therefore, there is an urgent need to produce the 99Mo and 99mTc isotopes by alternative ways. One such alternative is giant dipole resonance (GDR) based photonuclear transmutation of 100Mo to 99Mo. For 99Mo production with the LCS photons using GDR-based (γ,n) reaction, the gamma-ray energy should be around 15 MeV. This study indicates that optimization of LCS spectrum by varying the electron and laser energies within practical limits can enhance the transmutation of Mo-100 to M-99 quite significantly. It has been found that irradiation time should be rather short, e.g., less than 6 hours, to maximize the weekly production of Mo-99 in the GDR-based Mo-99 production facility using the LCS photons. The analysis shows that production of 99Mo using a high-performance LCS facility offers a potentially-promising alternative for the production of 99mTc.

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)

Molecular cloning of the α subunit of human and guinea pig leukocyte adhesion glycoprotein Mo1: Chromosomal localization and homology to the α subunits of integrins

International Nuclear Information System (INIS)

Arnaout, M.A.; Remold-O'Donnell, E.; Pierce, M.W.; Harris, P.; Tenen, D.G.

1988-01-01

The cell surface-glycoprotein Mo1 is a member of the family of leukocyte cell adhesion molecules (Leu-CAMs) that includes lymphocyte function-associated antigen 1 (LFA-1) and p150,95. Each Leu-CAM is a heterodimer with a distinct α subunit noncovalently associated with a common β subunit. The authors describe the isolation and analysis of two partial cDNA clones encoding the α subunit of the Leu-CAM Mo1 in humans and guinea pigs. A monoclonal antibody directed against an epitope in the carboxyl-terminal portion of the guinea pig α chain was used for immunoscreening a λgt11 expression library. The sequence of a 378-base-pair insert from one immunoreactive clone revealed a single continuous open reading frame encoding 126 amino acids including a 26-amino acid tryptic peptide isolated from the purified guinea pig α subunit. A cDNA clone of identical size was isolated from a human monocyte/lymphocyte cDNA library by using the guinea pig clone as a probe. The human clone also encoded a 126-amino acid peptide including the sequence of an additional tryptic peptide present in purified human Mo1α chain. Southern analysis of DNA from hamster-human hybrids localized the human Mo1α chain to chromosome 16, which has been shown to contain the gene for the α chain of lymphocyte function-associated antigen 1. These data suggest that the α subunits of Leu-CAMs evolved by gene duplication from a common ancestral gene and strengthen the hypothesis that the α subunits of these heterodimeric cell adhesion molecules on myeloid and lymphoid cells, platelets, and fibroblasts are evolutionary related

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)

[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

Study of methodologies for quality control of 99Mo used in 99Mo/99mTc generators

International Nuclear Information System (INIS)

Said, Daphne de Souza

2016-01-01

99m Tc is the most used radionuclide in nuclear medicine. In Brazil, the 99 Mo/ 99m Tc generators are exclusively produced by Radiopharmacy Center at IPEN-CNEN/ SP, by importing 99 Mo from different suppliers. 99 Mo (t 1/2 = 66 h) is a fission product of 235 U and it can have radionuclidic impurities that are prejudicial for human health. For safe use of generators, it is necessary to perform the evaluation of 99 Mo by quality control tests in order to assess if 99 Mo complies with the specifications. The European Pharmacopoeia (EP) presents a monograph for evaluation of the quality of the [ 99 Mo] solution as sodium molybdate,that is used as raw material for 99 Mo/ 99m Tc generators production, including specification parameters (identification, radiochemical purity and radionuclidic purity), analysis methods and limits. However, it has been observed difficulties on the execution and implementation of these methods by the generators producers, with a few literature about this subject, probably due to complexity of the proposed methods. In this work, many quality control parameters of 99 Mo described in the EP monograph were evaluated. Separation methods for 99M o from its radionuclidic impurities by solid phase extraction (SPE) and TLC were studied. After SPE separation, the quantification of metals by ICP-OES to evaluate the percentage of retention of Mo and the percentage of recovery of Ru, Te and Sr using different types of cartridges were proposed, replacing radiotracers use. It was observed that the specific type of SPE cartridge recommended by the EP for separation of 99 Mo presented low recoveries for Ru, compared to other available anion exchange SPE cartridges. 99 Mo samples from different worldwide suppliers were analyzed. It was observed that quantification of 103 Ru in 99 Mo samples with decay time higher than 4 weeks is possible. An alternative method for separation of 131 I from 99 Mo showed promising results by TLC. The quantification of beta and

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.

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.

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).

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

Radioactive decay pattern of actinides present in waste from Mo-99 production

Energy Technology Data Exchange (ETDEWEB)

Hiromoto, Goro; Dellamano, José Claudio, E-mail: hiromoto@ipen.br, E-mail: jcdellam@ipen.br [Instituto de PesquisasEnergéticas e Nucleares (GRR/IPEN/CNEN-SP), São Paulo, SP (Brazil). Gerência de Rejeitos Radioativos

2017-07-01

Brazil is currently planning to produce {sup 99}Mo from fission of LEU targets to meet the present national demand of {sup 99m}Tc. The {sup 99}Mo activity planned at the end of irradiation is 5000 Ci (185 TBq) per weekly cycle, in order to meet the present demand of 1000 Ci (37 TBq) per week, after target cooling and processing. To predict the activities that will be handled in the waste treatment facility, the computational code SCALE 6.0 was used to simulate the irradiation of the uranium targets and the decay of radioactive products. This study presents the findings of this research, mainly focused on the actinides activity that will be present in the waste and the respective radioactive decay pattern over a period of one hundred thousand years. (author)

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.

Waste Management Strategies for Production of Mo-99

Energy Technology Data Exchange (ETDEWEB)

Cozzi, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Johnson, F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-01-31

Production of Mo-99 for medical isotope use is being investigated using dissolved low enriched uranium (LEU) fissioned using an accelerator driven process. With the production and separation of Mo-99, a low level waste stream will be generated. Since the production facility is a commercial endeavor, waste disposition paths normally available for federally generated radioactive waste may not be available. Disposal sites for commercially generated low level waste are available, and consideration to the waste acceptance criteria (WAC) of the disposal site should be integral in flowsheet development for the Mo-99 production. Pending implementation of the “Uranium Lease and Take-Back Program for Irradiation for Production of Molybdenum-99 for Medical Use” as directed by the American Medical Isotopes Production Act of 2012, there are limited options for disposing of the waste generated by the production of Mo-99 using an accelerator. The commission of a trade study to assist in the determination of the most favorable balance of production throughput and waste management should be undertaken. The use of a waste broker during initial operations of a facility has several benefits that can offset the cost associated with using a subcontractor. As the facility matures, the development of in-house capabilities can be expanded to incrementally reduce the dependence on a subcontractor.

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)

U-8 wt %Mo and 7 wt %Mo alloys powder obtained by an hydride-de hydride process

International Nuclear Information System (INIS)

Balart, Silvia N.; Bruzzoni, Pablo; Granovsky, Marta S.; Gribaudo, Luis M. J.; Hermida, Jorge D.; Ovejero, Jose; Rubiolo, Gerardo H.; Vicente, Eduardo E.

2000-01-01

Uranium-molybdenum alloys are been tested as a component in high-density LEU dispersion fuels with very good performances. These alloys need to be transformed to powder due to the manufacturing requirements of the fuels. One method to convert ductile alloys into powder is the hydride-de hydride process, which takes advantage of the ability of the U-α phase to transform to UH 3 : a brittle and relatively low-density compound. U-Mo alloys around 7 and 8 wt % Mo were melted and heat treated at different temperature ranges in order to partially convert γ -phase to α -phase. Subsequent hydriding transforms this α -phase to UH 3 . The volume change associated to the hydride formation embrittled the material which ends up in a powdered alloy. Results of the optical metallography, scanning electron microscopy, X-ray diffraction during different steps of the process are shown. (author)

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)

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...

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

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.

Current status of production and supply of molybdenum-99 and 99Mo/99mTc generators in Indonesia

International Nuclear Information System (INIS)

Mutalib, A.

2003-01-01

Production of high-specific activity molybdenum-99 and 99 Mo/ 99m Tc Generators in Indonesia commenced when a new production facility supported by the presence of a 30 MW multipurpose reactor (RSG-GAS) was established in Serpong in 1990. This report describes the current production and supply of molybdenum-99m devoted mainly to fulfill the domestic demands in supplying 99 Mo/ 99m Tc Generators. Recent development on the use of LEU (Low Enriched Uranium) targets for replacing current HEU (High Enriched Uranium) targets in the production of 99 Mo will be reviewed briefly. (author)

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.

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

Analyses for inserting fresh LEU fuel assemblies instead of fresh HEU fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam

International Nuclear Information System (INIS)

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

2005-01-01

Analyses were performed by the RERTR Program to replace 36 burned HEU (36%) fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam with either 36 fresh fuel assemblies currently on-hand at the reactor or with LEU fuel assemblies to be procured. The study concludes that the current HEU (36%) WWR-M2 fuel assemblies can be replaced with LEU WWR-M2 fuel assemblies that are fully-qualified and have been commercially available since 2001 from the Novosibirsk Chemical Concentrates Plant in Russia. The current reactor configuration using re-shuffled HEU fuel began in June 2004 and is expected to allow normal operation until around August 2006. If 36 HEU assemblies each with 40.2 g 235 U are inserted without fuel shuffling over the next five operating cycles, the core could operate for an additional 10 years until June 2016. Alternatively, inserting 36 LEU fuel assemblies each containing 49.7 g 235 U without fuel shuffling over five operating cycles would allow normal operation for about 14 years from August 2006 until October 2020. The main reason for the longer service life of the LEU fuel is that its 235 U content is higher than the 235 U content needed simply to match the service life of the HEU fuel. Fast neutron fluxes in the experiment regions would be very nearly the same in both the HEU and LEU cores. Thermal neutron fluxes in the experiment regions would be lower by 1-5%, depending on the experiment type and location. (author)

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

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

LEUbased Fission Mo-99 Process with Reduced Solid Wastes

International Nuclear Information System (INIS)

Lee, Seungkon; Lee, Suseung; Jung, Sunghee; Hong, Soonbog; Jang, Kyungduk; Choi, Sang Mu; Lee, Jun Sig; Lim, Incheol

2014-01-01

99m Tc emits 140 keV of very low gamma-ray radiation energy, as low as conventional diagnostic X-ray, and has short half-life of 6.0058 hours. Therefore, as radioactive tracer, 99m Tc provides high quality diagnostic images but keeps total patient radiation exposure low. Depending on the tagging pharmaceuticals and procedures, 99m Tc can be applied for the diagnostics of various target organs and diseases: brain, myocardium, thyroid, lungs, liver, gallbladder, kidneys, skeleton, blood and tumors. More than 95% of 99 Mo is produced through fission of 235 U worldwide because, 99m o generated from the fission (fission 99 Mo) exhibits very high specific activity (<100 Ci/g). Over 90% of fission 99 Mo producers have been used highly enriched uranium (HEU) targets so far. However, the IAEA recommends the use of low enriched uranium (LEU) to the 99 Mo producers for nonproliferation reason. These days, worldwide 99 Mo supply is not only insufficient but also unstable. Because, most of the main 99 Mo production reactors are about 50 years old and suffered from frequent and unscheduled shutdown. Planned weekly productivity of 2000 Ci fission 99 Mo, in a 6-day reference, will cover 100% domestic demand of Korea, as well as 20% of international market. It is expected to replace 4.3 million USD ($800/Ci) of 99 Mo import for domestic market while exporting 82.8 million USD for world market, annually

Mo-99 production by fission and future projections

International Nuclear Information System (INIS)

Carranza, E.C.; Novello, A.; Bronca, M.; Cestau, D.; Bavaro, R.; Centurion, R.; Bravo, C.; Bronca, P.; Gualda, E.; Fraguas, F.; Giomi, A.; Ivaldi, L.

2012-01-01

Description of the I-131 and Mo-99 production process: The process starts with the irradiation of uranium-aluminum mini plates in the RA-3, Argentinean Reactor No.3, Ezeiza Atomic Center. In a nuclear reactor there is a constant flow of neutrons and when a neutron with proper energy impacts on a nucleus of U-235, it is absorbed at the same time generate an unstable configuration nuclear. For this reason, the nucleus formed is fission, getting two different atoms. Approximately 6% of the fissions produce Mo-99 and 3% produce I-131; the percentage remaining corresponds to formation of atoms without interest for use in medicine. In conclusion, the objective of the process developed in the Fission Plant, is starting from uranium mini plates, separate the Mo-99 and I-131 generated, the remaining elements formed. - Evolution of Mo-99 Production in the last 10 years: The Fission Mo-99 Plant Production begins routine production of Mo-99 in 1985, using targets made of uranium enriched at 90% U-235. In the 1990s, global concern regarding the use of highly enriched uranium, due to non-proliferation issues, caused the interruption of supply of nuclear material (HEU enriched at 90% of U-235). Following this, Argentina developed target based on low-enriched uranium (less than 20% U-235), becoming in 2002 the first country in the world to produce Mo-99 with LEU targets. From 2002 to date, the activity produced of Mo-99 has been tripled annually (author)

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

Electrocatalytic oxidation of methanol: study with Pt:Mo dispersed catalysts

Directory of Open Access Journals (Sweden)

Oliveira Neto Almir

2000-01-01

Full Text Available The electrocatalytic oxidation of methanol on Pt:Mo dispersed on carbon prepared using an alternative method recently developed in this laboratory was investigated. The EDX analysis confirmed that the simultaneous reduction of the precursor salts of Pt and Mo leads to the presence of these materials at the nominal composition initially calculated. The addition of Mo to Pt causes an increase of the oxidation currents, but does not improve the catalytic effect for methanol oxidation. Tafel plots for various methanol concentrations showed the presence of two slopes. On line differential electrochemical mass spectrometry (DEMS was used to investigate the distribution of products and intermediates in methanol oxidation.

76 FR 38357 - Reorganization of Foreign-Trade Zone 102, Under Alternative Site Framework; St. Louis, MO

Science.gov (United States)

2011-06-30

... Zone 102, Under Alternative Site Framework; St. Louis, MO Pursuant to its authority under the Foreign.../2010) as an option for the establishment or reorganization of general-purpose zones; Whereas, the St... the City of St. Louis and St. Louis County, Missouri, within and adjacent to the St. Louis Customs and...

Preparing for Molybdenum-99 Production In Malaysia [Country report: Malaysia

International Nuclear Information System (INIS)

Dahalan, Rehir; Masood, Zarina; Zulkifli, Mohd Hashim; Yusof, Mohd Abd Wahab

2015-01-01

The research reactor at Nuclear Malaysia, which has been in operation since June 1982, has a maximum flux of 1x10 13 n/cm 2 /s at its central position, has been utilized in production of neutron activated molybdenum-99 ( 99 Mo) and may be suitable for the new initiative for producing fission 99 Mo from low enriched uranium (LEU) targets if an upgrade involving its power and neutron flux were done. Currently, there is no fission 99 Mo production in place in Malaysia; however, there is an existing weekly 99 Mo/ 99m Tc generator production utilizing imported fission 99 Mo. Malaysia’s current demand for fission 99 Mo is relatively small but is still affected by the recent supply turmoil. At the request of the Malaysia Nuclear Agency, the IAEA organized a fact-finding mission to assess currently available infrastructure against that necessary to produce fission 99 Mo sufficient for domestic needs or additionally to contribute to regional fission 99 Mo supply security. During the mission, 99 Mo production from LEU and the alternative neutron activation method were considered. Taking into consideration sufficient upgrade of the current research reactor power and neutron flux, neutron activation could satisfy current national demand but offers little excess capacity to accommodate future growth or participation in the regional 99 Mo market. Also at a higher reactor power and neutron flux, LEU fission based technologies could produce adequate quantities for domestic and regional supply, but require significantly greater resource commitment than neutron activation production technologies particularly with respect to the management and ultimate disposition of all waste streams. In addition to the completion of the reactor power and flux upgrade, revising the operating mode to continuous operation is a prerequisite to fission 99 Mo production together with additional equipment for handling and transferring higher radiation dose target capsules from the reactor to the hot

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

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.

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)

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)

Association of the Leu72Met polymorphism of the ghrelin gene with the risk of Type 2 diabetes in subjects with impaired glucose tolerance in the Finnish Diabetes Prevention Study.

Science.gov (United States)

Mager, U; Lindi, V; Lindström, J; Eriksson, J G; Valle, T T; Hämäläinen, H; Ilanne-Parikka, P; Keinänen-Kiukaanniemi, S; Tuomilehto, J; Laakso, M; Pulkkinen, L; Uusitupa, M

2006-06-01

Ghrelin is a gut-brain regulatory peptide stimulating appetite and controlling energy balance. In previous studies, the Leu72Met polymorphism of the ghrelin gene has been associated with obesity and impaired insulin secretion. We investigated whether the Leu72Met polymorphism is associated with the incidence of Type 2 diabetes in subjects with impaired glucose tolerance (IGT) participating in the Finnish Diabetes Prevention Study (DPS). DPS was a longitudinal intervention study carried out in five participating centres in Finland. A total of 522 subjects with IGT were randomized into either an intervention or a control group and DNA was available from 507 subjects. The Leu72Met polymorphism was screened by the restriction fragment length polymorphism method. There were no differences in clinical and anthropometric characteristics among the genotypes at baseline. IGT subjects with the Met72 allele were at higher risk of developing Type 2 diabetes than subjects with the Leu72Leu genotype (P = 0.046). Our data also demonstrated that IGT subjects with the common Leu72Leu genotype developed Type 2 diabetes less frequently under intervention circumstances than subjects with the Met72 allele (OR = 0.28, 95% CI 0.10-0.79; P = 0.016). Subjects with the Leu72Leu genotype had a lower risk for the development of Type 2 diabetes. This was observed particularly in the study subjects who underwent an intensive diet and exercise intervention. Defective first-phase insulin secretion related to the Met72 allele might be one factor contributing to the conversion to Type 2 diabetes.

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.

ANL progress on the cooperation with CNEA for the MO-99 production: Base-side digestion process

International Nuclear Information System (INIS)

Gelis, A.V.; Quigley, K.J.; Aase, S.B.; Bakel, A.J.; Leyva, A.; Regalbuto, M.C.; Vandergrift, G.F.

2005-01-01

Conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) targets for the Mo-99 production requires certain modifications of the target design, the digestion and the purification processes. ANL and the Argentine Comision Nacional de Energia Atomica (CNEA) are collaborating to overcome all the concerns caused by the conversion of the CNEA process to use LEU foil targets. A new digester with stirring system has been successfully applied for the digestion of the low burn-up U foil targets in KMnO 4 alkaline media. In this paper, we report the progress on the development of the digestion procedure utilizing effective stirring and focusing on minimization of the liquid radioactive waste. (author)

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

Revaluation of 99Mo production by (n,γ) method at HANARO

International Nuclear Information System (INIS)

Jun, Byung Jin; Kimura, Akihiro; Hori, Naohiko; Izumo, Hironobu; Tsuchiya, Kunihiko; Lee, Byung Cheol

2010-07-01

After the feasibility study on 99 Mo production by (n,γ) method at HANARO was published by a KAERI report, worldwide supply of 99 Mo became worse and a need for early available alternative 99 Mo became stronger. Previous study indicated that the (n,γ) 99 Mo has a potential to be an alternative mass 99 Mo available earlier than those by any other methods. It can be realized when radioisotope industry of each country accepts the use of (n,γ) 99 Mo for a meaningful portion of national demand. A good backup supply system among high flux reactors in the region is a prerequisite to guarantee a stable and sufficient availability of the (n,γ) 99 Mo for the region, for which active collaboration among reactors is essential. As the initial stage of collaboration between HANARO and JMTR for the (n,γ) 99 Mo supply, the specific experience and 99 Mo production capability in HANARO have been discussed and revisited on the base of the previous report. (author)

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.

Safety analysis of an irradiation device for 99Mo production in RA-3 reactor

International Nuclear Information System (INIS)

Lerner, Ana Maria; Madariaga, Marcelo; Waldman, Ricardo

2000-01-01

The Argentine RA-3 research reactor (5 MW) has been converted to LEU fuel more than nine years ago. Since then, it has been operating with LEU fuel, which has been designed and fabricated at the National Atomic Energy Commission (CNEA). The Nuclear Regulatory Authority (ARN) is the institution in charge of the installation safety control. It is under this framework that the ARN has elaborated a neutronic calculation model for the RA-3 core, paying special attention to the device presently used for the irradiation of (HEU) 235 U targets required to obtain 9 '9Mo as a fission product. A regulatory analysis of results is carried out in the framework of ARN standards for fixed experiments. For such purpose, calculated reactivity values associated with such device are compared with recently measured values at the installation. Finally, and according to guidelines established in the first part of this work, a calculation model for a new device proposed by CNEA for the irradiation of metallic (LEU) uranium targets and still at its design stage, is here analysed. (author)

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)

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

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)

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.

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.

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).

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.

