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

Description: The mission of this Conversion and Blending Facility (CBF) will be to blend surplus HEU metal and alloy with depleted uranium metal to produce an LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal.
Date: July 5, 1995
Partner: UNT Libraries Government Documents Department

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

Description: This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials to pure HEU uranyl nitrate (UNH) and (2) blend pure HEU UNH with depleted and natural UNH to produce HEU UNH crystals. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU Will be produced as a waste suitable for storage or disposal.
Date: July 5, 1995
Partner: UNT Libraries Government Documents Department

Further Dosimetry Studies at Rhode Island Nuclear Science Center.

Description: The RINSC is a 2 mega-watt, light water and graphite moderated and cooled reactor that has a graphite thermal column built as a user facility for sample irradiation. Over the past decade, after the reactor conversion from a highly-enriched uranium core to a low-enriched one, flux and dose measurements and calculations had been performed in the thermal column to update the ex-core parameters and to predict the effect from in-core fuel burn-up and rearrangement. The most recent data from measurements and calculations that have been made at the RINSC thermal column since October of 2005 are reported.
Date: May 5, 2008
Creator: Reciniello,R.N.; Holden, N.E.; Hu, J.-P.; Johnson, D.G.; Meddleton, M. & Tehan, T.N.
Partner: UNT Libraries Government Documents Department

Method for fabricating {sup 99}Mo production targets using low enriched uranium, {sup 99}Mo production targets comprising low enriched uranium

Description: A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material (low enriched U) circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate.
Date: December 31, 1993
Creator: Wiencek, T.C.; Matos, J.E. & Hofman, G.L.
Partner: UNT Libraries Government Documents Department

Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel

Description: This document presents information concerning: analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU; changes to technical specifications mandated by the conversion of the GTRR to low enrichment fuel; changes in the Safety Analysis Report mandated by the conversion of the GTRR to low enrichment fuel; and copies of all changed pages of the SAR and the technical specifications.
Date: September 1, 1992
Creator: Matos, J.E.; Mo, S.C. & Woodruff, W.L.
Partner: UNT Libraries Government Documents Department

Processing of LEU targets for {sup 99}Mo production: Dissolution of U{sub 3}Si{sub 2} targets by alkaline hydrogen peroxide

Description: Low-enriched uranium silicide targets designed to recover fission product {sup 99}Mo were dissolved in alkaline hydrogen peroxide (H{sub 2}O{sub 2} plus NaOH) at about 90C. Sintering of matrix aluminum powder during irradiation and heat treatment retarded aluminum dissolution and prevented silicide particle dispersion. Gas evolved during dissolution is suspected to adhere to particles and block hydroxide ion contact with aluminum. Reduction of base concentrations from 5M to O.lM NaOH yielded similar silicide dissolution and peroxide destruction rates, simplifying later processing. Future work in particle dispersion enhancement, {sup 99}Mo separation, and waste disposal is also discussed.
Date: September 1, 1995
Creator: Buchholz, B.A. & Vandegrift, G.F.
Partner: UNT Libraries Government Documents Department

HEU to LEU Conversion and Blending Facility: UF{sub 6} blending alternative to produce LEU UF{sub 6} for commercial use

Description: US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials; the nuclear material will be converted to a form more proliferation- resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. Five technologies for blending HEU will be assessed; blending as UF{sub 6} to produce a UF{sub 6} product for commercial use is one of them. This document provides data to be used in the environmental impact analysis for the UF{sub 6} blending HEU disposition option. Resource needs, employment needs, waste and emissions from plant, hazards, accident scenarios, and intersite transportation are discussed.
Date: September 1, 1995
Partner: UNT Libraries Government Documents Department

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

Description: This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials into pure HEU oxide and (2) blend the pure HEU oxide with depleted and natural uranium oxide to produce an LWR grade LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU will be produced as a waste suitable for storage or disposal.
Date: July 5, 1995
Partner: UNT Libraries Government Documents Department