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.

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

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.

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.

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.

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...

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

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

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)

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

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)

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)

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...

Study on optimum conditions for Mo-99 adsorption by magnetite nanoparticles

International Nuclear Information System (INIS)

Holland, Helber; Yamaura, Mitiko; Damasceno, Marcos O.; Santos, Jacinete L.

2013-01-01

Radioisotopes play an important role in the peaceful uses of atomic energy. Technetium-99m is the most used radioisotope for diagnosis imaging in nuclear medicine and it is the decay product of Mo-99. One route to obtaining Mo-99 is in the form of fission product from Uranium targets irradiated in reactor. Uranium targets are dissolved by alkaline or acid process and the obtained solution is submitted to separation and purification steps of Mo-99 from the other fission products. Traditional separation techniques are inadequate for removing large volumes containing low concentrations metals due to the low operating efficiency and high costs processes. Therefore, alternative methods are being investigated as adsorption. Adsorption advantages over other techniques is low waste generation, easy metals recovery and reusability of adsorbents. Inorganic oxides are known for their ability to bind to metal ions in solution. At nanoscale range, this characteristic is highly potentialized. Thus, the use of nanoparticles has attracted attention for metal ions recovery by adsorption. Magnetite, Fe3O4, is an oxide formed by iron ions of valence 2+ and 3+. Due to the superparamagnetic behavior that arises in this material at nanoscale and crystal structure itself which favors surface adsorption, magnetite can be used as an adsorber agent to remove metal ions in solution. In this work, adsorption studies were performed to investigate best conditions for Mo-99 removal in solution. Influence of pH, stirring speed, contact time and initial concentration of Mo were studied. (author)

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

Fabrication Method of the Mo-99 Target with Advanced Planar Flow Casting

International Nuclear Information System (INIS)

Sim, M. S.; Lee, J. H.; Kim, C. K.; Kim, W. J.

2011-01-01

Mo-99 is a parent isotope of Tc-99m for medical diagnosis and very significant owing to its large fraction over 80% of the whole demand of medical radioisotopes in the all countries. Mo-99 isotope has been produced mainly by 235 U which is extracting fission products. All the major providers of fission Mo have used HEU as a target material. But RERTR program that is nonproliferation policy encourages using HEU to LEU. KAERI has developed a processing to be able to produce a uranium foil continuously at one go. This processing gave an opportunity for LEU target using uranium foil to be commercialized. It correspond RERTR program. KAERI developed a new process of making foil directly from uranium melt by PFC. This process is simple, productive, and cost-effective. But the foil's air-side surface is generally very rough. A typical transverse cross section had a minimum thickness of 65 μm and a maximum thickness of 205 μm. This roughness could affect target fabrication and irradiation behavior. After issuing this problem KAERI launched a further effort since 2008. A new equipment was designed and manufactured in the industry in 2009. While the new equipment being test-operating, some occurrence of appearing problems appeared. Since 2010, Equipment was moved to KAERI, we performed many experiments using depleted uranium, and go get satisfied some results. We have got interesting results and manufactured uranium foil. A typical transverse cross section had a minimum thickness of 87 μm and a maximum thickness of 194 μm. The average thickness is 120 μm as a result of calculation

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.

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.

Current status of research and development of 99Mo/99mTc generator in Korea

International Nuclear Information System (INIS)

Park, U.J.; Lee, J.S.; Son, K.J.; Nam, S.S.; Kwak, S.I.; Han, H.S.

2006-01-01

To supply 99m Tc in stable and economical manner in Korea, a chromatographic generator has been under development at KAERI since late 1980's. The chromatographic type of technetium generator is preferred in hospital because it is more convenient and less time-restricted for applications. Hence, the demand of 99m Tc in medical applications is keep increasing. In Korea, there are more than 200 gamma cameras including SPECT in hospitals. For these applications, approximately 100 units/week of 99m Tc generators of which annual cost reached 3 million US dollars were required in 2002. Hence, the development of 99m Tc generators and technology of fission 99 Mo processing were started while installing the generator loading facility (GLF) at KAERI. This facility is currently on a trial run for the commercial production of 99m Tc generators and expected to produce more than 200 generators per week in 2004. For the fission 99 Mo production, an annulus U foil target was considered as the LEU target. The designed LEU target is being used for target manufacturing test and accident analysis. The develop gel type 99m Tc generators by using a poly zirconium complex (PZC) and alumina column, KAERI has cooperated with a Japanese company since late 1990's. In 2003, experimental studies for molybdate adsorption capacity and elution characteristics of the PZC samples from three different batches produced by Kaken Co. were carried out. (author)

An Effort to Improve U Foil Fabrication Technology of Roll-casting for Fission Mo Target

Energy Technology Data Exchange (ETDEWEB)

Kim, Chang Kyu; Woo, Yun Myeong; Kim, Ki Hwan; Oh, Jong Myeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Sim, Moon Soo [Chungnam University, Green Energy Technology, Daejeon (Korea, Republic of)

2010-10-15

Mo-99 isotope has been produced mainly by extracting fission products of {sup 235}U. The targets for irradiating in reactor have used as stainless tube coated with highly enriched UO{sub 2} at the inside surface and highly enriched UAlx plate cladded with aluminum. In connection with non-proliferation policy the RERTR program developed a new process of Mo-99 using low enriched uranium (LEU) instead of highly enriched uranium (HEU). LEU should be put about five times more quantity than HEU because the {sup 235}U contents of LEU and HEU are 20% and higher than 90%, respectively. Accordingly pure uranium metal foil target was adopted as a promising target material due to high uranium density. ANL and BATAN developed a Cintichem process using uranium metal foil target of 130 {mu}m in thickness jointly and the RERTR program is trying to disseminate the new process world-widely. However, uranium foil is made by lots of times rolling work on uranium plate, which is laborious and tedious. In order to avoid this difficulty KAERI developed a new process of making foil directly from uranium melt by roll casting. This process is very much simple, productive, and cost-effective. But the outside surface of foil is generally very rough. A typical transverse cross section had a minimum thickness of 65 {mu}m and a maximum thickness of 205 {mu}m. This roughness could affect (1) target fabrication, where the U foil, or the Ni foil might be damaged during drawing, and (2) irradiation behavior, where gaps between the target walls and the U metal might affect cooling of the target

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.

ITO-free flexible organic photovoltaics with multilayer MoO3/LiF/MoO3/Ag/MoO3 as the transparent electrode

International Nuclear Information System (INIS)

Chen, Shilin; Dai, Yunjie; Zhang, Hongmei; Zhao, Dewei

2016-01-01

We present efficient flexible organic photovoltaics (OPVs) with multiple layers of molybdenum oxide (MoO 3 )/LiF/MoO 3 /Ag/MoO 3 as the transparent electrode, where the thin Ag layer yields high conductivity and the dielectric layer MoO 3 /LiF/MoO 3 has high transparency due to optical interference, leading to improved power conversion efficiency compared with indium tin oxide (ITO) based devices. The MoO 3 contacting organic active layer is used as a buffer layer for good hole extraction. Thus, the multilayer MoO 3 /LiF/MoO 3 /Ag/MoO 3 can improve light transmittance and also facilitate charge carrier extraction. Such an electrode shows excellent mechanical bendability with a 9% reduction of efficiency after 1000 cycles of bending due to the ductile nature of the thin metal layer and dielectric layer used. Our results suggest that the MoO 3 /LiF/MoO 3 /Ag/MoO 3 multilayer electrode is a promising alternative to ITO as an electrode in OPVs. (paper)

Quality control studies of 99Mo used in 99Mo/99mTc generators produced at IPEN/CNEN-SP, Brazil

International Nuclear Information System (INIS)

Said, Daphne S.; Brambilla, Tania P.; Matsuda, Margareth M.N.; Osso Junior, Joao A.

2015-01-01

99m Tc is the most used radionuclide in nuclear medicine. In Brazil, the 99 Mo/ 99m Tc generators are produced exclusively by the Center of Radiopharmacy at IPEN-CNEN/SP, by importing 99 Mo from different suppliers. 99 Mo (t 1/2 = 66 h) is a fission product of 235 U, therefore, it can be accompanied by several radioisotopes that are highly prejudicial for human health, demanding a strict quality control of this product for generators safe use. The European Pharmacopoeia established some parameters and limits that evaluate the quality of the solution of sodium [ 99 Mo]molybdate, that is used as raw material for generator's production. The European Pharmacopoeia also recommends some analytical methods to perform these evaluations, however, it has been observed difficulties on the implementation of these methods by the generator's producers. These difficulties are probably related to the lack of practicability of the proposed methods and the extensive list of utilized reagents. In this work some procedures of the European Pharmacopoeia's quality control method for 99 Mo were evaluated. Different types of solid phase exchanger cartridges were tested for retention of 99 Mo in 3 different conditions. Cartridges that presented percentages of retention higher than 90% were also tested for separation of 99 Mo from possible contaminants (Ru e Te). The results shown that solid phase exchanger cartridges that presented percentages of retention of Mo higher than 90% also presented significant percentages of retention of Ru and Te. An alternative method for separation of 99 Mo from 131 I (other contaminant) are also proposed. (author)

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...

Electroplating fission-recoil barriers onto LEU-metal foils for 99Mo-production targets

International Nuclear Information System (INIS)

Smaga, J.A.; Sedlet, J.; Conner, C.; Liberatore, M.W.; Walker, D.E.; Wygmans, D.G.; Vandegrift, G.F.

1997-01-01

Electroplating experiments on uranium foil have been conducted in order to develop low-enriched uranium composite targets suitable for the production of 99 Mo. Preparation of the foil surface prior to plating was found to play a key role in the quality of the resultant coating. A surface preparation procedure was developed that produces both zinc and nickel coatings with the desired level of coating adherence and coverage. Modifications of the existing plating processes now need investigation to improve to uniformity of the plating thickness, especially at the foil perimeter. (author)

Electroplating fission-recoil barriers onto LEU-metal foils for 99Mo-production targets

International Nuclear Information System (INIS)

Smaga, J.A.; Sedlet, J.; Conner, C.; Liberatore, M.W.; Walker, D.E.; Wygmans, D.G.; Vandegrift, G.F.

1997-10-01

Electroplating experiments on uranium foil have been conducted in order to develop low-enriched uranium composite targets suitable for the production of 99 Mo. Preparation of the foil surface prior to plating was found to play a key role in the quality of the resultant coating. A surface preparation procedure was developed that produces both zinc and nickel coatings with the desired level of coating adherence and coverage. Modifications of the existing plating processes now need investigation to improve to uniformity of the plating thickness, especially at the foil perimeter

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

Ternary alloying study of MoSi2

International Nuclear Information System (INIS)

Yi, D.; Li, C.; Akselsen, O.M.; Ulvensoen, J.H.

1998-01-01

Ternary alloying of MoSi 2 with adding a series of transition elements was investigated by X-ray diffraction (XRD), scanning electron microscopy, transmission electron microscopy (TEM), and energy dispersive spectroscopy (EDS). Iron, Co, Ni, Cr, V, Ti, and Nb were chosen as alloying elements according to the AB 2 structure map or the atomic size factor. The studied MoSi 2 base alloys were prepared by the arc melting process from high-purity metals. The EDS analysis showed that Fe, Co, and Ni have no solid solubility in as-cast MoSi 2 , while Cr, V, Ti, and Nb exhibit limited solid solubilities, which were determined to be 1.4 ± 0.7, 1.4 ± 0.4, 0.4 ± 0.1, and 0.8 ± 0.1. Microstructural characterization indicated that Mo-Si-M VIII (M VIII = Fe, Co, Ni) and Mo-Si-Cr alloys have a two-phase as-cast microstructure, i.e., MoSi 2 matrix and the second-phase FeSi 2 , CoSi, NiSi 2 , and CrSi 2 , respectively. In as-cast Mo-Si-V, Mo-Si-Ti, and Mo-Si-Nb alloys, besides MoSi 2 and C40 phases, the third phases were observed, which have been identified to be (Mo, V) 5 Si 3 , TiSi 2 , and (Mo, Nb) 5 Si 3

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)

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.)

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.

Reduced interaction layer growth of U–Mo dispersion in Al–Si

International Nuclear Information System (INIS)

Kim, Yeon Soo; Park, Jong Man; Ryu, Ho Jin; Jung, Yang Hong; Hofman, G.L.

2012-01-01

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 ∼50% burnup of initial 19.75% U-235 enrichment at ∼200 °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.

Development of uranium metal targets for 99Mo production

International Nuclear Information System (INIS)

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

1993-10-01

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

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.)

Fabrication Method of the Mo-99 Target with Advanced Planar Flow Casting

Energy Technology Data Exchange (ETDEWEB)

Sim, M. S.; Lee, J. H. [Chungnam University, Green Energy Technology, Daejeon (Korea, Republic of); Kim, C. K.; Woo, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-05-15

Mo-99 is a parent isotope of Tc-99m for medical diagnosis and very significant owing to its large fraction over 80% of the whole demand of medical radioisotopes in the all countries. Mo-99 isotope has been produced mainly by {sup 235}U which is extracting fission products. All the major providers of fission Mo have used HEU as a target material. But RERTR program that is non-proliferation policy encourages using HEU to LEU. KAERI has developed a processing to be able to produce a uranium foil continuously at one go. This processing gave an opportunity for LEU target using uranium foil to be commercialized. It correspond RERTR program. KAERI developed a new process of making foil directly from uranium melt by PFC. This process is simple, productive, and cost-effective. But the foil{center_dot}{center_dot}{center_dot}s air-side surface is generally very rough. A typical transverse cross section had a minimum thickness of 65 {mu}m and a maximum thickness of 205 {mu}m. This roughness could affect target fabrication and irradiation behavior. After issuing this problem KAERI launched a further effort since 2008. A new equipment was designed and manufactured in the industry in 2009. While the new equipment being test-operating, some occurrence of appearing problems appeared. Since 2010, Equipment was moved to KAERI, we performed many experiments using depleted uranium, and go get satisfied some results. We have got interesting results and manufactured uranium foil. A typical transverse cross section had a minimum thickness of 87 {mu}m and a maximum thickness of 194 {mu}m. However, the average thickness is 130 {mu}m as a result of measurement by a micrometer

EXAFS study of Mo2N and Mo nitrides supported on zeolites

International Nuclear Information System (INIS)

Liu Zhenlin; Meng Ming; Fu Yilu; Jiang Ming; Hu Tiandou; Xie Yaning; Liu Tao

2002-01-01

In the present study, the reaction is applied to prepare molybdenum nitrides with high surface area, and zeolites are used as supports. The EXAFS of the Mo K-absorption edge is measured and the change of coordination environment of Mo atoms before and after the nitridation is revealed

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

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.

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)

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

Comparison of Alternate and Original Items on the Montreal Cognitive Assessment.

Science.gov (United States)

Lebedeva, Elena; Huang, Mei; Koski, Lisa

2016-03-01

The Montreal Cognitive Assessment (MoCA) is a screening tool for mild cognitive impairment (MCI) in elderly individuals. We hypothesized that measurement error when using the new alternate MoCA versions to monitor change over time could be related to the use of items that are not of comparable difficulty to their corresponding originals of similar content. The objective of this study was to compare the difficulty of the alternate MoCA items to the original ones. Five selected items from alternate versions of the MoCA were included with items from the original MoCA administered adaptively to geriatric outpatients (N = 78). Rasch analysis was used to estimate the difficulty level of the items. None of the five items from the alternate versions matched the difficulty level of their corresponding original items. This study demonstrates the potential benefits of a Rasch analysis-based approach for selecting items during the process of development of parallel forms. The results suggest that better match of the items from different MoCA forms by their difficulty would result in higher sensitivity to changes in cognitive function over time.

Thermal expansion studies on Th(MoO4)2, Na2Th(MoO4)3 and Na4Th(MoO4)4

International Nuclear Information System (INIS)

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

2008-01-01

Thermal expansion behavior of Th(MoO 4 ) 2 , Na 2 Th(MoO 4 ) 3 and Na 4 Th(MoO 4 ) 4 was studied under vacuum in the temperature range of 298-1123 K by high temperature X-ray diffractometer. Th(MoO 4 ) 2 was synthesized by reacting ThO 2 with 2 mol of MoO 3 , at 1073 K in air and Na 2 Th(MoO 4 ) 3 and Na 4 Th(MoO 4 ) 4 were prepared by reacting Th(MoO 4 ) 2 with 1 and 2 mol of Na 2 MoO 4 , respectively at 873 K in air. The XRD data of Th(MoO 4 ) 2 was indexed on orthorhombic system where as XRD data of Na 2 Th(MoO 4 ) 3 and Na 4 Th(MoO 4 ) 4 were indexed on tetragonal system. The lattice parameters and cell volume of all the three compounds, fit into polynomial expression with respect to temperature, showed positive thermal expansion (PTE) up to 1123 K. The average value of thermal expansion coefficients for Th(MoO 4 ) 2 , Na 2 Th(MoO 4 ) 3 and Na 4 Th(MoO 4 ) 4 were determined from the high temperature data

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).

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

Electron microscopy studies on MoS2 nanocrystals

DEFF Research Database (Denmark)

Hansen, Lars Pilsgaard

Industrial-style MoS2-based hydrotreating catalysts are studied using electron microscopy. The MoS2 nanostructures are imaged with single-atom sensitivity to reveal the catalytically important edge structures. Furthermore, the in-situ formation of MoS2 crystals is imaged for the first time....

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)

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...

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)

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

Amorphous-crystalline transition studied in hydrated MoO3

International Nuclear Information System (INIS)

Camacho-Lopez, M.A.; Haro-Poniatowski, E.; Lartundo-Rojas, L.; Livage, J.; Julien, C.M.

2006-01-01

In this work we study the thermal behavior of hydrated MoO 3 synthesized via acidification of sodium molybdate. MoO 3 .nH 2 O (n = 1.4) amorphous compound was heated in air at increasing temperatures in order to obtain the crystalline MoO 3 phase. We have studied the structural changes as a function of annealing temperature by Raman spectroscopy. A statistical study to determine the average size of the crystallites at each annealing step has been realized by scanning electron microscopy. Results show that the hydrated MoO 3 .1.4H 2 O glass transforms in an amorphous MoO 3 .0.7H 2 O phase prior to its crystallization, while the sample heated at 500 deg. C crystallizes into the orthorhombic α-MoO 3 phase with micro-crystallites having an average size of 6.8 μm

Development of 99Mo/99mTc Generator System for Production of Medical Radionuclide 99mTc using a Neutron-activated 99Mo and Zirconium Based Material (ZBM as its Adsorbent

Directory of Open Access Journals (Sweden)

I. Saptiama

2016-12-01

Full Text Available Molybdenum produced from fission of U-235 is the most desirable precursor for 99Mo/99mTc generator system as it is non-carrier added and has high specific activity. However, in the last decade there has been short supply of 99Mo due to several constrains. Therefore, there have been many works performed for development of 99Mo/99mTc generator system using 99Mo which is not produced from either LEU or HEU. This report deals with development of 99Mo/99mTc generator system where zirconium-based material (ZBM is used as adsorbent of neutron-activated 99Mo. The system was prepared by firstly irradiating natural Mo in the G. A. Siwabessy reactor to produce neutron-activated 99Mo. The target was dissolved in NaOH 4N and then neutralized with 12 M HCl. The 99Mo solution was then mixed with a certain amount of ZBM followed by heating at 90°C for three hours to allow the 99Mo adsorbed on ZBM. The 99Mo-ZBM (9.36 GBq of 99Mo was Mo/ 4.2 g ZBM was packed on a fritz-glass column. This column was then fitted serially with an alumina column for trapping 99Mo breakthrough. The columns were then eluted daily with saline solution for up to one week. The yield of 99mTc was found to be between 53.7 – 74% (n= 5. All 99mTc eluates were clear solutions with pH of 5. Breakthrough of 99Mo in 99mTc eluates was found to be 0.031 ± 0.019 μCi 99Mo/ mCi 99mTc (n= 5 which was less than the maximum activity of 99Mo allowed in 99mTc solution ( 99%. Radiolabeling of this 99mTc towards methylene diphosphonate (MDP kit gave a radiolabelling efficiency of 99%. In summary, a new 99Mo/99mTc generator system that used neutron-activated 99Mo and ZBM as its adsorbent has been successfully prepared. The 99mTc produced from this new 99Mo/99mTc generator system attained the quality of 99mTc required for medical purposes.

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

Comparitive study of fluorescence lifetime quenching of rhodamine 6G by MoS2 and Au-MoS2

Science.gov (United States)

Shakya, Jyoti; Kasana, Parath; Mohanty, T.

2018-04-01

Time resolved fluorescence study of Rhodamine 6G (R6G) in the presence of Molybdenum disulfide (MoS2) nanosheets and gold doped MoS2 (Au-MoS2) have been carried out and discussed. We have analyzed the fluorescence decay curves of R6G and it is observed that Au-MoS2 is a better fluorescence lifetime quencher as compare to MoS2 nanosheets. Also, the energy transfer efficiency and energy transfer rate from R6G to MoS2 and Au-MoS2 has been calculated and found higher for Au-MoS2.

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.

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

2011 Progress Report on HEU Minimization Activities in Argentina

Energy Technology Data Exchange (ETDEWEB)

Bonini, A.; Cristini, P.; Lio, L. De; Dell' Occhio, L.; Gil, D.; Gonzalez, A.G.; Gonzalez, R.; Varela, C. Komar; Lopez, M.; Novara, O.; Taboada, H. [Comision Nacional de Energia Atomica, Av. Del Libertador 8250 (1429) Buenos Aires (Argentina)

2011-07-01

After the core conversion of the RA-6 reactor finished in March 2008, an extension of the original CNEA-NNSA DoE contract was signed to enhance the final national HEU inventories minimization. Before this process, CNEA reserved a small inventory of HEU for R and D uses in fission chambers, neutronic probes and standards. This minimization comprises that all fresh and irradiated HEU remnant inventories coming from fuels and Mo99 irradiation targets fabrication and irradiated HEU-oxides retained in production filters and solutions will be recovered, down-blended into LEU and purified or dispose as waste whenever its recovery would not be advisable due to cost-benefit consideration. CNEA has a R and D program to develop the fabrication technology of both dispersed U-Mo (Al-Si matrix and Al cladding) and monolithic (Zry-4 cladding) miniplates to support the qualification activities of the RERTR program. Some monolithic 58% enrichment and LEU 8%Mo and U10%Mo miniplates and plates were and are being delivered to INL-DoE to be irradiated in the ATR reactor core. CNEA, a worldwide leader on LEU technology for fission radioisotope production is providing Brazil with 1/3 of the national requirements on Mo99 by weekly deliveries. Australia has started the fission radioisotope production through several batches by week, based on CNEA's LEU technology provided by 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. Plans to recover and purify the LEU based inventories in Mo99 production filters, once the HEU to LEU campaign is over. 2. Fabrication and delivering to INL to be irradiated in the ATR core of U-8%Mo and U-10%Mo monolithic miniplates and development and fabrication of LEU very high density monolithic and dispersed U-Mo fuel plates with Zr cladding for the FUTURE-MONO experiment in the frame of the RERTR program. 3

Quality control studies of {sup 99}Mo used in {sup 99}Mo/{sup 99m}Tc generators produced at IPEN/CNEN-SP, Brazil

Energy Technology Data Exchange (ETDEWEB)

Said, Daphne S.; Brambilla, Tania P.; Matsuda, Margareth M.N.; Osso Junior, Joao A., E-mail: daphnesaid@usp.br, E-mail: taniabrambilla@yahoo.com.br, E-mail: mmatsuda@ipen.br, E-mail: jaosso@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

{sup 99m}Tc is the most used radionuclide in nuclear medicine. In Brazil, the {sup 99}Mo/{sup 99m}Tc generators are produced exclusively by the Center of Radiopharmacy at IPEN-CNEN/SP, by importing {sup 99}Mo from different suppliers. {sup 99}Mo (t{sub 1/2} = 66 h) is a fission product of {sup 235}U, therefore, it can be accompanied by several radioisotopes that are highly prejudicial for human health, demanding a strict quality control of this product for generators safe use. The European Pharmacopoeia established some parameters and limits that evaluate the quality of the solution of sodium [{sup 99}Mo]molybdate, that is used as raw material for generator's production. The European Pharmacopoeia also recommends some analytical methods to perform these evaluations, however, it has been observed difficulties on the implementation of these methods by the generator's producers. These difficulties are probably related to the lack of practicability of the proposed methods and the extensive list of utilized reagents. In this work some procedures of the European Pharmacopoeia's quality control method for {sup 99}Mo were evaluated. Different types of solid phase exchanger cartridges were tested for retention of {sup 99}Mo in 3 different conditions. Cartridges that presented percentages of retention higher than 90% were also tested for separation of {sup 99}Mo from possible contaminants (Ru e Te). The results shown that solid phase exchanger cartridges that presented percentages of retention of Mo higher than 90% also presented significant percentages of retention of Ru and Te. An alternative method for separation of {sup 99}Mo from {sup 131}I (other contaminant) are also proposed. (author)

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)

IEA Mobility Model (MoMo) and its use in the ETP 2008

International Nuclear Information System (INIS)

Fulton, Lew; Cazzola, Pierpaolo; Cuenot, Francois

2009-01-01

The IEA published 'Energy Technology Perspectives' (ETP) in June 2008. That document reports on IEA scenarios for baseline and low-CO 2 alternative scenarios to 2050, across the energy economy. The study included creating scenarios for transport, using the IEA Mobility Model (MoMo). This paper reports on the transport-related ETP scenarios and describes the model used in the analysis. According to the ETP Baseline scenario, world transport energy use and CO 2 emissions will more than double by 2050. In the most challenging scenario, called 'BLUE', transport emissions are reduced by 70% in 2050 compared to their baseline level in that year (and about 25% below their 2005 levels). There are several versions of the BLUE scenario, but all involve: a 50% or greater improvement in LDV efficiency, 30-50% improvement in efficiency of other modes (e.g. trucks, ships and aircraft), 25% substitution of liquid fossil fuels by biofuels, and considerable penetration of electric and/or fuel-cell vehicles. In the second half of this paper, an overview of the MoMo model is provided. Details on the complete analysis are contained in the ETP 2008 document, available at (www.iea.org). Details of the LDV fuel economy analysis are contained in a separate paper in this collection.

DFT study of the reactions of Mo and Mo with CO2 in gas phase

Indian Academy of Sciences (India)

understanding the mechanism of second-row metal reacting with CO2. The minimum energy ... et al.18 performed an IR study on the reaction of laser- ablated Mo atom .... indicate that the weak electrostatic interaction between. Mo. + and CO2 ...

The recovery of 99Mo from solutions of irradiated Uranium using a column with nanoparticles of Titanium Dioxide

International Nuclear Information System (INIS)

Androne, G. E.; Petre, M.; Lazar, C. G.

2016-01-01

Molyibdenum-99 (T½ = 66.02 h) decays by beta emission to 99 Tcm (T½ = 6.02 h). The latter nuclide is used in many nuclear medicine applications. The 99 Mo is produced from irradiated high (HEU) or low (LEU) enriched uranium. In this work a sensitive and selective method for recovering Mo from uranium solution, using a column with titanium dioxide nanoparticles, is developed. The titanium dioxide (TiO 2 ) nanoparticles were synthesized via sol-gel method using titanium tetra-chloride as starting material and urea as a reacting medium. A 40 ml uranium solution containing 450 g/L uranyl nitrate, 1 M HNO 3 , and 4 mg Mo was loaded on a column containing 6 g of TiO 2 sorbent at 75°C. After loading, the column was washed with 1 M HNO 3 and H 2 O. Mo was stripped from the column with 0.1 M NaOH at 25°C. The ICP-MS results indicate that 80-95% of the initial mass of Mo was loaded on the column, and 90-94% of this quantity was recovered in the strip fraction. (authors)

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.

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)

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.)

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)

Diffusion barrier performances of thin Mo, Mo-N and Mo/Mo-N films between Cu and Si

International Nuclear Information System (INIS)

Song Shuangxi; Liu Yuzhang; Mao Dali; Ling Huiqin; Li Ming

2005-01-01

In this work, we have studied the diffusion barrier performances of Mo, Mo-N and Mo/Mo-N metallization layers deposited by sputtering Mo in Ar/N 2 atmospheres, respectively. Samples were subsequently annealed at different temperatures ranging from 400 to 800 deg C in vacuum condition. The film properties and their suitability as diffusion barriers and protective coatings in silicon devices were characterized using four-point probe measurement, X-ray diffractometry, scanning electron microscopy, Auger electron spectroscopy and transmission electron microscopy analyses. Experimental results revealed that the Mo (20 nm)/Mo-N (30 nm) layer was able to prevent the diffusion reaction between Cu and Si substrate after being annealed at 600 deg C for 30 min. The adhesion between layers and the content of N atoms are the key parameters to improve the properties of Mo-based barrier materials. The Mo layer interposed between Cu and Mo-N diluted the high nitrogen concentration of the barrier and so enhanced the barrier performances

Benchmark experiment for the cross section of the 100Mo(p,2n)99mTc and 100Mo(p,pn)99Mo reactions

Science.gov (United States)

Takács, S.; Ditrói, F.; Aikawa, M.; Haba, H.; Otuka, N.

2016-05-01

As nuclear medicine community has shown an increasing interest in accelerator produced 99mTc radionuclide, the possible alternative direct production routes for producing 99mTc were investigated intensively. One of these accelerator production routes is based on the 100Mo(p,2n)99mTc reaction. The cross section of this nuclear reaction was studied by several laboratories earlier but the available data-sets are not in good agreement. For large scale accelerator production of 99mTc based on the 100Mo(p,2n)99mTc reaction, a well-defined excitation function is required to optimise the production process effectively. One of our recent publications pointed out that most of the available experimental excitation functions for the 100Mo(p,2n)99mTc reaction have the same general shape while their amplitudes are different. To confirm the proper amplitude of the excitation function, results of three independent experiments were presented (Takács et al., 2015). In this work we present results of a thick target count rate measurement of the Eγ = 140.5 keV gamma-line from molybdenum irradiated by Ep = 17.9 MeV proton beam, as an integral benchmark experiment, to prove the cross section data reported for the 100Mo(p,2n)99mTc and 100Mo(p,pn)99Mo reactions in Takács et al. (2015).

Molybdenum adsorption by alumina and Dowex 1x8 resin for the separation and purification process of fission 99Mo

International Nuclear Information System (INIS)

Yamaura, M.; Damasceno, M.O.; Freitas, A.A.; Camilo, R.L.; Araujo, I.C.; Forbicini, C.A.L.G. de O.

2011-01-01

Molybdenum-99 is the most widely employed radioisotope in nuclear medicine, due to its decay product, Technetium-99m, a radioisotope used in over 80% of diagnostic tests. Since 2009, the production of generators 99 Mo/ 99m Tc suffers a crisis of global supply. The raw material, the 99 Mo, is produced mainly by fission of 235 U in the reactor in uranium targets. Brazilian government invests in building of a research reactor suitable for the domestic production of 99 Mo from LEU (Low Enriched Uranium) targets and the IPEN/CNEN develops the production technology. This work is part of the research for the development of production technology of 99 Mo at the IPEN/CNEN-SP. The study has evaluated the adsorption behaviour of molybdenum from the alkaline dissolution of aluminum plates by the alumina and by the anionic resin Dowex 1x8 aiming at their use in the process of separation and purification in chromatography columns. Influences of pH and of aluminum concentration in the retention of molybdenum were investigated. Results showed high performance in the wide pH range. However in strongly acid solutions containing aluminum, alumina showed higher adsorption percentage than that achieved by the resin Dowex 1x8. (author)

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)

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)

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.

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.

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)

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.

Design of ET(B) receptor agonists: NMR spectroscopic and conformational studies of ET7-21[Leu7, Aib11, Cys(Acm)15].

Science.gov (United States)

Hewage, Chandralal M; Jiang, Lu; Parkinson, John A; Ramage, Robert; Sadler, Ian H

2002-03-01

In a previous report we have shown that the endothelin-B receptor-selective linear endothelin peptide, ET-1[Cys (Acm)1,15, Ala3, Leu7, Aib11], folds into an alpha-helical conformation in a methanol-d3/water co-solvent [Hewage et al. (1998) FEBS Lett., 425, 234-238]. To study the requirements for the structure-activity relationships, truncated analogues of this peptide were subjected to further studies. Here we report the solution conformation of ET7-21[Leu7, Aib11, Cys(Acm)15], in a methanol-d3/water co-solvent at pH 3.6, by NMR spectroscopic and molecular modelling studies. Further truncation of this short peptide results in it displaying poor agonist activity. The modelled structure shows that the peptide folds into an alpha-helical conformation between residues Lys9-His16, whereas the C-terminus prefers no fixed conformation. This truncated linear endothelin analogue is pivotal for designing endothelin-B receptor agonists.

Thermochemical study of MoS2 oxidation

International Nuclear Information System (INIS)

Filimonov, D.S.; Topor, N.D.; Kesler, Ya.A.

1990-01-01

Thermochemical studies of oxidation processes of metallic molybdenum, sulfur, molybdenum disulfide under different conditions in microcalorimeter are conducted. Values of thermal effects which are used to calculate standard formation enthalpy of MoS 2 and which correlate well are obtained. Δ f H 0 (MoS 2 ,298.15 K) recommended value constitutes (-223.0±16.7) kJ/mol

Full size U-10Mo monolithic fuel foil and fuel plate fabrication-technology development

International Nuclear Information System (INIS)

Moore, G.A.; Jue, J-F.; Rabin, B.H.; Nilles, M.J.

2010-01-01

Full-size U-10Mo foils are being developed for use in high density LEU monolithic fuel plates. The application of a zirconium barrier layer to the foil is performed using a hot co-rolling process. Aluminium clad fuel plates are fabricated using Hot Isostatic Pressing (HIP) or a Friction Bonding (FB) process. An overview is provided of ongoing technology development activities, including: the co-rolling process, foil shearing/slitting and polishing, cladding bonding processes, plate forming, plate-assembly swaging, and fuel plate characterization. Characterization techniques being employed include, Ultrasonic Testing (UT), radiography, and microscopy. (author)

Amorphous-crystalline transition studied in hydrated MoO{sub 3}

Energy Technology Data Exchange (ETDEWEB)

Camacho-Lopez, M.A. [Facultad de Quimica, Universidad Autonoma del Estado de Mexico, Paseo Colon y Tollocan, Toluca Edo. de Mexico 50110 (Mexico); Haro-Poniatowski, E. [Departamento de Fisica, Laboratorio de Optica Cuantica, Universidad Autonoma Metropolitana-Iztapalapa, Apdo. Postal 55-534, Mexico, D.F. 09340 (Mexico); Lartundo-Rojas, L. [Laboratorio de Microscopia, Universidad Autonoma Metropolitana-Iztapalapa, Apdo. Postal 55-534, Mexico, D.F. 09340 (Mexico); Livage, J. [Chimie de la Matiere Condensee, Universite Pierre et Marie Curie, 4 Place Jussieu, 75252 Paris Cedex 05 (France); Julien, C.M. [Institut des Nano-Sciences de Paris, UMR 7588, Universite Pierre et Marie Curie, Campus Boucicaut, 140 rue de Lourmel, 75015 Paris (France)]. E-mail: Christian.Julien@insp.jussieu.fr

2006-11-25

In this work we study the thermal behavior of hydrated MoO{sub 3} synthesized via acidification of sodium molybdate. MoO{sub 3}.nH{sub 2}O (n = 1.4) amorphous compound was heated in air at increasing temperatures in order to obtain the crystalline MoO{sub 3} phase. We have studied the structural changes as a function of annealing temperature by Raman spectroscopy. A statistical study to determine the average size of the crystallites at each annealing step has been realized by scanning electron microscopy. Results show that the hydrated MoO{sub 3}.1.4H{sub 2}O glass transforms in an amorphous MoO{sub 3}.0.7H{sub 2}O phase prior to its crystallization, while the sample heated at 500 deg. C crystallizes into the orthorhombic {alpha}-MoO{sub 3} phase with micro-crystallites having an average size of 6.8 {mu}m.

Molybdenum adsorption by alumina and Dowex 1x8 resin for the separation and purification process of fission {sup 99}Mo

Energy Technology Data Exchange (ETDEWEB)

Yamaura, M.; Damasceno, M.O.; Freitas, A.A.; Camilo, R.L.; Araujo, I.C.; Forbicini, C.A.L.G. de O., E-mail: myamaura@ipen.b, E-mail: molidam@ipen.b, E-mail: afreitas@ipen.b, E-mail: rcamilo@ipen.b, E-mail: cruz.araujo@uol.com.b, E-mail: cforbici@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

Molybdenum-99 is the most widely employed radioisotope in nuclear medicine, due to its decay product, Technetium-99m, a radioisotope used in over 80% of diagnostic tests. Since 2009, the production of generators {sup 99}Mo/{sup 99m}Tc suffers a crisis of global supply. The raw material, the {sup 99}Mo, is produced mainly by fission of {sup 235}U in the reactor in uranium targets. Brazilian government invests in building of a research reactor suitable for the domestic production of {sup 99}Mo from LEU (Low Enriched Uranium) targets and the IPEN/CNEN develops the production technology. This work is part of the research for the development of production technology of {sup 99}Mo at the IPEN/CNEN-SP. The study has evaluated the adsorption behaviour of molybdenum from the alkaline dissolution of aluminum plates by the alumina and by the anionic resin Dowex 1x8 aiming at their use in the process of separation and purification in chromatography columns. Influences of pH and of aluminum concentration in the retention of molybdenum were investigated. Results showed high performance in the wide pH range. However in strongly acid solutions containing aluminum, alumina showed higher adsorption percentage than that achieved by the resin Dowex 1x8. (author)

Association between the ghrelin Leu72Met polymorphism and type 2 diabetes risk: a meta-analysis.

Science.gov (United States)

Liao, Ning; Xie, Zi-Kang; Huang, Jian; Xie, Zheng-Fu

2013-04-01

Data on the association between the ghrelin Leu72Met polymorphism and type 2 diabetes are conflicting. A meta-analysis was performed on this topic. We searched for case-control studies using electronic databases (Medline and PubMed) and reference lists of studies. Odds ratios (OR) and 95% confidence intervals (CI) assuming dominant, recessive and homozygote comparison genetic models were calculated. Six case-control studies involving a total of 3417 cases and 3081 controls were included in this meta-analysis. No association was found between the ghrelin Leu72Met polymorphism and type 2 diabetes risk in the overall population in dominant, recessive and homozygote comparison models. However, in subgroup analyses stratified by ethnicity, we found that the risk for type 2 diabetes was decreased in subjects with Met72+ genotypes in Caucasians (OR=0.79, 95% CI: 0.64-0.98, P(z)=0.030). The ghrelin Leu72Met polymorphism was protective against type 2 diabetes in Caucasians. Future studies performed in larger sample size are needed to allow a more definitive conclusion. Copyright © 2012 Elsevier B.V. All rights reserved.

Spectrophotometric and potentiometric studies of oxidation of Mo(III) by Mo(VI) in phosphoric acid medium

Energy Technology Data Exchange (ETDEWEB)

Kumar, A; Verma, G S.P. [Ranchi Coll. (India). Dept. of Chemistry

1975-12-01

Oxidation of Mo(III) (green) by Mo(VI) in an inert atmosphere and in orthophosphoric acid medium at various acid concentrations is reported. Potentiometric and spectrophotometric data suggest that oxidation of Mo(III) proceeds to Mo(V) through a binuclear species Mo(III) Mo(IV) absorbing at 400 nm. The formation of this species is facilitated at high acid concentrations. It is further found that quantitative conversion of Mo(III) into Mo(V) takes place at fairly high acid concentrations. In high phosphoric acid concentrations, solution of Mo(III) has been found to be oxidized to Mo(VI) by air and hence this can be used as a good oxygen absorber.

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)

Conceptual study of 99Mo production in JRR-3

International Nuclear Information System (INIS)

Hirose, Akira; Komeda, Masao; Kinase, Masami; Sorita, Takami; Wada, Shigeru

2010-06-01

We investigated the production process of 99 Mo, which is parent nuclide of 99m Tc, in JRR-3. 99m Tc is most commonly used as a radiopharmaceutical in the field of nuclear medicine. Currently the supplying of 99 Mo is only dependent on imports from foreign countries, so JAEA is aiming at domestic production of a part of 99 Mo in cooperation with the industrial arena. This report presents the technical study for the production process of 99 Mo by using the neutron radiation method of (n, γ) reaction in JRR-3. (author)

Study of the activation of targets containing Mo for the production of 99Mo by the 98Mo(n,γ)99Mo nuclear reaction and the behaviour of the radionuclidic impurities of the process

International Nuclear Information System (INIS)

Nieto, Renata Correa

1998-01-01

The most used radioisotope in Nuclear Medicine is 99m Tc, in the 99 Mo- 99m Tc generator form. 99 Mo can be produced by several nuclear reactions in reactors and cyclotrons. The cyclotron production is not technically and economically viable. The production in the reactor can be done in two different ways: by the fission of 235 U and by 98 Mo(n,γ) 99 Mo reaction. A project for the production of 99 Mo by the activation of Mo and the preparation of gel type generators is under development at the 'Instituto de Pesquisas Energeticas e Nucleares'. In the present work, the radionuclidic impurities produced in the activation of MOO 3 and MoZr gel were evaluated, and these represent the two possible ways of preparing the gel of MoZr. A target of metallic Mo was also studied. The radionuclidic purity of 99m Tc eluted from generators prepared in these ways was also measured and compared with the generators prepared with fission 99 Mo. The results showed that, by all the parameters analysed, the best way of preparing the generator of 99 Mo - 99m Tc is the irradiation of MOO 3 and further preparation of the gel and the generators. (author)

Spectrophotometric and potentiometric studies of oxidation of Mo(III) by Mo(VI) in phosphoric acid medium

International Nuclear Information System (INIS)

Kumar, Arvind; Verma, G.S.P.