Low-enriched uranium holdup measurements in Kazakhstan

Description: Quantification of the residual nuclear material remaining in process equipment has long been a challenge to those who work with nuclear material accounting systems. Fortunately, nuclear material has spontaneous radiation emissions that can be measured. If gamma-ray measurements can be made, it is easy to determine what isotope a deposit contains. Unfortunately, it can be quite difficult to relate this measured signal to an estimate of the mass of the nuclear deposit. Typically, the measurement expert must work with incomplete or inadequate information to determine a quantitative result. Simplified analysis models, the distribution of the nuclear material, any intervening attenuation, background(s), and the source-to-detector distance(s) can have significant impacts on the quantitative result. This presentation discusses the application of a generalized-geometry holdup model to the low-enriched uranium fuel pellet fabrication plant in Ust-Kamenogorsk, Kazakhstan. Preliminary results will be presented. Software tools have been developed to assist the facility operators in performing and documenting the measurements. Operator feedback has been used to improve the user interfaces.
Date: December 31, 1998
Creator: Barham, M.A.; Ceo, R.N. & Smith, S.E.
Partner: UNT Libraries Government Documents Department

Development and processing of LEU targets for {sup 99}Mo production

Description: Most of the world`s supply of {sup 99m}Tc for medical purposes is currently produced from the decay of {sup 99}Mo derived from the fissioning of high-enriched uranium (HEU). Substantial progress has been made in developing targets and chemical processes for producing {sup 99}Mo using low-enriched uranium (LEU). Target development has been focused on a uranium-metal foil target as a replacement for the coated-UO{sub 2} Cintichem-type target. Although the first designs were not successful because of ion mixing-induced bonding of the uranium foil to the target tubes, recent irradiations of modified targets have proven successful. Only minor modifications of the Cintichem chemical process are required for the uranium-metal foil targets. A demonstration using prototypically irradiated targets is anticipated in February 1997. Progress has also been made in basic dissolution of both uranium-metal foil and aluminum-clad U{sub 3}Si{sub 2} dispersion fuel targets.
Date: April 1, 1997
Creator: Snelgrove, J.L.; Vandegrift, G.F. & Hofman, G.L.
Partner: UNT Libraries Government Documents Department

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

Description: Presented here are recent experimental results on tests of a modified Cintichem process for producing {sup 99}Mo from low enriched uranium (LEU). Studies were focused in three areas: (1) testing the effects on {sup 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 {sup 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 {sup 99}Mo product. The results also show that the selected barrier materials, Cu, Fe, and Ni, do not interfere with {sup 99}Mo recovery and can be removed during chemical processing of the LEU target.
Date: December 31, 1996
Creator: Wu, D.; Landsberger, S. & Vandegrift, G.F.
Partner: UNT Libraries Government Documents Department

LEU conversion status of US research reactors, September 1996

Description: This paper summarizes the conversion status of research and test reactors in the United States from the use of fuels containing highly- enriched uranium (HEU, greater than or equal to 20%) to the use of fuels containing low-enriched uranium (LEU, < 20%). Estimates of the uranium densities required for conversion are made for reactors with power levels greater than or equal to 1 MW that are not currently involved in the LEU conversion process.
Date: October 7, 1996
Creator: Matos, J.E.
Partner: UNT Libraries Government Documents Department

Development of very high-density low-enriched uranium fuels

Description: The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm{sup 3}, based on the use of {gamma}-stabilized uranium alloys. Fabrication development teams and facilities are being put into place and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and first results should be available by end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun.
Date: February 1, 1997
Creator: Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L. & Wiencek, T.C.
Partner: UNT Libraries Government Documents Department

Development and processing of LEU targets for {sup 99}Mo production

Description: Substituting LEU for HEU in targets for producing fission-product {sup 99}Mo requires changes in target design and chemical processing. We have made significant progress in developing targets and chemical processes for this purpose. Target development was concentrated on a U- metal foil target as a replacement for the coated-UO{sub 2} Cintichem- type target. Although the first designs were not successful because of ion mixing-induced bonding of the U foil to the target tubes, recent irradiations of modified targets have proven successful. It was shown that only minor modifications of the Cintichem chemical process are required for the U-metal foil targets. A demonstration using prototypically irradiated targets is anticipated by the end of 1996. Progress was also made in basic dissolution of both U-metal foil and Al-clad U{sub 3}Si{sub 2} dispersion fuel targets, and work in this area is also continuing.
Date: February 1, 1997
Creator: Snelgrove, J.L.; Vandergrift, G.F. & Hofman, G.L.
Partner: UNT Libraries Government Documents Department