1975-01-01

Oxidation of Mo(III) (green) by Mo(VI) in an inert atmosphere and in orthophosphoric acid medium at various acid concentrations is reported. Potentiometric and spectrophotometric data suggest that oxidation of Mo(III) proceeds to Mo(V) through a binuclear species Mo(III) Mo(IV) absorbing at 400 nm. The formation of this species is facilitated at high acid concentrations. It is further found that quantitative conversion of Mo(III) into Mo(V) takes place at fairly high acid concentrations. In high phosphoric acid concentrations, solution of Mo(III) has been found to be oxidized to Mo(VI) by air and hence this can be used as a good oxygen absorber. (author)

LEU-fueled SLOWPOKE-2 modelling with MCNP4A

International Nuclear Information System (INIS)

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

1996-01-01

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

Large-scale studies of the Leu72Met polymorphism of the ghrelin gene in relation to the metabolic syndrome and associated quantitative traits

DEFF Research Database (Denmark)

Bing, C; Ambye, L; Fenger, M

2005-01-01

Recently, low-frequency polymorphisms in the coding region of the ghrelin gene were suggested to be involved in the aetiology of obesity and to modulate glucose-induced insulin secretion in different ethnic study groups. The objective of the present large study was to investigate whether the Leu7...

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)

Thermal transport properties of MoS ₂ and MoSe ₂ monolayers

Energy Technology Data Exchange (ETDEWEB)

Kandemir, Ali; Yapicioglu, Haluk; Kinaci, Alper; Çağın, Tahir; Sevik, Cem

2016-01-11

Isolation of single to few layer transition metal dichalgogenides open alternate venues in application of 2 dimensional materials to nanoelectronics. Either for general overheating issues or specific application in thermoelectric devices, the characterization of the thermal transport in these new low dimensional materials is needed for their efficient implementation. In this study, lattice thermal conductivities of single layer MoS₂ and MoSe₂ are evaluated using classical molecular dynamics method. The interactions between atoms are defined by Stillinger-Weber type empirical potentials that are developed to represent structural, mechanical, and vibrational properties of the given materials. In parameterization of the potentials, a stochastic optimization algorithm, namely particle swarm optimization is utilized. The final parameter sets produce quite consistent results with DFT in terms of lattice parameters, bond distances, elastic constants and vibrational properties of both single layer MoS₂ and MoSe₂. The predicted thermal properties of both materials are in very good agreement with earlier first principles calculations. The discrepancies between calculations and experimental measurements are most likely to be caused by pristine nature of the structures in our simulations.

Development on UO3-K2O and MoO3-K2O binary systems and study of UO2MoO4-MoO3 domain within UO3-MoO3-K2O ternary system

International Nuclear Information System (INIS)

Dion, C.; Noel, A.

1983-01-01

This paper confirms the previous study on the MoO 3 -K 2 O system, and constitutes a clarity of the UO 3 -K 2 O system. Four distinct uranates VI with alkaline metal/uranium ratio's 2, 1, 0,5 and 0,285 exist. Preparation conditions and powder diffraction spectra of these compounds are given. Additional informations relative to K 2 MoO 4 allotropic transformations are provided. Study of UO 2 MoO 4 -K 2 MoO 4 diagram has brought three new phases into prominence: (B) K 6 UMo 4 O 18 incongruently melting point, (E) K 2 UMo 2 O 10 congruently melting and (F) K 2 U 3 Mo 4 O 22 incongruently melting point. Within MoO 3 -K 2 MoO 4 -UO 2 MoO 4 ternary system, no new phase is found. The general appearance of ternary liquidus and crystallization fields of several compounds are given. These three new compounds become identified with these of UO 2 MoO 4 -Na 2 MoO 4 binary system [fr

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

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

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.

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

Feasibility neutronic conceptual design for the core configuration of a 75 kWth Aqueous Homogeneous Reactor for 99Mo production

International Nuclear Information System (INIS)

Milian, D.; Milian, D. E.; Rodriguez, L. P.; Salomon, J.; Cadavid, N.

2015-01-01

99m Tc is a very useful radioisotope, which is used in nearly 80% of all nuclear medicine procedures. 99m Tc is produced from 99 Mo decay. Since 2007 the medical community has been plagued by 99 Mo shortages due to aging reactors, such as the National Research Universal reactor in Canada and the High Flux Reactor in Petten, The Netherlands. At present, most of the world's supply of 99 Mo for medical isotope production involves the neutron fission of 235 U in multipurpose research reactors. 99 Mo mostly results from the fission reaction of 235 U targets with a fission yield about 6.1%. After irradiation in the reactor, the target is digested in acid or alkaline solutions and 99 Mo is recovered through a series of extraction (separation) and purification steps. 99 Mo production system in an Aqueous Homogeneous Reactor (AHR) offers a better method, because all of the 99 Mo can be extracted from the fuel solution. Over 30 AHRs has been built and operated around the world with 149 years of combined experience. In this paper, an AHR conceptual design using LEU (Low Enriched Uranium) is optimized to meet the South American demand for 99 Mo for the coming years. Aspect related with the neutronic behavior such as optimal reflector thickness, critical height, medical isotope production and others are evaluated. The neutronic calculations have been performed with the well-known MCNPX computational code. A benchmarking experiments performed at the Russian Research Center 'Kurchatov Institute' in order to validate that the developed models of AHRs with MCNPX code and the available library in XSDIR, ENDF/B VI.2, are adequate for studies of aqueous fuel solutions. (Author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-11-01

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

NMR and XAS Study of Fe-Mo Double Perovskites

International Nuclear Information System (INIS)

Zajac, D.A.; Kapusta, C.; Borowiec, M.; Sikora, M.; Marquina, C.; Blasco, J.; Ibarra, M.R.

2005-01-01

The results of NMR and XAS measurements of the A 2 FeMoO 6 double perovskites (DP) (A 2 =Sr 2 , SrBa, Ba 2 , Ca 2 ) at the Fe and Mo K edges are reported and the information on the individual site electronic and magnetic properties is analysed. The compounds studied belong to the family of materials exhibiting a high field '' colossal '' magnetoresistance as well as a low field '' giant '' magnetoresistance. Magnetoresistive properties of the compounds arise from their half-metallicity, i.e. only one spin direction being populated in the conduction band, which consists of overlapping spin down 3d Fe, 2p O and 4d Mo electron bands. Within the model, a spin-down electron undergoes a fast hopping through unoccupied oxygen 2p orbitals between Fe 3+ (3d 5 - spin up) and Mo 6+ (4d 0 ) ionic cores. This mechanism implicates an anti-parallel coupling of the Fe and Mo spins and leads to non-integer magnetic moments and a metallic character below TC. The interaction, in analogy with the '' double exchange '' (DE) in manganites, is called '' double exchange-like '' interaction. The superexchange interaction (SE) is also expected to be present, resulting also in an anti-parallel coupling of 3d Fe 3+ and 4d Mo 5+ spins through occupied oxygen 2p orbitals. The insulating character of SE is connected with an increase of the tilt angle of the Fe-O-Mo bond, which is related to a change of the structural tolerance factor f and results in structural distortions. The molybdenum NMR measurements revealed the existence of a non-integer magnetic moment at Mo and Fe, which can be attributed to the DE-like interaction. However, experiments using Moessbauer spectroscopy have shown the existence of two Fe ionisation states - with integer (SE) and non integer (DE) magnetic moments. The 95 Mo and 97 Mo NMR measurements on A 2 FeMoO 6 (A 2 =Sr 2 , SrBa, Ba 2 , Ca 2 ) presented in this work show different values of the Mo hyperfine field and the corresponding magnetic moment. This is attributed

Comparative study of NiW, NiMo and MoW prepared by mechanical alloying

International Nuclear Information System (INIS)

Gonzalez, G.; Sagarzazu, A.; Villalba, R.; Ochoa, J.

2007-01-01

The present work concern the amorphisation process induced by mechanical alloying in the NiW, NiMo and MoW systems. The alloys chosen combine a group of transition elements varying from very similar atomic radius and electronic valences (MoW) to different ones (NiW and NiMo). The three systems achieved an amorphous state after 50 h of milling. The mechanism of amorphisation proposed for NiW and NiMo was the combined effect of an excess concentration of the solute atoms entering into the structure of one of the elements and a critical concentration of defects. Continuous formation of an amorphous phase at the interface of the crystalline phase was observed during the process. MoW seems to amorphize by continuous reduction of grain size down to a critical value where the amorphisation takes place

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Studies of the effect of irradiation in a nuclear reactor, of targets containing Mo used for the preparation of 99Mo gel, material that constitutes the 99Mo - 99mTc generators

International Nuclear Information System (INIS)

Nieto, Renata Correa

2004-01-01

The most used radioisotope in Nuclear Medicine is 99m Tc, obtained in the 99 Mo - 99m Tc generator form. 99 Mo can be produced by several nuclear reactions in Cyclotron and Reactor. The production in Cyclotron is not technically and commercially feasible. The production in Nuclear Reactor can be made in two ways: 235 U fission and 99 Mo (n,γ) 99 Mo reaction. A project aiming the production of 99 Mo by activation of Mo is under way at IPEN, producing a gel type MoZr generator. There are two ways of preparing the gel and the generators: by irradiating MoO 3 and preparing the gel or by the preparation of the gel and further irradiation. This work consists in the study of the irradiation effects in several targets containing Mo for the production of 99 Mo by the 98 Mo (n,γ) 99 Mo reaction and further preparation of the gel for use as a gel type 99 Mo - 99m Tc generator. Three rinds of gel were studied: zirconium, titanium and cerium molybdate, and their morphology, infrared structure and elution yield of 99m Tc were analysed. The best results were achieved with the generators prepared with MoZr post formed gel, with amorphous structure and better elution yields. The pre formed gel induced crystallinity and worst performance of the generators. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-09-30

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

Highly efficient residue-selective labeling with isotope-labeled Ile, Leu, and Val using a new auxotrophic E. coli strain

International Nuclear Information System (INIS)

Miyanoiri, Yohei; Ishida, Yojiro; Takeda, Mitsuhiro; Terauchi, Tsutomu; Inouye, Masayori; Kainosho, Masatsune

2016-01-01

We recently developed a practical protocol for preparing proteins bearing stereo-selectively 13 C-methyl labeled leucines and valines, instead of the commonly used 13 C-methyl labeled precursors for these amino acids, by E. coli cellular expression. Using this protocol, proteins with any combinations of isotope-labeled or unlabeled Leu and Val residues were prepared, including some that could not be prepared by the precursor methods. However, there is still room for improvement in the labeling efficiencies for Val residues, using the methods with labeled precursors or Val itself. This is due to the fact that the biosynthesis of Val could not be sufficiently suppressed, even by the addition of large amounts of Val or its precursors. In this study, we completely solved this problem by using a mutant strain derived from E. coli BL21(DE3), in which the metabolic pathways depending on two enzymes, dihydroxy acid dehydratase and β-isopropylmalate dehydrogenase, are completely aborted by deleting the ilvD and leuB genes, which respectively encode these enzymes. The ΔilvD E. coli mutant terminates the conversion from α,β-dihydroxyisovalerate to α-ketoisovalerate, and the conversion from α,β-dihydroxy-α-methylvalerate to α-keto-β-methylvalerate, which produce the preceding precursors for Val and Ile, respectively. By the further deletion of the leuB gene, the conversion from Val to Leu was also fully terminated. Taking advantage of the double-deletion mutant, ΔilvDΔleuB E. coli BL21(DE3), an efficient and residue-selective labeling method with various isotope-labeled Ile, Leu, and Val residues was established.

Highly efficient residue-selective labeling with isotope-labeled Ile, Leu, and Val using a new auxotrophic E. coli strain

Energy Technology Data Exchange (ETDEWEB)

Miyanoiri, Yohei [Nagoya University, Structural Biology Research Center, Graduate School of Science (Japan); Ishida, Yojiro [Rutgers University-Robert Wood Johnson Medical School, Center for Advanced Biotechnology and Medicine (United States); Takeda, Mitsuhiro [Nagoya University, Structural Biology Research Center, Graduate School of Science (Japan); Terauchi, Tsutomu [Tokyo Metropolitan University, Graduate School of Science and Engineering (Japan); Inouye, Masayori [Rutgers University-Robert Wood Johnson Medical School, Center for Advanced Biotechnology and Medicine (United States); Kainosho, Masatsune, E-mail: kainosho@tmu.ac.jp [Nagoya University, Structural Biology Research Center, Graduate School of Science (Japan)

2016-06-15

We recently developed a practical protocol for preparing proteins bearing stereo-selectively {sup 13}C-methyl labeled leucines and valines, instead of the commonly used {sup 13}C-methyl labeled precursors for these amino acids, by E. coli cellular expression. Using this protocol, proteins with any combinations of isotope-labeled or unlabeled Leu and Val residues were prepared, including some that could not be prepared by the precursor methods. However, there is still room for improvement in the labeling efficiencies for Val residues, using the methods with labeled precursors or Val itself. This is due to the fact that the biosynthesis of Val could not be sufficiently suppressed, even by the addition of large amounts of Val or its precursors. In this study, we completely solved this problem by using a mutant strain derived from E. coli BL21(DE3), in which the metabolic pathways depending on two enzymes, dihydroxy acid dehydratase and β-isopropylmalate dehydrogenase, are completely aborted by deleting the ilvD and leuB genes, which respectively encode these enzymes. The ΔilvD E. coli mutant terminates the conversion from α,β-dihydroxyisovalerate to α-ketoisovalerate, and the conversion from α,β-dihydroxy-α-methylvalerate to α-keto-β-methylvalerate, which produce the preceding precursors for Val and Ile, respectively. By the further deletion of the leuB gene, the conversion from Val to Leu was also fully terminated. Taking advantage of the double-deletion mutant, ΔilvDΔleuB E. coli BL21(DE3), an efficient and residue-selective labeling method with various isotope-labeled Ile, Leu, and Val residues was established.

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)

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...

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.

Development of annular targets for 99MO production-1999

International Nuclear Information System (INIS)

Conner, C.; Lewandowski, E. F.; Snelgrove, J. L.; Liberatore, M. W.; Walker, D. E.; Wiencek, T. C.; McGann, D. J.; Hofman, G. L.; Vandegrift, G. F.

1999-01-01

The new annular target performed well during irradiation. The target is inexpensive and provides good heat transfer during irradiation. Based on these and previous tests, we conclude that targets with zirconium tubes and either nickel-plated or zinc-plated foils work well. We proved that we could use aluminum target tubes, which are much cheaper and easier to work with than the zirconium tubes. In aluminum target tubes nickel-plated fission-recoil barriers work well and prevent bonding of the foil to the new target tubes during irradiation. Also, zinc-plated and aluminum-foil barriers appear promising in anodized aluminum tubes. Additional tests are anticipated to address such issues as fission-recoil barrier thickness and uranium foil composition. Overall, however, the target was successful and will provide an inexpensive, efficient way to irradiate LEU metal foil for the production of 99 Mo

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)

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

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.

Abstracts and papers of the 2000 International RERTR Meeting

International Nuclear Information System (INIS)

2000-10-01

This meeting was devoted to progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the year 2000. The main activities planned for the year 2001 were discussed. Year 2000 was held by important accomplishments and events for the RERTR program. Four additional shipments containing 503 spent fuel assemblies from foreign research reactors were accepted by the U.S. Altogether, 3,740 spent fuel assemblies from foreign research reactors have been received by the U.S. under the acceptance policy. Qualification of the U-Mo dispersion fuels is proceeding on schedule. Test fuel elements with uranium density of 6 g/cm3 are being fabricated by BWXT and are scheduled to begin undergoing irradiation in the HFR-Petten in the spring of 2001, with a goal of qualifying this fuel by the end of 2003. U-Mo fuel with uranium density of 8-9 g/cm3 is planned to be qualified by the end of 2005. Joint LEU conversion feasibility studies were completed for HFR-Petten and for SAFARI-1. Significant improvements were made in the design of LEU metal-foil annular targets that would allow efficient production of fission 99Mo. Progress was made on irradiation testing of LEU UO 2 dispersion fuel and on LEU conversion feasibility studies in the Russian RERTR program. Conversion of the BER-II reactor in Germany, was completed and conversion of the La Reina reactor in Chile, began. In the fuel development area, the RERTR program is pursuing qualification of high-density LEU U-Mo dispersion fuels, with the dual goal of enabling further conversions and of developing a substitute for LEU silicide fuels that can be more easily disposed of after expiration of the FRR SNF Acceptance Program. The 99Mo effort has reached the point where it appears feasible for all the 99Mo producers of the world to agree jointly to a common course of action leading to the elimination of HEU use in their processes. It was concluded

Interdiffusion studies on hot rolled U-10Mo/AA1050

Energy Technology Data Exchange (ETDEWEB)

Saliba-Silva, A.M.; Martins, I.C.; Carvalho, E.U.; Durazzo, M.; Riella, H.G. [Instituto de Pesquisas Energeticas e Nucleares (CCN/IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Combustivel Nuclear], e-mail: saliba@ipen.br

2010-07-01

The U-Mo alloys are investigated with the goal of becoming nuclear material to fabricate high-density fuel elements for high performance research reactors. This enrichment level suggests that the U-Mo alloys should be between 6 to 10wt%, which can give up to 9gU/cm{sup 3} as fuel density. Nevertheless, the U-Mo alloys are very reactive with Al. Interdiffusion reaction products are formed since nuclear fission promotes chemical interaction layer during operation, leading to potential structural failure. Present studies were made with treated hot rolled diffusion couples of U-10Mo inserted in Al (AA1050). The U-10Mo/AA1050 pairs were treated in two temperature (150 degree C and 550 degree C) with three soaking times (5h, 40h and 80h). From microstructure analyses, rapid diffusion of Al happened inside U-10Mo in the first heating at 540 degree C during 15 min, reaching 8 at%Al in a range of 200 {mu}m towards U-10Mo. Longer time (5, 40, 80h) at 550 degree C maintain this level of Al-content up to 1000 {mu}m inside U-10Mo. A minor depth ({approx}1 {mu}m) near the interdiffusion contact had higher Al-content, but not sufficient to form identifiable (U,Mo)Al{sub x} structures. Probably, residual elements reduced drastically the interdiffusion phenomena between U-10Mo and AA1050, maybe due to silicon presence. (author)

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)

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.

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)

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)

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

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.

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

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

RERTR activities in Argentina

International Nuclear Information System (INIS)

Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; De La Fuente, M.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Perez, A.; Piazza, A.; Ruggirello, G.; Taboada, H.

2000-01-01

The Atomic Energy Commission of Argentina has been involved in the Reduced Enrichment for Research and Test Reactors Program since 1978. The most relevant milestones of the program, regarding fuel R and D activities, were the development and manufacturing at industrial scale of U 3 O 8 dispersed fuel assemblies and the conversion of the RA-3 reactor core to LEU fuel. More recently, the activities were focused in the development of high density U 3 Si 2 fuel with a density of 4.8 gU/cm 3 and the improvement of the manufacturing process of U 3 Si 2 powder. Currently one of the main objectives is to develop and qualify the technology for the production of high-density LEU fuel elements using U-Mo alloy. Several alternative ways to obtain U-Mo powder are under development with the aim of evaluating plant scale production and costs. To boost this program the main research reactor of Argentina, the RA-3, will be upgraded to 10 MW early in 2001 and the hot cells at the Ezeiza Atomic Center are fully operational after important investments. Significant progresses were also carried out in the development of LEU targets for the production of Mo 99 . Experimental work has demonstrated the feasibility of the manufacturing and radiochemical processing of miniplate targets prepared with dispersed UAl x , maintaining the geometry and the alkaline processing of the HEU targets used so far. (author)

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.

Compatibility studies on Mo-coating systems for nuclear fuel cladding applications

Science.gov (United States)

Koh, Huan Chin; Hosemann, Peter; Glaeser, Andreas M.; Cionea, Cristian

2017-12-01

To improve the safety factor of nuclear power plants in accident scenarios, molybdenum (Mo), with its high-temperature strength, is proposed as a potential fuel-cladding candidate. However, Mo undergoes rapid oxidation and sublimation at elevated temperatures in oxygen-rich environments. Thus, it is necessary to coat Mo with a protective layer. The diffusional interactions in two systems, namely, Zircaloy-2 (Zr2) on a Mo tube, and iron-chromium-aluminum (FeCrAl) on a Mo rod, were studied by aging coated Mo substrates in high vacuum at temperatures ranging from 650 °C to 1000° for 1000 h. The specimens were characterized using scanning electron microscopy (SEM), energy-dispersive spectrometry (EDS) and nanoindentation. In both systems, pores in the coating increased in size and number with increasing temperature over time, and cracks were also observed; intermetallic phases formed between the Mo and its coatings.

Uranium Anodic Dissolution under Slightly Alkaline Conditions Progress Report Full-Scale Demonstration with DU Foil

Energy Technology Data Exchange (ETDEWEB)

Gelis, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Brown, M. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Wiedmeyer, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, G. F. [Argonne National Lab. (ANL), Argonne, IL (United States)

2014-02-18

Argonne National Laboratory (Argonne) is developing an alternative method for digesting irradiated low enriched uranium (LEU) foil targets to produce ⁹⁹Mo in neutral/alkaline media. This method consists of the electrolytic dissolution of irradiated uranium foil in sodium bicarbonate solution, followed by precipitation of base-insoluble fission and activation products, and uranyl-carbonate species with CaO. The addition of CaO is vital for the effective anion exchange separation of ⁹⁹MoO₄^2- from the fission products, since most of the interfering anions (e.g., CO₃^2-) are removed from the solution, while molybdate remains in solution. An anion exchange is used to retain and to purify the ⁹⁹Mo from the filtrate. The electrochemical dissolver has been designed and fabricated in 304 stainless-steel (SS), and tested for the dissolution of a full-size depleted uranium (DU) target, wrapped in Al foil. Future work will include testing with low-burn-up DU foil at Argonne and later with high-burn-up LEU foils at Oak Ridge National Laboratory.

Excitation functions of deuteron induced nuclear reactions on natMo up to 21 MeV. An alternative route for the production of 99mTc and 99Mo

International Nuclear Information System (INIS)

Sonck, M.; Hermanne, A.; Takacs, S.; Szelecsenyi, F.; Tarkanyi, F.

1999-01-01

Cross sections of deuteron induced nuclear reactions on natural molybdenum have been studied in the frame of a systematic investigation of charged particle induced nuclear reactions on metals for different applications. The excitation functions of 92m,95 Nb-, 93,94g,94m,95g,95m,96,99m Tc- and 99 Mo were measured up to 21 MeV deuteron energy by using stacked foil technique and activation method. The goal of this work was to study the production possibility of the medical important 94m,99m Tc- and 99 Mo-nuclides. Production of 99m Tc and 99 Mo is of importance for their use in nuclear medicine, whereas 94m Tc is of interest regarding quantification of kinetics of well-established 99m Tc-radiopharmaceuticals. The production possibilities of 99m Tc and 99 Mo above 20 MeV deuteron energies up to 50 MeV were estimated and was found that beside the proton induced reactions the deuteron induced reactions on enriched molybdenum target are very promising. (author)

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

An experimental study: Role of different ambient on sulfurization of MoO{sub 3} into MoS{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Kumar, Prabhat, E-mail: prabhat89k@gmail.com; Singh, Megha; Sharma, Rabindar K.; Reddy, G.B.

2016-06-25

Molybdenum disulfide (MoS{sub 2}) nanostructured thin films (NTFs) were synthesised by sulfurizing MoO{sub 3} NTFs using three different non-conventional methods (named methods 1–3). Method 1 uses sulfur vapors, second employs H{sub 2}S/Ar gas and third adopts plasma of H{sub 2}S/Ar gas. HRTEM revealed formation of core–shell nanostructures with maximum shell thickness obtained in method 3. The samples showed uniform nanoflakes (NFs) throughout substrate, revealed by SEM, same as their precursor MoO{sub 3.} XRD and Raman analysis disclosed crystalline MoS{sub 2} and degree of crystallinity was greatest in case of sulfurization in plasma ambient. Quantitative analysis of sulfurized films carried out by XPS shows presence of MoS{sub 2} in all three methods with percentage found to be 18%, 87% and ∼100% respectively. The effect of sulfurizing ambient on its efficiency to convert MoO{sub 3} into MoS{sub 2} has been studied and it was found out that plasma ambient has resulted in high quality of MoS{sub 2} NTFs based on parameters as crystallinity, purity, uniformity and stoichiometry control. - Highlights: • Comparison of three non-conventional methods of sulfurization. • Parameters used for comparison are crystallinity, purity, sulfurized thickness, uniformity and stoichiometry. • H{sub 2}S/Ar plasma based method came out to be best among other techniques. • A soft template reactions for sulfurization of MoO{sub 3} nanoflake is proposed.

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

Preparation and Evaluation at the Delta Opioid Receptor of a Series of Linear Leu-Enkephalin Analogues Obtained by Systematic Replacement of the Amides

Science.gov (United States)

2013-01-01

Leu-enkephalin analogues, in which the amide bonds were sequentially and systematically replaced either by ester or N-methyl amide bonds, were prepared using classical organic chemistry as well as solid phase peptide synthesis (SPPS). The peptidomimetics were characterized using competition binding, ERK1/2 phosphorylation, receptor internalization, and contractility assays to evaluate their pharmacological profile over the delta opioid receptor (DOPr). The lipophilicity (LogD7.4) and plasma stability of the active analogues were also measured. Our results revealed that the last amide bond can be successfully replaced by either an ester or an N-methyl amide bond without significantly decreasing the biological activity of the corresponding analogues when compared to Leu-enkephalin. The peptidomimetics with an N-methyl amide function between residues Phe and Leu were found to be more lipophilic and more stable than Leu-enkephalin. Findings from the present study further revealed that the hydrogen-bond donor properties of the fourth amide of Leu-enkephalin are not important for its biological activity on DOPr. Our results show that the systematic replacement of amide bonds by isosteric functions represents an efficient way to design and synthesize novel peptide analogues with enhanced stability. Our findings further suggest that such a strategy can also be useful to study the biological roles of amide bonds. PMID:23650868

Phase formation in Na2MoO4 - MgMoO4 - Cr2(MoO4)3 system

International Nuclear Information System (INIS)

Kotova, I.Yu.; Kozhevnikova, N.M.

1998-01-01

Interaction within Na 2 MoO 4 - MgMoO 4 - Cr 2 (MoO 4 ) 3 ternary system is studied by X ray phase and DTA methods. State diagram of NaCr(MoO 4 ) 2 - MgMoO 4 section is plotted. Production of ternary molybdates of Na 1-x Mg 1-x Cr 1+x (MoO 4 ) 3 , where 0 ≤ x ≤ 0.3, and NaMg 3 Cr(MoO 4 ) 5 composition is ascertained [ru

The lanthanum(III molybdate(VI La4Mo7O27

Directory of Open Access Journals (Sweden)

Petra Becker

2009-08-01

Full Text Available Crystals of the orthorhombic phase La4Mo7O27 (lanthanum molybdenum oxide were obtained from a non-stoichiometric melt in the pseudo-ternary system La2O3–MoO3–B2O3. In the crystal structure, distorted square-antiprismatic [LaO8] and monocapped square-antiprismatic [LaO9] polyhedra are connected via common edges and faces into chains along [010]. These chains are arranged in layers that alternate with layers of [MoO4] and [MoO5] polyhedra parallel to (001. In the molybdate layers, a distorted [MoO5] trigonal bipyramid is axially connected to two [MoO4] tetrahedra, forming a [Mo3O11] unit.

Montreal Cognitive Assessment (MoCA): validation study for frontotemporal dementia.

Science.gov (United States)

Freitas, Sandra; Simões, Mário R; Alves, Lara; Duro, Diana; Santana, Isabel

2012-09-01

The Montreal Cognitive Assessment (MoCA) is a brief instrument developed for the screening of milder forms of cognitive impairment, having surpassed the well-known limitations of the Mini-Mental State Examination (MMSE). The aim of the present study was to validate the MoCA as a cognitive screening test for behavioral-variant frontotemporal dementia (bv-FTD) by examining its psychometric properties and diagnostic accuracy. Three matched subgroups of participants were considered: bv-FTD (n = 50), Alzheimer disease (n = 50), and a control group of healthy adults (n = 50). Compared with the MMSE, the MoCA demonstrated consistently superior psychometric properties and discriminant capacity, providing comprehensive information about the patients' cognitive profiles. The diagnostic accuracy of MoCA for bv-FTD was extremely high (area under the curve AUC [MoCA] = 0.934, 95% confidence interval [CI] = 0.866-.974; AUC [MMSE] = 0.772, 95% CI = 0.677-0.850). With a cutoff below 17 points, the MoCA results for sensitivity, specificity, positive predictive value, negative predictive value, and classification accuracy were significantly superior to those of the MMSE. The MoCA is a sensitive and accurate instrument for screening the patients with bv-FTD and represents a better option than the MMSE.

Screening for Cognitive Impairment in Parkinson's Disease: Improving the Diagnostic Utility of the MoCA through Subtest Weighting.

Directory of Open Access Journals (Sweden)

Sophie Fengler

Full Text Available Given the high prevalence of cognitive impairment in Parkinson's disease (PD, cognitive screening is important in clinical practice. The Montreal Cognitive Assessment (MoCA is a frequently used screening test in PD to detect mild cognitive impairment (PD-MCI and Parkinson's disease dementia (PD-D. However, the proportion in which the subtests are represented in the MoCA total score does not seem reasonable. We present the development and preliminary evaluation of an empirically based alternative scoring system of the MoCA which aims at increasing the overall diagnostic accuracy.In study 1, the MoCA was administered to 40 patients with PD without cognitive impairment (PD-N, PD-MCI, or PD-D, as defined by a comprehensive neuropsychological test battery. The new MoCA scoring algorithm was developed by defining Areas under the Curve (AUC for MoCA subtests in a Receiver Operating Characteristic (ROC and by weighting the subtests according to their sensitivities and specificities. In study 2, an independent sample of 24 PD patients (PD-N, PD-MCI, or PD-D was tested with the MoCA. In both studies, diagnostic accuracy of the original and the new scoring procedure was calculated.Diagnostic accuracy increased with the new MoCA scoring algorithm. In study 1, the sensitivity to detect cognitive impairment increased from 62.5% to 92%, while specificity decreased only slightly from 77.7% to 73%; in study 2, sensitivity increased from 68.8% to 81.3%, while specificity stayed stable at 75%.This pilot study demonstrates that the sensitivity of the MoCA can be enhanced substantially by an empirically based weighting procedure and that the proposed scoring algorithm may serve the MoCA's actual purpose as a screening tool in the detection of cognitive dysfunction in PD patients better than the original scoring of the MoCA. Further research with larger sample sizes is necessary to establish efficacy of the alternate scoring system.

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)

Analysis of manganese superoxide dismutase (MnSOD: Ala-9Val and glutathione peroxidase (GSH-Px: Pro 197 Leu gene polymorphisms in mood disorders.

Directory of Open Access Journals (Sweden)

Birgül Elbozan Cumurcu

2013-05-01

Full Text Available We investigated the etiopathogenetic role of manganese superoxide dismutase (MnSOD (Ala-9Val and glutathione peroxidase (GSH-Px (Pro 197 Leu gene polymorphisms in patients diagnosed with major depressive disorder (MDD and bipolar I disorder (BD. Eighty patients with MDD, 82 patients with BD (total 162 patients and 96 healthy controls were enrolled in this study and genotyped using a Real Time-Quantitative Polymer Chain Reaction (RT-qPCR-based method. The patients with BD and MDD and the controls had a similar distribution of the genotypes and alleles in the Ala-9Val MnSOD gene polymorphism. Comparison of the MDD group and control group regarding the Pro197 Leu GSH-Px gene polymorphism revealed similar genotype distribution but different allele distribution. The BD group and control group were similar both for genotypes and for alleles when compared regarding the Pro 197 Leu GSH-Px gene polymorphism. The combined analysis (MDD plus BD also failed to find any association between the Ala-9Val MnSOD and Pro 197 Leu GSH-Px gene polymorphism. Although small statistical power of the current study the significant difference between patients with depression and the control group for the Pro 197 Leu GSH-Px polymorphism indicates that the distribution of these alleles may have a contribution in the physiopathogenesis of depression. One of the limitation of the current study is that the sample size is too small. Understanding of the exact role of Pro 197 LeuGSH-Px polymorphism in the development of depression needs to further studies with more sample size and high statistical power.

Variable system: An alternative approach for the analysis of mediated moderation.

Science.gov (United States)

Kwan, Joyce Lok Yin; Chan, Wai

2018-06-01

Mediated moderation (meMO) occurs when the moderation effect of the moderator (W) on the relationship between the independent variable (X) and the dependent variable (Y) is transmitted through a mediator (M). To examine this process empirically, 2 different model specifications (Type I meMO and Type II meMO) have been proposed in the literature. However, both specifications are found to be problematic, either conceptually or statistically. For example, it can be shown that each type of meMO model is statistically equivalent to a particular form of moderated mediation (moME), another process that examines the condition when the indirect effect from X to Y through M varies as a function of W. Consequently, it is difficult for one to differentiate these 2 processes mathematically. This study therefore has 2 objectives. First, we attempt to differentiate moME and meMO by proposing an alternative specification for meMO. Conceptually, this alternative specification is intuitively meaningful and interpretable, and, statistically, it offers meMO a unique representation that is no longer identical to its moME counterpart. Second, using structural equation modeling, we propose an integrated approach for the analysis of meMO as well as for other general types of conditional path models. VS, a computer software program that implements the proposed approach, has been developed to facilitate the analysis of conditional path models for applied researchers. Real examples are considered to illustrate how the proposed approach works in practice and to compare its performance against the traditional methods. (PsycINFO Database Record (c) 2018 APA, all rights reserved).

Influence of chemical composition in crystallographic texture Fe-Cr-Mo alloys; Influencia da composicao quimica na textura cristalografica de ligas Fe-Cr-Mo

Energy Technology Data Exchange (ETDEWEB)

Moura, L.B.; Guimaraes, R.F. [Instituto Federal de Educacao, Ciencia e Tecnologia do Ceara, Fortaleza, CE (Brazil). Dept. da Industria; Abreu, H.F.G. [Universidade Federal do Ceara (UFC), Fortaleza, CE (Brazil)

2010-07-01

The use of steels with higher contents of Mo in the oil industry has been an alternative to reduce the effect of naphthenic corrosion in refining units. The addition of Mo in Fe-Cr alloys in the same manner that increases resistance to corrosion naphthenic causes some difficulties such as difficulty of forming, welding and embrittlement. In this work, experimental ingots of Fe-Cr-Mo alloys (Cr - 9, 15 and 17%, Mo - 5, 7 and 9%) were melted in vacuum induction furnace and hot and cold rolled in a laboratory rolling mill. The influence of chemical composition on crystallographic texture of samples subjected to the same thermo-mechanical treatment was analyzed by x-ray diffraction. The results indicate that fiber (111) becomes more intense with increasing Mo and/or Cr contents. (author)

Study of phase equilibria in LiIn(MoO4)2 - MeIn(MoO4)2 (Me - K, Rb) systems

International Nuclear Information System (INIS)

Smirnyagina, N.N.; Kozhevnikova, N.M.; Alekseev, F.P.; Mokhosoev, M.V.

1983-01-01

To determine the possibilities of formation of ternary molybdates, containing two different alkali cations and a cation of trivalent element, the qUasibinary LiIn(MoO 4 ) 2 -MeIn(MoO 4 ) 2 cross-sections of quaternary Li 2 O-Me 2 O-In 2 O 3 -MoO 3 , (Me-K, Rb) systems have been studied. Methods of X-ray phase-, differential thermal- and crystal optical analyses were used. The studied systems are eutectics with segregation; ternary compounds are not formed in theM

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-12-01

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

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...

Synthesis, structure and magnetic properties of the one-dimensional bimetallic oxide [Cu(terpy)Mo2O7

International Nuclear Information System (INIS)

Burkholder, Eric; Gabriel Armatas, N.; Golub, Vladimir; O'Connor, Charles J.; Zubieta, Jon

2005-01-01

The hydrothermal reaction of Cu(CH 3 CO 2 ) 2 .H 2 O, Na 2 MoO 4 and terpyridine at 140 deg. C for 48 h yields [Cu(terpy)Mo 2 O 7 ] (1), a bimetallic one-dimensional oxide. The structure consists of ruffled chains of edge- and corner-sharing {MoO 5 } square pyramids, decorated with {CuN 3 O 2 } '4+1' axially distorted square pyramids. The Cu(II) polyhedra are disposed so as to produce an alternating pattern of Cu-Cu distances across the {Mo 2 O 2 } rhombi of the chain of 6.25 and 6.82 A. This structural feature is reflected in the magnetic properties which are characteristic of a dimer rather than a linear chain, consistent with an alternating antiferromagnetic Heisenberg chain. -- Graphical abstract: Hydrothermal synthesis provided the one-dimensional bimetallic oxide [Cu(terpy)Mo 2 O 7 ], a material consisting of a zig-zag {Mo 2 O 7 } n 2 n - chain, decorated with {Cu(terpy)} 2+ groups exhibiting alternating short-long Cu-Cu distances between copper sites

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.)

Phase formation in the K2MoO4-Lu2(MoO4)3-Hf(MoO4)2 system and the structural study of triple molybdate K5LuHf(MoO4)6

International Nuclear Information System (INIS)

Romanova, E.Yu.; Bazarov, B.G.; Tushinova, Yu.L.; Fedorov, K.N.; Bazarova, Zh.G.; Klevtsova, R.F.; Glinskaya, L.A.

2007-01-01

Interactions in the ternary system K 2 MoO 4 -Lu 2 (MoO 4 ) 3 -Hf(MoO 4 ) 2 have been studied by X-ray powder diffraction and differential thermal analysis. A new triple (potassium lutetium hafnium) molybdate with the 5 : 1 : 2 stoichiometry has been found. Monocrystals of this molybdate have been grown. Its X-ray diffraction structure has been refined (an X8 APEX automated diffractometer, MoK α radiation, 1960 F(hkl), R = 0.0166). The trigonal unit cell has the following parameters: a = 10.6536(1) A, c = 37.8434(8) A, V=3719.75(9) A, Z = 6, space group R3-bar c. The mixed 3D framework of the structure is built of Mo tetrahedra sharing corners with two independent (Lu,Hf)O 6 octahedra. Two sorts of potassium atoms occupy large framework voids [ru

Evaluation of proton induced reactions on sup 1 sup 0 sup 0 Mo. New cross sections for production of sup 9 sup 9 sup m Tc and sup 9 sup 9 Mo

CERN Document Server

Takács, S; Tarkanyi, F; Hermanne, A; Sonck, M

2003-01-01

The use of the sup 9 sup 9 Mo -> sup 9 sup 9 sup m Tc generator in nuclear medicine is well established world wide. The production of the sup 9 sup 9 Mo (T sub 1 sub / sub 2 = 66 h) parent as a fission product of sup 2 sup 3 sup 5 U is largely based on the use of reactor technology. From the early 1990's accelerator based production methods to provide either direct produced sup 9 sup 9 sup m Tc or the parent sup 9 sup 9 Mo, were studied and suggested as potential alternatives to the reactor based production of sup 9 sup 9 Mo. A possible pathway for the charged particle production of sup 9 sup 9 sup m Tc and sup 9 sup 9 Mo is irradiation of molybdenum metal with protons via the reaction sup 1 sup 0 sup 0 Mo(p,2n) sup 9 sup 9 sup m Tc and sup 1 sup 0 sup 0 Mo(p,pn) sup 9 sup 9 Mo, respectively. The earlier published excitation functions show large differences in their maximum that result in large differences in the calculated yields. Study the excitation function for these proton-induced reactions was decided. ...

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.

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)

Preliminary Thermohydraulic Analysis of a New Moderated Reactor Utilizing an LEU-Fuel for Space Nuclear Thermal Propulsion

International Nuclear Information System (INIS)

Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung

2015-01-01

The Korea Advanced NUclear Thermal Engine Rocket utilizing an LEU fuel (KANUTER-LEU) is a non-proliferative and comparably efficient NTR engine with relatively low thrust levels of 40 - 50 kN for in-space transportation. The small modular engine can expand mission versatility, when flexibly used in a clustered engine arrangement, so that it can perform various scale missions from low-thrust robotic science missions to high-thrust manned missions. In addition, the clustered engine system can enhance engine redundancy and ensuing crew safety as well as the thrust. The propulsion system is an energy conversion system to transform the thermal energy of the reactor into the kinetic energy of the propellant to produce the powers for thrust, propellant feeding and electricity. It is mainly made up of a propellant Feeding System (PFS) comprising a Turbo-Pump Assembly (TPA), a Regenerative Nozzle Assembly (RNA), etc. For this core design study, an expander cycle is assumed to be the propulsion system. The EGS converts the thermal energy of the EHTGR in the idle operation (only 350 kW th power) to electric power during the electric power mode. This paper presents a preliminary thermohydraulic design analysis to explore the design space for the new reactor and to estimate the referential engine performance. The new non-proliferative NTR engine concept, KANUTER-LEU, is under designing to surmount the nuclear proliferation obstacles on allR and Dactivities and eventual commercialization for future generations. To efficiently implement a heavy LEU fuel for the NTR engine, its reactor design innovatively possesses the key characteristics of the high U density fuel with high heating and H 2 corrosion resistances, the thermal neutron spectrum core and also minimizing non-fission neutron loss, and the compact reactor design with protectively cooling capability. To investigate feasible design space for the moderated EHTGR-LEU and resultant engine performance, the preliminary design

Electronic structure investigation of MoS2 and MoSe2 using angle-resolved photoemission spectroscopy and ab initio band structure studies.