Progress in dissolving modified LEU Cintichem targets

Description: A process is under development to use low-enriched uranium (LEU) metal targets for production of {sup 99}Mo. The first step is to dissolve the irradiated foil. In past work, this has been done by heating a closed (sealed) vessel containing the foil and a solution of nitric and sulfuric acids. In this work, the authors have demonstrated that (1) the dissolver solution can contain nitric acid alone, (2) uranium dioxide is also dissolved by nitric acid alone, and (3) barrier metals of Cu, Fe, or Ni on the U foil are also dissolved by nitric acid. Changes to the dissolver design and operation needed to accommodate the uranium foil are discussed, including (1) simple operations that are easy to do in a remote-maintenance facility, (2) heat removal from the irradiated LEU foil, and (3) cold trap operation with high dissolver pressures.
Date: December 31, 1996
Creator: Leonard, R.A.; Chen, L.; Mertz, C.J. & Vandegrift, G.F.
Partner: UNT Libraries Government Documents Department

Transient analysis for the tajoura critical facility with IRT-2M HEU fuel and IRT-4M leu fuel : ANL independent verification results.

Description: Calculations have been performed for postulated transients in the Critical Facility at the Tajoura Nuclear Research Center (TNRC) in Libya. These calculations have been performed at the request of staff of the Renewable Energy and Water Desalinization Research Center (REWDRC) who are performing similar calculations. The transients considered were established during a working meeting between ANL and REWDRC staff on October 1-2, 2005 and subsequent email correspondence. Calculations were performed for the current high-enriched uranium (HEU) core and the proposed low-enriched uranium (LEU) core. These calculations have been performed independently from those being performed by REWDRC and serve as one step in the verification process.
Date: December 2, 2005
Creator: Garner, P. L. & Hanan, N. A.
Partner: UNT Libraries Government Documents Department

The LEU conversion status of U.S. Research Reactors.

Description: This paper summarizes the conversion status of US research and test reactors and estimates uranium densities needed to convert reactors with power levels 21 MW from HEU ({ge} 20% U-235) to LEU (&lt;20% U-235) fuels. Detailed conversion studies for each reactor need to be completed in order to establish the feasibility of using LEU fuels.
Date: November 14, 1997
Creator: Matos, J. E.
Partner: UNT Libraries Government Documents Department

U.S. transparency monitoring of HEU oxide conversion and blending to LEU hexafluoride at three Russian blending plants

Description: The down-blending of Russian highly enriched uranium (HEU) takes place at three Russian gaseous centrifuge enrichment plants. The fluorination of HEU oxide and down-blending of HEU hexafluoride began in 1994, and shipments of low enriched uranium (LEU) hexafluoride product to the United States Enrichment Corporation (USEC) began in 1995 US transparency monitoring under the HEU Purchase Agreement began in 1996 and includes a permanent monitoring presence US transparency monitoring at these facilities is intended to provide confidence that HEU is received and down-blended to LEU for shipment to USEC The monitoring begins with observation of the receipt of HEU oxide shipments, including confirmation of enrichment using US nondestructive assay equipment The feeding of HEU oxide to the fluorination process and the withdrawal of HEU hexafluoride are monitored Monitoring is also conducted where the blending takes place and where shipping cylinders are filled with LEU product. A series of process and material accountancy documents are provided to US monitors.
Date: July 27, 1998
Creator: Leich, D.
Partner: UNT Libraries Government Documents Department

Low enrichment fuel conversion for Iowa State University. Final report

Description: The UTR-10 research and teaching reactor at Iowa State University (ISU) has been converted from high-enriched fuel (HEU) to low- enriched fuel (LEU) under Grant No. DE-FG702-87ER75360 from the Department of Energy (DOE). The original contract period was August 1, 1987 to July 31, 1989. The contract was extended to February 28, 1991 without additional funding. Because of delays in receiving the LEU fuel and the requirement for disassembly of the HEU assemblies, the contract was renewed first through May 31, 1992, then through May 31, 1993 with additional funding, and then again through July 31, 1994 with no additional funding. In mid-August the BMI cask was delivered to Iowa State. Preparations are underway to ship the HEU fuel when NRC license amendments for the cask are approved.
Date: October 17, 1996
Creator: Bullen, D.B. & Wendt, S.E.
Partner: UNT Libraries Government Documents Department

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

Description: 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, <20%) silicide fuel with 4.8 g/cm{sup 3} that provides nearly the same neutron flux for experiments as the HEU design, but has a less favorable fuel cycle economy. If an LEU fuel with a uranium density of 6.0 - 6.5 g/cm{sup 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{sup 3} would enhance the performance of the LEU core. The RERTR 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.
Date: December 1, 1995
Creator: Mo, S.C.; Hanan, N.A. & Matos, J.E.
Partner: UNT Libraries Government Documents Department