Science.gov (United States)

Mahatha, S K; Patel, K D; Menon, Krishnakumar S R

2012-11-28

Angle-resolved photoemission spectroscopy (ARPES) and ab initio band structure calculations have been used to study the detailed valence band structure of molybdenite, MoS(2) and MoSe(2). The experimental band structure obtained from ARPES has been found to be in good agreement with the theoretical calculations performed using the linear augmented plane wave (LAPW) method. In going from MoS(2) to MoSe(2), the dispersion of the valence bands decreases along both k(parallel) and k(perpendicular), revealing the increased two-dimensional character which is attributed to the increasing interlayer distance or c/a ratio in these compounds. The width of the valence band and the band gap are also found to decrease, whereas the valence band maxima shift towards the higher binding energy from MoS(2) to MoSe(2).

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)

The crisis of 99Mo. Current trends and challenges

International Nuclear Information System (INIS)

Leyva, René; Duatti, Adriano

2016-01-01

In recent years they have exacerbated the problems with the availability of 99 Mo in the international market for the production of radio isotopic generators 99 Mo / 99 mTc. The main situation is given by the shutdown, maintenance or damage to the nuclear reactors of the reductor plants. The search for solutions, analysis of alternatives and development of new methods for the 99 Mo, the well-known workhorse of nuclear medicine, are addressed in this paper. Are interesting new routes using cyclotrons. Also shown are the steps taken internationally to alleviate the growing demand against a limited supply in production. (author)

Improving the representation of modal choice into bottom-up optimization energy system models - The MoCho-TIMES model

DEFF Research Database (Denmark)

Tattini, Jacopo; Ramea, Kalai; Gargiulo, Maurizio

2018-01-01

and mathematical expressions required to develop the approach. This study develops MoCho-TIMES in the standalone transportation sector of TIMES-DK, the integrated energy system model for Denmark. The model is tested for the Business as Usual scenario and for four alternative scenarios that imply diverse......This study presents MoCho-TIMES, an original methodology for incorporating modal choice into energy-economy-environment-engineering (E4) system models. MoCho-TIMES addresses the scarce ability of E4 models to realistically depict behaviour in transport and allows for modal shift towards transit...

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)

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

Dipole strength distribution below the giant dipole resonance in {sup 92}Mo, {sup 98}Mo and {sup 100}Mo

Energy Technology Data Exchange (ETDEWEB)

Rusev, G.Y.

2006-07-01

Investigations of the dipole-strength distributions in {sup 92}Mo, {sup 98}Mo and {sup 100}Mo were carried out by means of the method of nuclear resonance fluorescence. The low-lying excitations in the nuclides {sup 92}Mo, {sup 98}Mo and {sup 100}Mo have been studied in photon-scattering experiments at an electron energy of 6 MeV at the ELBE accelerator and at electron energies from 3.2 to 3.8 MeV at the Dynamitron accelerator. Five levels were observed in {sup 92}Mo. Five levels in {sup 98}Mo and 14 in {sup 100}Mo were identified for the first time in the energy range from 2 to 4 MeV. Dipole-strength distributions up to the neutron-separation energies in the nuclides {sup 92}Mo, {sup 98}Mo and {sup 100}Mo have been investigated at the ELBE accelerator. Because of the possible observation of transitions in the neighboring nuclei produced via ({gamma},n) reaction, additional measurements at electron energies of 8.4 and 7.8 MeV, below the neutron-separation energy, were performed on {sup 98}Mo and {sup 100}Mo, respectively. The number of transitions assigned to {sup 92}Mo, {sup 98}Mo and {sup 100}Mo is 340, 485 and 499, respectively, the main part of them being dipole transitions. Statistical properties of the observed transitions are obtained. The continuum contains the ground-state transitions as well as the branching transitions to the low-lying levels and the subsequent deexcitations of these levels. (orig.)

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)

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

Microstructure and properties of MoSi2-MoB and MoSi2-Mo5Si3 molybdenum silicides

International Nuclear Information System (INIS)

Schneibel, J.H.; Sekhar, J.A.

2003-01-01

MoSi 2 -based intermetallics containing different volume fractions of MoB or Mo 5 Si 3 were fabricated by hot-pressing MoSi 2 , MoB, and Mo 5 Si 3 powders in vacuum. Both classes of alloys contained approximately 5 vol.% of dispersed silica phase. Additions of MoB or Mo 5 Si 3 caused the average grain size to decrease. The decrease in the grain size was typically accompanied by an increase in flexure strength, a decrease in the room temperature fracture toughness, and a decrease in the hot strength (compressive creep strength) measured around 1200 deg. C, except when the Mo 5 Si 3 effectively became the major phase. Oxidation measurements on the two classes of alloys were carried out in air. Both classes of alloys were protected from oxidation by an in-situ adherent scale that formed on exposure to high temperature. The scale, although not analyzed in detail, is commonly recognized in MoSi 2 containing materials as consisting mostly of SiO 2 . The MoB containing materials showed an increase in the scale thickness and the cyclic oxidation rate at 1400 deg. C when compared with pure MoSi 2 . However, in contrast with the pure MoSi 2 material, oxidation at 1400 deg. C began with a weight loss followed by a weight gain and the formation of the protective silica layer. The Mo 5 Si 3 containing materials experienced substantial initial weight losses followed by regions of small weight changes. Overall, the MoB and Mo 5 Si 3 additions to MoSi 2 tended to be detrimental for the mechanical and oxidative properties

Sustainable one-step synthesis of hierarchical microspheres of PEGylated MoS2 nanosheets and MoO3 nanorods: Their cytotoxicity towards lung and breast cancer cells

Science.gov (United States)

Kumar, Neeraj; George, Blassan Plackal Adimuriyil; Abrahamse, Heidi; Parashar, Vyom; Ngila, Jane Catherine

2017-02-01

Nanotechnology provides an emerging potent alternate mode of cancer therapy. Nanomaterials dispersion or solubility is of particular concern in utilising their full potential applications in biomedical fields. PEGylation of nanomaterials is considered to provide products with stealth properties, and physiological environment with no obvious adverse effects. The purpose of this work was to develop a sustainable one-step method for fabrication of hierarchical microspheres of PEGylated MoS2 nanosheets using a stoichiometric ratio of Mo(VI) and thiourea. This study further investigated the cytotoxicity of the PEGylated MoS2 nanosheets towards lung (A549) and breast cancer (MCF-7) cell lines by analysing morphological changes and performing dose-dependent cell proliferation, and cytotoxicity analysis using adenosine 5‧-triphosphate (ATP), and lactate dehydrogenase (LDH) assay. For comparison, MoO3 nanorods were synthesised by simple chemical route and their cytotoxicity towards lung (A549) and breast cancer (MCF-7) cell lines were checked. The findings suggested that PEGylated MoS2 nanosheets have excellent cytotoxicity towards breast cancer (MCF-7) cell lines, and MoO3 have better cytotoxicity towards lung (A549) cancer cell lines. This work envisages an accessible foundation for engineering sophisticated biomolecule-MoS2 nanosheets conjugation due to the defect-rich biocompatible surface, to achieve great versatility, additional functions, and further advances in the biomedical field.

Alternative generators of the 99mTc

International Nuclear Information System (INIS)

Khujaev, S.

2004-01-01

9 9mTc is the most widely used radionuclide in nuclear medicine. 9 9mTc radionuclide is obtained from a generator, in which 9 9Mo servest as the parent radionuclide. In the generator 9 9mTc and 9 9Mo radionuclides are found in genetic balance and 9 9mTc radionuclide is chemically extracted from the system periodically. Although there already exists many ways and variants of manufacturing 9 9mTc generators, search for new variants of the 9 9Mo → 9 9mTc generator systems continue. An example is the investigations carried out with the support of the IAEA. In these research works, generators based on elution of poly molybdate gels have been developed and evaluated. These generators will be serving as alternative technologies for production of 9 9mTc radionuclide, which use 9 9Mo produced by non-fission means. It is known that in Australia and China more than 30% of 9 9mTc generators are gel-generators. The works of authors are devoted to the problem of searching new perspective materials as a column material that will serve as adsorbent. The main purpose of all the research on alternative technologies is the usage of parent radionuclide 9 9Mo that is obtained from 9 8Mo(n, γ) 9 9Mo reaction instead of as a fission product. Our work examines the possibility of reception of generators 9 9Mo → 9 9mTc using non-fission 9 9Mo that is based on insoluble salts of molybdate

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

Studies of mixed HEU-LEU-MTR cores using 3D models

Energy Technology Data Exchange (ETDEWEB)

Haenggi, P.; Lehmann, E.; Hammer, J.; Christen, R. [Paul Scherrer Institute, Villigen (Switzerland)

1997-08-01

Several different core loadings were assembled at the SAPHIR research reactor in Switzerland combining the available types of MTR-type fuel elements, consisting mainly of both HEU and LEU fuel. Bearing in mind the well known problems which can occur in such configurations (especially power peaking), investigations have been carried out for each new loading with a 2D neutron transport code (BOXER). The axial effects were approximated by a global buckling value and therefore the radial effects could be studied in considerably detail. Some of the results were reported at earlier RERTR meetings and were compared to those obtained by other methods and with experimental values. For the explicit study of the third dimension of the core, another code (SILWER), which has been developed in PSI for LWR power plant cores, has been selected. With the help of an adapted model for the MTR-core of SAPHIR, several important questions have been addressed. Among other aspects, the estimation of the axial contribution to the hot channel factors, the influence of the control rod position and of the Xe-poisoning on the power distribution were studied. Special attention was given to a core position where a new element was assumed placed near a empty, water filled position. The comparison of elements of low and high enrichments at this position was made in terms of the induced power peaks, with explicit consideration of axial effects. The program SILWER has proven to be applicable to MTR-cores for the investigation of axial effects. For routine use as for the support of reactor operation, this 3D code is a good supplement to the standard 2D model.

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

International Nuclear Information System (INIS)

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

2015-01-01

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. - Highlights: • U–Mo alloys in as-cast as well as in annealed conditions have been studied using Optical Microscope, SEM, XRD. • The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. • The dendritic microstructure of γ-(U,Mo) and B.C.C. ‘Mo’ phase of 33 at.% U–Mo alloy have been analysed. • Rietveld analysis has been done to optimize lattice parameters and calculate phase fractions in annealed alloys. • The Vickers microhardness of U_2Mo phase shows lower hardness than two phase microstructures in annealed alloys.

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

Energy Technology Data Exchange (ETDEWEB)

Chakraborty, S., E-mail: sibasis@barc.gov.in [Radiometallurgy Division, Bhabha Atomic Research Centre, Mumbai, 400085 (India); Choudhuri, G. [Atomic Fuels Division, Bhabha Atomic Research Centre, Mumbai, 400085 (India); Banerjee, J. [Radiometallurgy Division, Bhabha Atomic Research Centre, Mumbai, 400085 (India); Agarwal, Renu [Product Development Division, Bhabha Atomic Research Centre, Mumbai, 400085 (India); Khan, K.B.; Kumar, Arun [Radiometallurgy Division, Bhabha Atomic Research Centre, Mumbai, 400085 (India)

2015-12-15

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. - Highlights: • U–Mo alloys in as-cast as well as in annealed conditions have been studied using Optical Microscope, SEM, XRD. • The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. • The dendritic microstructure of γ-(U,Mo) and B.C.C. ‘Mo’ phase of 33 at.% U–Mo alloy have been analysed. • Rietveld analysis has been done to optimize lattice parameters and calculate phase fractions in annealed alloys. • The Vickers microhardness of U{sub 2}Mo phase shows lower hardness than two phase microstructures in annealed alloys.

Comparison between a commercial solid-phase extraction cartridge and a home-made silver containing charcoal column: purification of Mo-99 from I-131 and Te-121

International Nuclear Information System (INIS)

Dias, Carla Roberta; Teodoro, Rodrigo; Osso Junior, Joao

2011-01-01

Among the radioisotopes used for medical application in Nuclear Medicine, 99m Tc, readily available from the elution of 99 Mo/ 99m Tc generators, is the most used, responsible for more than eighty percent of the total applications. These generators use the 99 Mo radioisotope that is produced in nuclear reactors and IPEN imports all the 99 Mo used in Brazil, mainly from Canada (Nordion). Due to the increasing needs of the Nuclear Medicine in Brazil and the world shortage of 99 Mo observed since 2008, IPEN decided to construct a new research reactor named Brazilian Multipurpose Reactor (BMR) as well as to develop the production of 99 Mo through the route of 235 U fission using a CINTICHEM modified separation process. The 99 Mo obtained from this process contains some contaminants and need to be purified. The aim of this work is to compare the preliminary results of the purification step of the solution containing 99 Mo and the contaminants, 131 I and 121 Te in the silver containing charcoal column and a solid-phase extraction cartridge. The purification process of 99 Mo coming from fission LEU foils was performed by adsorption chromatography using a home-made activated charcoal containing silver column (AC-Ag) and a commercial solid-phase extraction cartridge (OnGuard II Ag). High yields of 99 Mo elution and high retention of 131 I were achieved in the AC-Ag column and silver cartridge but 121 Te was more retained in the cartridge than in the AC-Ag column. (author)

First-principles study of van der Waals interactions in MoS2 and MoO3

International Nuclear Information System (INIS)

Peelaers, H; Van de Walle, C G

2014-01-01

Van der Waals interactions play an important role in layered materials such as MoS 2 and MoO 3 . Within density functional theory, several methods have been developed to explicitly include van der Waals interactions. We compare the performance of several of these functionals in describing the structural and electronic properties of MoS 2 and MoO 3 . We include functionals based on the local density or generalized gradient approximations, but also based on hybrid functionals. The coupling of the semiempirical Grimme D2 method with the hybrid functional HSE06 is shown to lead to a very good description of both structural and electronic properties. (paper)

Adsorption of DNA/RNA nucleobases onto single-layer MoS2 and Li-Doped MoS2: A dispersion-corrected DFT study

Science.gov (United States)

Sadeghi, Meisam; Jahanshahi, Mohsen; Ghorbanzadeh, Morteza; Najafpour, Ghasem

2018-03-01

The kind of sensing platform in nano biosensor plays an important role in nucleic acid sequence detection. It has been demonstrated that graphene does not have an intrinsic band gap; therefore, transition metal dichalcogenides (TMDs) are desirable materials for electronic base detection. In the present work, a comparative study of the adsorption of the DNA/RNA nucleobases [Adenine (A), Cytosine (C) Guanine (G), Thymine (T) and Uracil (U)] onto the single-layer molybdenum disulfide (MoS2) and Li-doped MoS2 (Li-MoS2) as a sensing surfaces was investigated by using Dispersion-corrected Density Functional Theory (D-DFT) calculations and different measure of equilibrium distances, charge transfers and binding energies for the various nucleobases were calculated. The results revealed that the interactions between the nucleobases and the MoS2 can be strongly enhanced by introducing metal atom, due to significant charge transfer from the Li atom to the MoS2 when Lithium is placed on top of the MoS2. Furthermore, the binding energies of the five nucleobases were in the range of -0.734 to -0.816 eV for MoS2 and -1.47 to -1.80 eV for the Li-MoS2. Also, nucleobases were adsorbed onto MoS2 sheets via the van der Waals (vdW) force. This high affinity and the renewable properties of the biosensing platform demonstrated that Li-MoS2 nanosheet is biocompatible and suitable for nucleic acid analysis.

Creep Rupture Analysis and Life Estimation of 1.25Cr-0.5Mo, 2.25Cr-1Mo and Modified 9Cr-1Mo Steel: A Comparative Study

Science.gov (United States)

Roy, Prabir Kumar

2018-04-01

This paper highlights a comparative assessment of creep life of 1.25Cr-0.5Mo, 2.25Cr-1Mo and modified 9Cr-1Mo steels based on accelerated creep rupture tests. Creep rupture test data have been analysed and creep life of the above mentioned materials have been assessed using Larson Miller parameter at the stress levels of 60 and 42 MPa for different temperatures. Limiting steam temperatures for minimum design life of 105 h at 42 and 60 MPa for the above mentioned steels have also been calculated. Microstructural studies for the three above mentioned steels are also done.

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

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

Crystal structure and thermal stability of AgIn(MoO4)2

International Nuclear Information System (INIS)

Klevtsov, P.V.; Solodovnikov, S.F.; Perepelitsa, A.P.; Klevtsova, R.F.

1984-01-01

Tetragonal crystals of double molybdate AgIn(MoO 4 ) 2 are prepared bi crystallization from solution in Ag 2 Mo 2 O 7 melt (a=4.998, c=36.725 A, space group I4 1 , Z=6). Its crystal structure is determined (autodaffractometer ''Syntex P2 1 '', MoKsub(α)-radiation, 876 reflections, R=0.054) in which along with Mo-tetrahedrons Mo-octahedrons are present. By mutual edges latter are united into bands forming fragments of wolframite structure alonside with (In, Ag) octahedrons. In the direction of c axis wolframite fragments alternate with scheelite fragments consisting of Mo-tetrahedrons and Ag-octavertices. The crystallochemical formula of the compound is Ag(Insub(0.75)Agsub(0.25))sub(2)Mosub(2)Osub(8) [MoO 4 ]. At a temperature of about 600 deq C AgIn-molybdate transforms into modification with NaIn(MoO 4 ) 2 structure NaIn(MoO 4 ) 2 and melts at 650 deg C decomposing into In 2 (MoO 4 ) 3 solid phase and Ag 2 MoO 4 melt

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)

Preliminary Thermohydraulic Analysis of a New Moderated Reactor Utilizing an LEU-Fuel for Space Nuclear Thermal Propulsion

Energy Technology Data Exchange (ETDEWEB)

Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)

2015-10-15

The Korea Advanced NUclear Thermal Engine Rocket utilizing an LEU fuel (KANUTER-LEU) is a non-proliferative and comparably efficient NTR engine with relatively low thrust levels of 40 - 50 kN for in-space transportation. The small modular engine can expand mission versatility, when flexibly used in a clustered engine arrangement, so that it can perform various scale missions from low-thrust robotic science missions to high-thrust manned missions. In addition, the clustered engine system can enhance engine redundancy and ensuing crew safety as well as the thrust. The propulsion system is an energy conversion system to transform the thermal energy of the reactor into the kinetic energy of the propellant to produce the powers for thrust, propellant feeding and electricity. It is mainly made up of a propellant Feeding System (PFS) comprising a Turbo-Pump Assembly (TPA), a Regenerative Nozzle Assembly (RNA), etc. For this core design study, an expander cycle is assumed to be the propulsion system. The EGS converts the thermal energy of the EHTGR in the idle operation (only 350 kW{sub th} power) to electric power during the electric power mode. This paper presents a preliminary thermohydraulic design analysis to explore the design space for the new reactor and to estimate the referential engine performance. The new non-proliferative NTR engine concept, KANUTER-LEU, is under designing to surmount the nuclear proliferation obstacles on allR and Dactivities and eventual commercialization for future generations. To efficiently implement a heavy LEU fuel for the NTR engine, its reactor design innovatively possesses the key characteristics of the high U density fuel with high heating and H{sub 2} corrosion resistances, the thermal neutron spectrum core and also minimizing non-fission neutron loss, and the compact reactor design with protectively cooling capability. To investigate feasible design space for the moderated EHTGR-LEU and resultant engine performance, the

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.

Ternary system of Na2MoO4-Cs2MoO4-MoO3

International Nuclear Information System (INIS)

Zueva, V.P.; Shabanova, A.N.; Drobasheva, T.I.

1982-01-01

Using the methods of thermal analysis interaction of components in ternary system Na 2 MoO 4 -Cs 2 MoO 4 -MoO 3 has been studied. Crystallization surface consists of nine fields belonging to initial components and compounds of lateral sides. Triangulation of the system is carried out and the character of nonvariant points is clarified, the temperature of 360 deg C corresponds to low-melting eutectics

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-30

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-11-01

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

Stacking stability of MoS2 bilayer: An ab initio study

International Nuclear Information System (INIS)

Tao Peng; Guo Huai-Hong; Yang Teng; Zhang Zhi-Dong

2014-01-01

The study of the stacking stability of bilayer MoS 2 is essential since a bilayer has exhibited advantages over single layer MoS 2 in many aspects for nanoelectronic applications. We explored the relative stability, optimal sliding path between different stacking orders of bilayer MoS 2 , and (especially) the effect of inter-layer stress, by combining first-principles density functional total energy calculations and the climbing-image nudge-elastic-band (CI-NEB) method. Among five typical stacking orders, which can be categorized into two kinds (I: AA, AB and II: AA', AB', A'B), we found that stacking orders with Mo and S superposing from both layers, such as AA' and AB, is more stable than the others. With smaller computational efforts than potential energy profile searching, we can study the effect of inter-layer stress on the stacking stability. Under isobaric condition, the sliding barrier increases by a few eV/(ucGPa) from AA' to AB', compared to 0.1 eV/(ucGPa) from AB to [AB]. Moreover, we found that interlayer compressive stress can help enhance the transport properties of AA'. This study can help understand why inter-layer stress by dielectric gating materials can be an effective means to improving MoS 2 on nanoelectronic applications. (condensed matter: structural, mechanical, and thermal properties)

Sustainable one-step synthesis of hierarchical microspheres of PEGylated MoS2 nanosheets and MoO3 nanorods: Their cytotoxicity towards lung and breast cancer cells

International Nuclear Information System (INIS)

Kumar, Neeraj; George, Blassan Plackal Adimuriyil; Abrahamse, Heidi; Parashar, Vyom; Ngila, Jane Catherine

2017-01-01

Highlights: • Microspheres of PEGylated MoS 2 nanosheets were synthesised by hydrothermal route. • PEGylated MoS 2 have shown good cytotoxicity towards breast cancer (MCF-7) cells. • For comparison, h-MoO 3 nanorods were prepared by simple chemical route. • h-MoO 3 have exhibited excellent cytotoxicity towards lung (A549) cancer cells. - Abstract: Nanotechnology provides an emerging potent alternate mode of cancer therapy. Nanomaterials dispersion or solubility is of particular concern in utilising their full potential applications in biomedical fields. PEGylation of nanomaterials is considered to provide products with stealth properties, and physiological environment with no obvious adverse effects. The purpose of this work was to develop a sustainable one-step method for fabrication of hierarchical microspheres of PEGylated MoS 2 nanosheets using a stoichiometric ratio of Mo(VI) and thiourea. This study further investigated the cytotoxicity of the PEGylated MoS 2 nanosheets towards lung (A549) and breast cancer (MCF-7) cell lines by analysing morphological changes and performing dose-dependent cell proliferation, and cytotoxicity analysis using adenosine 5′-triphosphate (ATP), and lactate dehydrogenase (LDH) assay. For comparison, MoO 3 nanorods were synthesised by simple chemical route and their cytotoxicity towards lung (A549) and breast cancer (MCF-7) cell lines were checked. The findings suggested that PEGylated MoS 2 nanosheets have excellent cytotoxicity towards breast cancer (MCF-7) cell lines, and MoO 3 have better cytotoxicity towards lung (A549) cancer cell lines. This work envisages an accessible foundation for engineering sophisticated biomolecule–MoS 2 nanosheets conjugation due to the defect-rich biocompatible surface, to achieve great versatility, additional functions, and further advances in the biomedical field.

Progress of the RERTR programme in 1999

International Nuclear Information System (INIS)

Travelli, A.

1999-01-01

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during 1999 and discusses planned activities for the coming year. The past year was characterized by exceptionally important accomplishments and events for the RERTR program. - Three additional shipments containing 1,006 spent fuel assemblies from foreign research reactors were accepted by the U.S. Altogether, 3,237 spent fuel assemblies from foreign research reactors have been received by the U.S. under the acceptance policy. - Postirradiation examinations of the first two batches of microplates revealed good irradiation behavior of U-6Mo, and excellent irradiation behavior of U-Mo alloys with higher Mo content or with small Ru additions. Irradiation of a new batch of microplates to investigate the behavior of these fuels at high temperatures is scheduled to begin in October 1999. 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. - Progress on irradiation testing and safety analyses was made in the Russian RERTR program, which aims to develop and demonstrate the technical means needed to convert Russian supplied research reactors to LEU fuels. - The U.S. Government has decided to aggressively pursue, in cooperation with the Russian Government, eventual conversion of three Russian plutonium production reactors to the use of low-enriched UO 2 -Al dispersion fuel. This effort is now proceeding, with assistance from RERTR personnel. - At the request of the German government, the RERTR program has addressed the performance of a new alternative LEU FRM-II core design that could be installed in the same building structure erected for the current 20 MW HEU design with the same 50-day fuel life. The results have been favorable. - Significant improvements were made in the design of an LEU metal-foil target

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.

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

International Nuclear Information System (INIS)

Westmacott, Chad; Cameron, Ron

2012-01-01

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

Homogeneous SLOWPOKE reactors for Mo-99/Tc-99m production in North America

Energy Technology Data Exchange (ETDEWEB)

Hilborn, J.W., E-mail: hilbovanw@sympatico.ca [Deep River, Ontario (Canada); Bonin, H.W. [Royal Military College of Canada, Kingston, Ontario (Canada)

2014-07-01

The 15 month shutdown of NRU in 2009 - 2010 caused an overall isotope shortage of approximately 30%; and in North America, the annual Tc-99m demand decreased from an estimated 20 million unit doses to about 15 million unit doses. Mo-99/Tc-99m is produced from HEU targets, irradiated in NRU for 11 days, and after chemical removal of uranium it is shipped to Nordion in Kanata, Ontario. Nordion further purifies the material and sends it to Lantheus Medical Imaging in the USA for manufacture of Mo-99 generators, which are then distributed to hundreds of hospital radiopharmacies throughout North America. One other American company, Covidien, manufactures and distributes Mo-99 generators like Lantheus, but they import bulk Mo-99 from Europe or South Africa. At the hospitals, Tc-99m is chemically extracted daily from the Mo-99 generators and loaded into syringes for immediate clinical use. Fortuitously, the 66 hour half-life of Mo-99 allows the replenishment of Tc-99m in the generator over a growth period of about 20 hours; and a generator can be 'milked' daily for up to two weeks. A more efficient model is the direct production and distribution of Tc-99m unit doses to regional hospitals from 10 'industrial' radiopharmacies located at existing licensed reactor sites in North America. A 20 kW homogeneous SLOWPOKE reactor at each site would deliver 15 litres of irradiated uranyl sulphate fuel solution daily to industrial-scale hot cells for extraction of Mo-99, which would be incorporated in large Mo-99/Tc-99m generators for extraction of Tc-99m five days a week; and the Low Enriched Uranium (LEU) would be recycled. Each automated hot-cell facility would be designed to load up to 7,000 Tc-99m syringes daily, for courier delivery to all of the Nuclear Medicine hospitals within a 3 hour average range by road transport. Typically, the delivered doses would be in the range 10 to 30 mCi. Assuming an average unit dose of 25 mCi at the hospital and 5 x 52

U-10Mo Baseline Fuel Fabrication Process Description

Energy Technology Data Exchange (ETDEWEB)

Hubbard, Lance R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Arendt, Christina L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Dye, Daniel F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Clayton, Christopher K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lerchen, Megan E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lombardo, Nicholas J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zacher, Alan H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-09-27

This document provides a description of the U.S. High Power Research Reactor (USHPRR) low-enriched uranium (LEU) fuel fabrication process. This document is intended to be used in conjunction with the baseline process flow diagram (PFD) presented in Appendix A. The baseline PFD is used to document the fabrication process, communicate gaps in technology or manufacturing capabilities, convey alternatives under consideration, and as the basis for a dynamic simulation model of the fabrication process. The simulation model allows for the assessment of production rates, costs, and manufacturing requirements (manpower, fabrication space, numbers and types of equipment, etc.) throughout the lifecycle of the USHPRR program. This document, along with the accompanying PFD, is updated regularly

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.

Status and progress of the RERTR program in the year 2000

International Nuclear Information System (INIS)

Travelli, Armando

2000-01-01

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the year 2000 and discusses the main activities planned for the year 2001. The past year was held by important accomplishments and events for the RERTR program: Four additional shipments containing 503 spent fuel assemblies from foreign research reactors were accepted by the U.S. Altogether, 3,740 spent fuel assemblies from foreign research reactors have been received by the U.S. under the acceptance policy; Postirradiation examinations of three batches 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 miniplates of greater sizes is in progress 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 is proceeding on schedule. Test fuel elements with uranium density of 6 g/cm 3 are being fabricated by BWXT and are scheduled to begin undergoing irradiation in the HFR-Petten in the spring of 2001, with a goal of qualifying this fuel by the end of 2003. U-Mo fuel with uranium density of 8-9 g/cm 3 is planned to be qualified by the end of 2005; Joint LEU conversion feasibility studies were completed for HFR-Petten and for SAFARI-1; Significant improvements were made in the design of LEU metal-foil annular targets that would allow efficient production of fission 99 Mo. Irradiations in the RAS-GAS reactor showed that these targets can formed from aluminum tubes, and that the yield and purity of their product from the acidic process were at least as good as those from the HEU Cintichem targets; Progress was made on irradiation testing of

Status and progress of the RERTR program in the year 2000

International Nuclear Information System (INIS)

Travelli, A.

2000-01-01

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the year 2000 and discusses the main activities planned for the year 2001. The past year was characterized by important accomplishments and events for the RERTR program. Four additional shipments containing 503 spent fuel assemblies from foreign research reactors were accepted by the U.S. Altogether, 3,740 spent fuel assemblies from foreign research reactors have been received by the U.S. under the acceptance policy. Postirradiation examinations of three batches of microplates have continued to reveal excellent irradiation behavior of U-MO dispersion fuels in a variety of compositions and irradiating conditions. h-radiation of two new batches of miniplates of greater sizes is in progress in the ATR to investigate me 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 is proceeding on schedule. Test fuel elements with 6 gU/cm 3 are being fabricated by BWXT and are scheduled to begin undergoing irradiation in the HFR-Petten in the spring of 2001, with a goal of qualifying this fuel by the end of 2003. U-Mo with 8-9 gU/cm 3 is planned to be qualified by the end of 2005. Joint LEU conversion feasibility studies were completed for HFR-Petten and for SAFARI-1. Significant improvements were made in the design of LEU metal-foil annular targets that would allow efficient production of fission 99 Mo. Irradiations in the RAS-GAS reactor showed that these targets can formed from aluminum tubes, and that the yield and purity of their product from the acidic process were at least as good as those from the HEU Cintichem targets. Progress was made on irradiation testing of LEU UO 2 dispersion fuel and on

Study of the 93Mo nucleus

International Nuclear Information System (INIS)

Mittal, V.K.; Avasthi, D.K.; Kumar, A; Govil, I.M.

1981-01-01

The level structure of 93 Mo has been studied through 93 Nb(p,nγ) reaction at incident proton energy of 4.2 MeV. Spin assignments and multipole mixing ratios for various levels and transitions have been deduced from the analysis of angular distributions of singles γ-ray spectra. Lifetimes of various levels have been obtained using Doppler Shift Attenuation (DSA) method. (author)

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)

99Mo production by 100Mo(n,2n)99Mo using accelerator neutrons

International Nuclear Information System (INIS)

Sato, Nozomi; Kawabata, Masako; Nagai, Yasuki; Hashimoto, Kazuyuki; Hatsukawa, Yuichi; Saeki, Hideya; Motoishi, Shoji; Kin, Tadahiro; Konno, Chikara; Ochiai, Kentaro; Takakura, Kosuke; Minato, Futoshi; Iwamoto, Osamu; Iwamoto, Nobuyuki; Hashimoto, Shintaro

2013-01-01

We proposed a new route to produce a medical radioisotope 99 Mo by the 100 Mo(n,2n) 99 Mo reaction using accelerator neutrons. A high-quality 99 Mo with a minimum level of radioactive waste can be obtained by the proposed reaction. The decay product of 99 Mo, 99m Tc, is separated from 99 Mo by the sublimation method. The proposed route could bring a major breakthrough in the solution of ensuring a constant and reliable supply of 99 Mo. (author)

Influence of chemical composition in crystallographic texture Fe-Cr-Mo alloys

International Nuclear Information System (INIS)

Moura, L.B.; Guimaraes, R.F.

2010-01-01

The use of steels with higher contents of Mo in the oil industry has been an alternative to reduce the effect of naphthenic corrosion in refining units. The addition of Mo in Fe-Cr alloys in the same manner that increases resistance to corrosion naphthenic causes some difficulties such as difficulty of forming, welding and embrittlement. In this work, experimental ingots of Fe-Cr-Mo alloys (Cr - 9, 15 and 17%, Mo - 5, 7 and 9%) were melted in vacuum induction furnace and hot and cold rolled in a laboratory rolling mill. The influence of chemical composition on crystallographic texture of samples subjected to the same thermo-mechanical treatment was analyzed by x-ray diffraction. The results indicate that fiber (111) becomes more intense with increasing Mo and/or Cr contents. (author)

Altered [99mTc]Tc-MDP biodistribution from neutron activation sourced 99Mo.

Science.gov (United States)

Demeter, Sandor; Szweda, Roman; Patterson, Judy; Grigoryan, Marine

2018-01-01

Given potential worldwide shortages of fission sourced 99 Mo/ 99m Tc medical isotopes there is increasing interest in alternate production strategies. A neutron activated 99 Mo source was utilized in a single center phase III open label study comparing 99m Tc, as 99m Tc Methylene Diphosphonate ([ 99m Tc]Tc-MDP), obtained from solvent generator separation of neutron activation produced 99 Mo, versus nuclear reactor produced 99 Mo (e.g., fission sourced) in oncology patients for which an [ 99m Tc]Tc-MDP bone scan would normally have been indicated. Despite the investigational [ 99m Tc]Tc-MDP passing all standard, and above standard of care, quality assurance tests, which would normally be sufficient to allow human administration, there was altered biodistribution which could lead to erroneous clinical interpretation. The cause of the altered biodistribution remains unknown and requires further research.

Wear behaviour of wear-resistant adaptive nano-multilayered Ti-Al-Mo-N coatings

Science.gov (United States)

Sergevnin, V. S.; Blinkov, I. V.; Volkhonskii, A. O.; Belov, D. S.; Kuznetsov, D. V.; Gorshenkov, M. V.; Skryleva, E. A.

2016-12-01

Coating samples in the Ti-Al-Mo-N system were obtained by arc-PVD method at variable bias voltage Ub applied to the substrate, and the partial pressure of nitrogen P(N2) used as a reaction gas. The deposited coatings were characterized by a nanocrystalline structure with an average grain size of 30-40 nm and multilayered architecture with alternating layers of (Ti,Al)N nitride and Mo-containing phases with a thickness comparable to the grain size. Coatings of (Ti,Al)N-Mo-Mo2N and (Ti,Al)N-Mo2N compositions were obtained by changing deposition parameters. The obtained coatings had hardness of 40 GPa and the relative plastic deformation under microindentation up to 60%. (Ti,Al)N-Mo2N coatings demonstrated better physicomechanical characteristics, showing high resistance to crack formation and destruction through the plastic deformation mechanism without brittle fracturing, unlike (Ti,Al)N-Mo-Mo2N. The friction coefficient of the study coatings (against Al2O3 balls under dry condition using a pin-on-disc method) reached the values of 0.35 and 0.5 at 20 °C and 500 °C respectively, without noticeable wear within this temperature range. These tribological properties were achieved by forming MoO3 acting as a solid lubricant. At higher temperatures the deterioration in the tribological properties is due to the high rate of MoO3 sublimation from friction surfaces.

Phase formation in the Li2MoO4–K2MoO4–In2(MoO4)3 system and crystal structures of new compounds K3InMo4O15 and LiK2In(MoO4)3

International Nuclear Information System (INIS)

Khal’baeva, Klara M.; Solodovnikov, Sergey F.; Khaikina, Elena G.; Kadyrova, Yuliya M.; Solodovnikova, Zoya A.; Basovich, Olga M.

2012-01-01

XRD study of solid-phase interaction in the Li 2 MoO 4 –K 2 MoO 4 –In 2 (MoO 4 ) 3 system was performed. The boundary K 2 MoO 4 –In 2 (MoO 4 ) 3 system is an non-quasibinary join of the K 2 O–In 2 O 3 –MoO 3 system where a new polymolybdate K 3 InMo 4 O 15 isotypic to K 3 FeMo 4 O 15 was found. In the structure (a=33.2905(8), b=5.8610(1), c=15.8967(4) Å, β=90.725(1)°, sp. gr. C2/c, Z=8, R(F)=0.0407), InO 6 octahedra, Mo 2 O 7 diortho groups and MoO 4 tetrahedra form infinite ribbons {[In(MoO 4 ) 2 (Mo 2 O 7 )] 3− } ∞ along the b-axis. Between the chains, 8- to 10-coordinate potassium cations are located. A subsolidus phase diagram of the Li 2 MoO 4 –K 2 MoO 4 –In 2 (MoO 4 ) 3 system was constructed and a novel triple molybdate LiK 2 In(MoO 4 ) 3 was revealed. Its crystal structure (a=7.0087(2), b=9.2269(3), c=10.1289(3) Å, β=107.401(1)°, sp. gr. P2 1 , Z=2, R(F)=0.0280) contains an open framework of vertex-shared MoO 4 tetrahedra, InO 6 octahedra and LiO 5 tetragonal pyramids with nine- and seven-coordinate potassium ions in the framework channels. - Graphical abstract: Exploring the Li 2 MoO 4 –K 2 MoO 4 –In 2 (MoO 4 ) 3 system showed its partial non-quasibinarity and revealed new compounds K 3 InMo 4 O 15 (isotypic to K 3 FeMo 4 O 15 ) and LiK 2 In(MoO 4 ) 3 which were structurally studied. An open framework of the latter is formed by vertex-shared MoO 4 tetrahedra, InO 6 octahedra and LiO 5 tetragonal pyramids. Highlights: ► Subsolidus phase relations in the Li 2 MoO 4 –K 2 MoO 4 –In 2 (MoO 4 ) 3 system were explored. ► The K 2 MoO 4 –In 2 (MoO 4 ) 3 system is a non-quasibinary join of the K 2 O–In 2 O 3 –MoO 3 system. ► New compounds K 3 InMo 4 O 15 and LiK 2 In(MoO 4 ) 3 were obtained and structurally studied. ► K 3 InMo 4 O 15 is isotypic to K 3 FeMo 4 O 15 and carries bands of InO 6 , MoO 4 and Mo 2 O 7 units. ► An open framework of LiK 2 In(MoO 4 ) 3 is formed by polyhedra MoO 4 , InO 6 and LiO 5 .

Conversion and start up of Tehran Research Reactor with LEU fuel

International Nuclear Information System (INIS)

Zaker, M.

2004-01-01

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

Structural studies of amorphous Mo-Ge alloys using synchrotron radiation

International Nuclear Information System (INIS)

Kortright, J.B.

1984-06-01

Structural changes in sputtered amorphous Mo-Ge alloy films with composition varying from a-Ge to about 70 at. % Mo have been studied with several x-ray techniques. Results of individual techniques are presented and discussed in separate chapters. The complementary nature of information obtained from EXAFS and scattering for these materials is discussed in a separate chapter. A concluding chapter summarizes the results and structural changes with composition

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-09-30

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

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

Phase relations in the M2MoO4 - Ag2MoO4 - Hf(MoO4)2 (M=Li, Na) systems

International Nuclear Information System (INIS)

Bazarova, Zh.G.; Bazarov, B.G.; Balsanova, L.V.

2002-01-01

The M 2 MoO 4 - Ag 2 MoO 4 - Hf(MoO 4 ) 2 (M=Li, Na) systems were studied by X-ray diffraction and differential thermal analyses in the subsolidus area (450 - 500 Deg C) for the first time. The formation of the binary compound with the variable composition Li 4-x Hf 1+0.2x (MoO 4 ) 4 (0 ≤ x ≤ 0.6) in the Li 2 MoO 4 - Hf(MoO 4 ) 2 system and the ternary molybdates Li 4 Ag 2 Hf(MoO 4 ) 5 (S 1 ) and Na 2 Ag 2 Hf(MoO 4 ) 4 (S 2 ) was established and the thermal characteristics of the prepared compounds were examined. The new binary molybdate Ag 2 Hf(MoO 4 ) 3 was prepared by the reaction between Ag 2 MoO 4 and Hf(MoO 4 ) 2 [ru

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)

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

International Nuclear Information System (INIS)

Ofori, Y. T.

2013-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-01

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

Electronic structure of structural open derivatives of the [Mo6X14]2- cluster: [Mo5Cl13]2- and [Mo4I11]2-

International Nuclear Information System (INIS)

Miessner, H.; Korol'kov, D.V.

1983-01-01

The electronic structure of structural open derivatives of the [Mo 6 X 14 ] 2 - -cluster [Mo 5 Cl 13 ] 2 - and [Mo 4 I 11 ] 2 - has been studied by the EHMO method. In [Mo 5 Cl 13 ] 2 - 9 occupied MO's with dominant Mo4d character are responsible for the formation of the 8 metal-metal bonds. In [Mo 4 I 11 ] 2 - the stronger covalent character of the Mo-I bonds affects the localization and the energy of molecular orbitals and also the charge distribution. The metal-metal bonds are formed by 8 MO's containing considerable participation of halogen AO's contrary to the chloride cluster. There is no bonding between the Mo atoms at the wing tips of the Mo 4 butterfly and the reason for decreasing the dihedral angle between the Mo 3 planes in [Mo 4 I 11 ] 2 - compared with the octahedral angle is apparently the stabilization of the whole system (Mo-Mo and Mo-I bonds). The unpaired electron occupies in both clusters a slightly antibonding (with regard to the Mo-Mo bonds) orbital. (author)

Leu72Met408 Polymorphism of the Ghrelin Gene Is Associated With Early Phase of Gastric Emptying in the Patients With Functional Dyspepsia in Japan.