Peak fitting applied to low-resolution enrichment measurements

Description: Materials accounting at bulk processing facilities that handle low enriched uranium consists primarily of weight and uranium enrichment measurements. Most low enriched uranium processing facilities draw separate materials balances for each enrichment handled at the facility. The enrichment measurement determines the isotopic abundance of the {sup 235}U, thereby determining the proper strata for the item, while the weight measurement generates the primary accounting value for the item. Enrichment measurements using the passive gamma radiation from uranium were developed for use in US facilities a few decades ago. In the US, the use of low-resolution detectors was favored because they cost less, are lighter and more robust, and don`t require the use of liquid nitrogen. When these techniques were exported to Europe, however, difficulties were encountered. Two of the possible root causes were discovered to be inaccurate knowledge of the container wall thickness and higher levels of minor isotopes of uranium introduced by the use of reactor returns in the enrichment plants. the minor isotopes cause an increase in the Compton continuum under the 185.7 keV assay peak and the observance of interfering 238.6 keV gamma rays. The solution selected to address these problems was to rely on the slower, more costly, high-resolution gamma ray detectors when the low-resolution method failed. Recently, these gamma ray based enrichment measurement techniques have been applied to Russian origin material. The presence of interfering gamma radiation from minor isotopes was confirmed. However, with the advent of fast portable computers, it is now possible to apply more sophisticated analysis techniques to the low-resolution data in the field. Explicit corrections for Compton background, gamma rays from {sup 236}U daughters, and the attenuation caused by thick containers can be part of the least squares fitting routine. Preliminary results from field measurements in Kazakhstan will be discussed.
Date: December 1, 1998
Creator: Bracken, D.; McKown, T.; Sprinkle, J.K. Jr.; Gunnink, R.; Kartoshov, M.; Kuropatwinski, J. et al.
Partner: UNT Libraries Government Documents Department

Analyses of Weapons-Grade MOX VVER-1000 Neutronics Benchmarks: Pin-Cell Calculations with SCALE/SAS2H

Description: A series of unit pin-cell benchmark problems have been analyzed related to irradiation of mixed oxide fuel in VVER-1000s (water-water energetic reactors). One-dimensional, discrete-ordinates eigenvalue calculations of these benchmarks were performed at ORNL using the SAS2H control sequence module of the SCALE-4.3 computational code system, as part of the Fissile Materials Disposition Program (FMDP) of the US DOE. Calculations were also performed using the SCALE module CSAS to confirm the results. The 238 neutron energy group SCALE nuclear data library 238GROUPNDF5 (based on ENDF/B-V) was used for all calculations. The VVER-1000 pin-cell benchmark cases modeled with SAS2H included zero-burnup calculations for eight fuel material variants (from LEU UO{sub 2} to weapons-grade MOX) at five different reactor states, and three fuel depletion cases up to high burnup. Results of the SAS2H analyses of the VVER-1000 neutronics benchmarks are presented in this report. Good general agreement was obtained between the SAS2H results, the ORNL results using HELIOS-1.4 with ENDF/B-VI nuclear data, and the results from several Russian benchmark studies using the codes TVS-M, MCU-RFFI/A, and WIMS-ABBN. This SAS2H benchmark study is useful for the verification of HELIOS calculations, the HELIOS code being the principal computational tool at ORNL for physics studies of assembly design for weapons-grade plutonium disposition in Russian reactors.
Date: January 11, 2001
Creator: Ellis, R.J.
Partner: UNT Libraries Government Documents Department

SHEBA-II as a criticality safety benchmark experiment

Description: SHEBA-II (Solution High Energy Burst Assembly-II) is a critical assembly experiment currently (1995) being operated at the Los Alamos Critical Experiments Facility. It is a bare assembly fueled with an aqueous solution of about 5% enriched uranyl fluoride that is stored in four critically safe steel tanks. The solution is transferred to the critical assembly vessel (CAV) by applying gas pressure to the storage tanks. Reactivity is controlled by varying the solution level, and a safety rod may be inserted in a thimble along the central axis of the CAV for fast shutdown. The simple geometry provided by this cylindrical system allows for easily applied calculational methods, and thus SHEBA-II is ideally suited for use as a criticality safety benchmark experiment.
Date: July 1, 1995
Creator: LaBauve, R.J. & Sapir, J.L.
Partner: UNT Libraries Government Documents Department