Science.gov (United States)

Yamawaki, Hiroshi; Futagami, Seiji; Shimpuku, Mayumi; Shindo, Tomotaka; Maruki, Yuuta; Nagoya, Hiroyuki; Kodaka, Yasuhiro; Sato, Hitomi; Gudis, Katya; Kawagoe, Tetsuro; Sakamoto, Choitsu

2015-01-01

There are no available data about the relationship between ghrelin gene genotypes and early phase of gastric emptying in functional dyspepsia (FD) as defined by Rome III classification. We enrolled 74 patients presenting with typical symptoms of FD and 64 healthy volunteers. Gastric motility was evaluated using the 13C-acetate breath test. We used Rome III criteria to evaluate upper abdominal symptoms and self-rating questionnaires for depression (SRQ-D) scores to determine status of depression. The Arg51Gln (346G->A), preproghrelin (3056T->C), Leu72Met (408C->A), Gln90Leu (3412T->A) and G-protein 3 (825C->T) polymorphisms were analyzed in the DNA from blood samples of enrolled subjects. Genotyping was performed by polymerase chain reaction. There was a significant relationship between the Gln90Leu3412 genotype and SRQ-D score in FD patients (P = 0.009). Area under the curve at 15 minutes (AUC15) value was significantly associated with the Leu72Met408 genotype (P = 0.015) but not with entire gastric emptying. The Leu72Met (408C->A) single nucleotide polymorphism was significantly associated with early phase of gastric emptying in FD patients. Further studies will be necessary to clarify the association between ghrelin gene single nucleotide polymorphisms and early phase of gastric emptying in FD patients.

First-principle study of hydrogenation on monolayer MoS2

International Nuclear Information System (INIS)

Xu, Yong; Li, Yin; Chen, Xi; Zhang, Ru; Zhang, Chunfang; Lu, Pengfei

2016-01-01

The structural and electronic properties of hydrogenation on 1H-MoS 2 and 1T-MoS 2 have been systematically explored by using density functional theory (DFT) calculations. Our calculated results indicate an energetically favorable chemical interaction between H and MoS 2 monolayer for H adsorption when increasing concentration of H atoms. For 1H-MoS 2 , single H atom adsorption creates midgap approaching the Fermi level which increases the n-type carrier concentration effectively. As a consequence, its electrical conductivity is expected to increase significantly. For 1T-MoS 2 , H atoms adsorption can lead to the opening of a direct gap of 0.13 eV compared to the metallic pristine 1T-MoS 2 .

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)

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.

Composition-Graded MoWSx Hybrids with Tailored Catalytic Activity by Bipolar Electrochemistry.

Science.gov (United States)

Tan, Shu Min; Pumera, Martin

2017-12-06

Among transition metal dichalcogenide (TMD)-based composites, TMD/graphene-related material and bichalcogen TMD composites have been widely studied for application toward energy production via the hydrogen evolution reaction (HER). However, scarcely any literature explored the possibility of bimetallic TMD hybrids as HER electrocatalysts. The use of harmful chemicals and harsh preparation conditions in conventional syntheses also detracts from the objective of sustainable energy production. Herein, we present the conservational alternative synthesis of MoWS x via one-step bipolar electrochemical deposition. Through bipolar electrochemistry, the simultaneous fabrication of composition-graded MoWS x hybrids, i.e., sulfur-deficient Mo x W (1-x) S 2 and Mo x W (1-x) S 3 (MoWS x /BPE cathodic and MoWS x /BPE anodic , respectively) under cathodic and anodic overpotentials, was achieved. The best-performing MoWS x /BPE cathodic and MoWS x /BPE anodic materials exhibited Tafel slopes of 45.7 and 50.5 mV dec -1 , together with corresponding HER overpotentials of 315 and 278 mV at -10 mA cm -2 . The remarkable HER activities of the composite materials were attributed to their small particle sizes, as well as the near-unity value of their surface Mo/W ratios, which resulted in increased exposed HER-active sites and differing active sites for the concurrent adsorption of protons and desorption of hydrogen gas. The excellent electrocatalytic performances achieved via the novel methodology adopted here encourage the empowerment of electrochemical deposition as the foremost fabrication approach toward functional electrocatalysts for sustainable energy generation.

Edge termination of MoS2 and CoMoS catalyst particles

DEFF Research Database (Denmark)

Byskov, Line Sjolte; Nørskov, Jens Kehlet; Clausen, B. S.

2000-01-01

The edge termination of MoS2 and CoMoS catalyst particles is studied by density functional calculations. We show that for structures without vacancies Mo-terminated edges have the lowest edge energies. Creation of vacancies, which are believed to be active sites in these catalyst systems, leads...

Sustainable one-step synthesis of hierarchical microspheres of PEGylated MoS{sub 2} nanosheets and MoO{sub 3} nanorods: Their cytotoxicity towards lung and breast cancer cells

Energy Technology Data Exchange (ETDEWEB)

Kumar, Neeraj [Department of Applied Chemistry, University of Johannesburg, Doornfontein 2028, South Africa, (South Africa); George, Blassan Plackal Adimuriyil; Abrahamse, Heidi [Laser Research Centre, Faculty of Health Sciences, University of Johannesburg, Doornfontein 2028 (South Africa); Parashar, Vyom, E-mail: vyomparashar@gmail.com [Department of Applied Chemistry, University of Johannesburg, Doornfontein 2028, South Africa, (South Africa); Ngila, Jane Catherine, E-mail: jcngila@uj.ac.za [Department of Applied Chemistry, University of Johannesburg, Doornfontein 2028, South Africa, (South Africa)

2017-02-28

Highlights: • Microspheres of PEGylated MoS{sub 2} nanosheets were synthesised by hydrothermal route. • PEGylated MoS{sub 2} have shown good cytotoxicity towards breast cancer (MCF-7) cells. • For comparison, h-MoO{sub 3} nanorods were prepared by simple chemical route. • h-MoO{sub 3} have exhibited excellent cytotoxicity towards lung (A549) cancer cells. - Abstract: Nanotechnology provides an emerging potent alternate mode of cancer therapy. Nanomaterials dispersion or solubility is of particular concern in utilising their full potential applications in biomedical fields. PEGylation of nanomaterials is considered to provide products with stealth properties, and physiological environment with no obvious adverse effects. The purpose of this work was to develop a sustainable one-step method for fabrication of hierarchical microspheres of PEGylated MoS{sub 2} nanosheets using a stoichiometric ratio of Mo(VI) and thiourea. This study further investigated the cytotoxicity of the PEGylated MoS{sub 2} nanosheets towards lung (A549) and breast cancer (MCF-7) cell lines by analysing morphological changes and performing dose-dependent cell proliferation, and cytotoxicity analysis using adenosine 5′-triphosphate (ATP), and lactate dehydrogenase (LDH) assay. For comparison, MoO{sub 3} nanorods were synthesised by simple chemical route and their cytotoxicity towards lung (A549) and breast cancer (MCF-7) cell lines were checked. The findings suggested that PEGylated MoS{sub 2} nanosheets have excellent cytotoxicity towards breast cancer (MCF-7) cell lines, and MoO{sub 3} have better cytotoxicity towards lung (A549) cancer cell lines. This work envisages an accessible foundation for engineering sophisticated biomolecule–MoS{sub 2} nanosheets conjugation due to the defect-rich biocompatible surface, to achieve great versatility, additional functions, and further advances in the biomedical field.

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.

Interface morphology of Mo/Si multilayer systems with varying Mo layer thickness studied by EUV diffuse scattering.

Science.gov (United States)

Haase, Anton; Soltwisch, Victor; Braun, Stefan; Laubis, Christian; Scholze, Frank

2017-06-26

We investigate the influence of the Mo-layer thickness on the EUV reflectance of Mo/Si mirrors with a set of unpolished and interface-polished Mo/Si/C multilayer mirrors. The Mo-layer thickness is varied in the range from 1.7 nm to 3.05 nm. We use a novel combination of specular and diffuse intensity measurements to determine the interface roughness throughout the multilayer stack and do not rely on scanning probe measurements at the surface only. The combination of EUV and X-ray reflectivity measurements and near-normal incidence EUV diffuse scattering allows to reconstruct the Mo layer thicknesses and to determine the interface roughness power spectral density. The data analysis is conducted by applying a matrix method for the specular reflection and the distorted-wave Born approximation for diffuse scattering. We introduce the Markov-chain Monte Carlo method into the field in order to determine the respective confidence intervals for all reconstructed parameters. We unambiguously detect a threshold thickness for Mo in both sample sets where the specular reflectance goes through a local minimum correlated with a distinct increase in diffuse scatter. We attribute that to the known appearance of an amorphous-to-crystallization transition at a certain thickness threshold which is altered in our sample system by the polishing.

Preparation of MoB and MoB-MoSi2 composites by combustion synthesis in SHS mode

International Nuclear Information System (INIS)

Yeh, C.L.; Hsu, W.S.

2007-01-01

Combustion synthesis in the mode of self-propagating high-temperature synthesis (SHS) was carried out in the Mo-B and Mo-B-Si systems for the preparation of molybdenum boride MoB and the composite of MoB-MoSi 2 from elemental powder compacts. Under a preheating temperature above 150 deg. C , the reaction of Mo with boron in the sample compact of Mo:B = 1:1 is characterized by a planar combustion front propagating in a self-sustaining and steady manner. As the preheating temperature or sample compaction density increased, combustion temperature was found to increase and the propagation rate of the combustion front was correspondingly enhanced. Moreover, the XRD analysis provides evidence of yielding nearly single-phase α-MoB from the Mo-B sample at equiatomic stoichiometry. In the synthesis of MoB-MoSi 2 composites, the starting stoichiometry of the Mo-B-Si powder compact was varied so as to produce the final composites containing 20-80 mol% MoB. It was also found the increase of flame-front velocity and combustion temperature with increasing MoB content formed in the composite. The composition analysis by XRD shows excellent conversion from the Mo-B-Si powder compact to the MoB-MoSi 2 composite through the SHS reaction; that is, in addition to a small amount of Mo 5 Si 3 , the as-synthesized composite is composed entirely of MoB and MoSi 2

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Studies of techniques for the post-elution concentration of 99mTc obtained from gel type 99Mo/99mTc generators

International Nuclear Information System (INIS)

Suzuki, Katia Noriko

2009-01-01

On average 80% of the radiopharmaceuticals used in Nuclear Medicine are labeled with 99 mTc due to its physical properties and easy attainment through of 99 Mo/ 99 mTc generators. The Directory of Radiopharmacy (DIRF) of IPEN-CNEN/SP developed a gel type chromatographic generator of MoZr with 99 Mo produced by 98 Mo(n,γ) 99 Mo reaction that occurs at the IEA-R1 Nuclear Reactor. The gel is composed of zirconium molybdate with elution volume of 12 mL with an activity of 11100 MBq (300 mCi) producing a radioactive concentration of 925 MBq (25 mCi)/mL. The fission generator gives a higher radioactive concentration around 1850 MBq (50 mCi)/mL. The aim of this work is to study a system of post-elution concentration of 99 mTc for the attainment of a high enough radioactive concentration to meet the demands of the market, with a proved quality. Two types of systems of post-elution concentration were developed: the single and the tandem. The most appropriate system for the gel generator of 99 Mo/ 99 mTc, being at the same time sterile and vacuum automated, was the tandem system using Dionex 2.5 cc/QMA cartridges. The gel generator is eluted with 10 mL of solution of 0.1% NaCl and the pertechnetate anion is retained in the QMA cartridge and further eluted with 4 mL of saline. The process takes no more than 30 minutes. The elution efficiency of the system of concentration was 90 %. At the beginning of 2009 a global crisis in the supply of 99 Mo took place making it necessary the development of alternative technologies for the production of 99 Mo/ 99 mTc generators using fission produced 99 Mo and the development of an appropriate method to extend the useful life of this generator. The results of this study showed that the same system developed for the post- concentration of the gel generator can be employed for the fission generator, using the tandem system, giving a concentration factor of 3 for the elution of 99 mTc. (author)

Adsorption studies of alcohol molecules on monolayer MoS_2 nanosheet—A first-principles insights

International Nuclear Information System (INIS)

Nagarajan, V.; Chandiramouli, R.

2017-01-01

Highlights: • The adsorption of methanol, ethanol & 1-propanol on MoS_2 nanosheet are studied. • The PDOS & band structure confirms adsorption of alcohol vapors on MoS_2 nanosheet. • The adsorption of 1-propanol vapor on MoS_2 nanosheet is more favorable. • The alcohol molecules adsorption on MoS_2 nanosheet is explored in atomistic level. - Abstract: The electronic and adsorption properties of three different alcohol molecules namely methanol, ethanol and 1-propanol vapors on MoS_2 nanosheet is investigated using DFT method. The structural stability of MoS_2 nanosheet is ascertained with formation energy. The adsorption properties of alcohol molecules on MoS_2 base material is discussed in terms of average energy gap variation, Mulliken charge transfer, energy band gap and adsorption energy. The prominent adsorption sites of methanol, ethanol and 1-propanol vapors on MoS_2 nanosheet are studied in atomistic level. The projected density of states (PDOS) spectrum gives the clear insights on the electronic properties of MoS_2 nanosheet. The PDOS and energy band structure confirmed the adsorption of alcohol vapors on MoS_2 nanosheet. The variation in the band structure and PDOS is noticed upon adsorption of methanol, ethanol and 1-propanol molecules on MoS_2 nanosheet. The PDOS spectrum also reveals the variation in peak maxima owing to transfer of electron between alcohol molecules and MoS_2 base material. The adsorption of 1-propanol vapor on MoS_2 nanosheet is observed to be more favorable than other alcohol molecules. The findings confirm that monolayer MoS_2 nanosheet can be used to detect the presence of alcohol vapors in the environment.

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

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.)

Possible doping strategies for MoS 2 monolayers: An ab initio study

KAUST Repository

Dolui, Kapildeb

2013-08-14

Density functional theory is used to systematically study the electronic properties of doped MoS2 monolayers, where the dopants are incorporated both via S/Mo substitution or as adsorbates. Among the possible substitutional dopants at the Mo site, Nb is identified as suitable p-type dopant, while Re is the donor with the lowest activation energy. When dopants are simply adsorbed on a monolayer we find that alkali metals shift the Fermi energy into the MoS2 conduction band, making the system n type. Finally, the adsorption of charged molecules is considered, mimicking an ionic liquid environment. We find that molecules adsorption can lead to both n- and p-type conductivity, depending on the charge polarity of the adsorbed species. © 2013 American Physical Society.

Possible doping strategies for MoS 2 monolayers: An ab initio study

KAUST Repository

Dolui, Kapildeb; Rungger, Ivan; Das Pemmaraju, Chaitanya; Sanvito, Stefano

2013-01-01

Density functional theory is used to systematically study the electronic properties of doped MoS2 monolayers, where the dopants are incorporated both via S/Mo substitution or as adsorbates. Among the possible substitutional dopants at the Mo site, Nb is identified as suitable p-type dopant, while Re is the donor with the lowest activation energy. When dopants are simply adsorbed on a monolayer we find that alkali metals shift the Fermi energy into the MoS2 conduction band, making the system n type. Finally, the adsorption of charged molecules is considered, mimicking an ionic liquid environment. We find that molecules adsorption can lead to both n- and p-type conductivity, depending on the charge polarity of the adsorbed species. © 2013 American Physical Society.

Study of the water-gas shift reaction on Mo2C/Mo catalytic coatings for application in microstructured fuel processors

NARCIS (Netherlands)

Rebrov, E.V.; Kuznetsov, S.A.; Croon, de M.H.J.M.; Schouten, J.C.

2007-01-01

The activity and stability of two types of molybdenum carbide coatings deposited on molybdenum substrates (Mo2C/Mo) were compared in the water-gas shift reaction at 513–631 K. The activity of the Mo2C/Mo coatings obtained by carburization of preoxidized molybdenum substrates in a CH4/H2 mixture at

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

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

Phase formation in the Li2MoO4-Rb2MoO4-Ln2(MoO4)3 systems and the properties of LiRbLn2(MoO4)4

International Nuclear Information System (INIS)

Basovich, O.M.; Khajkina, E.G.; Vasil'ev, E.V.; Frolov, A.M.

1995-01-01

Phase equilibria within subsolidus range of ternary salt systems Li 2 MoO 4 -Rb 2 MoO 4 -Ln 2 (MoO 4 ) 4 (Ln - Nd, Er) are analyzed. Formation of ternary molybdate LiRbNd 2 (MoO 4 ) 4 is proved along LiNd(MoO 4 ) 2 -RbNd(MoO 4 )-2 cross-section. Phase diagram of this cross-section is plotted. Similar compounds are synthesized for Ln = La-Eu. The parameters of their monoclinic elementary cells are determined. Luminescent properties of LiRbLa 2 (MoO 4 ) 4 -Nd 3+ are studied. 17 refs., 4 figs., 2 tabs

Impact of reduced graphene oxide on MoS{sub 2} grown by sulfurization of sputtered MoO{sub 3} and Mo precursor films

Energy Technology Data Exchange (ETDEWEB)

Pacley, Shanee, E-mail: shanee.pacley@us.af.mil; Brausch, Jacob; Beck-Millerton, Emory [U.S. Air Force Research Laboratory (AFRL)/Wright Patterson Air Force Base, Wright Patterson, Ohio 45433-7707 (United States); Hu, Jianjun; Jespersen, Michael [University of Dayton Research Institute, 300 College Park, Dayton, Ohio 45469 (United States); Hilton, Al [Wyle Laboratories, 4200 Colonel Glenn Hwy, Beavercreek, Ohio 45431 (United States); Waite, Adam [University Technology Corporation, 1270 N Fairfield Rd., Beavercreek, Ohio 45432 (United States); Voevodin, Andrey A. [Department of Materials Science and Engineering, University of North Texas, 1155 Union Circle, Denton, Texas 76203 (United States)

2016-07-15

Monolayer molybdenum disulfide (MoS{sub 2}), a two dimensional semiconducting dichalcogenide material with a bandgap of 1.8–1.9 eV, has demonstrated promise for future use in field effect transistors and optoelectronics. Various approaches have been used for MoS{sub 2} processing, the most common being chemical vapor deposition. During chemical vapor deposition, precursors such as Mo, MoO{sub 3}, and MoCl{sub 5} have been used to form a vapor reaction with sulfur, resulting in thin films of MoS{sub 2}. Currently, MoO{sub 3} ribbons and powder, and MoCl{sub 5} powder have been used. However, the use of ribbons and powder makes it difficult to grow large area-continuous films. Sputtering of Mo is an approach that has demonstrated continuous MoS{sub 2} film growth. In this paper, the authors compare the structural properties of MoS{sub 2} grown by sulfurization of pulse vapor deposited MoO{sub 3} and Mo precursor films. In addition, they have studied the effects that reduced graphene oxide (rGO) has on MoS{sub 2} structure. Reports show that rGO increases MoS{sub 2} grain growth during powder vaporization. Herein, the authors report a grain size increase for MoS{sub 2} when rGO was used during sulfurization of both sputtered Mo and MoO{sub 3} precursors. In addition, our transmission electron microscopy results show a more uniform and continuous film growth for the MoS{sub 2} films produced from Mo when compared to the films produced from MoO{sub 3}. Atomic force microscopy images further confirm this uniform and continuous film growth when Mo precursor was used. Finally, x-ray photoelectron spectroscopy results show that the MoS{sub 2} films produced using both precursors were stoichiometric and had about 7–8 layers in thickness, and that there was a slight improvement in stoichiometry when rGO was used.

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.

Investigation of BaMoO4-Ln2(MoO4)3 systems (Ln = Nd, Sm, Yb)

International Nuclear Information System (INIS)

Vakulyuk, V.V.; Evdokimov, A.A.; Khomchenko, G.P.

1982-01-01

Using the methods of X-ray phase and differential-thermal analyses phase ratios in the systems BaMoO 4 -Ln 2 (MoO 4 ) 3 (Ln=Nd, Sm, Yb); BaNd 2 (MoO 4 ) 4 -MaGd 2 (MoO 4 ) are studied. Unit cell parameters and the character of melting of the compounds BaLn 2 (MoO 4 ) 4 are specified. Effect of growth conditions on laminated nature of BaGd 2 (MoO 4 ) 4 monocrystals is studied

The Leu72Met Polymorphism of the Prepro-ghrelin Gene is Associated With Alcohol Consumption and Subjective Responses to Alcohol: Preliminary Findings.

Science.gov (United States)

Suchankova, Petra; Yan, Jia; Schwandt, Melanie L; Stangl, Bethany L; Jerlhag, Elisabet; Engel, Jörgen A; Hodgkinson, Colin A; Ramchandani, Vijay A; Leggio, Lorenzo