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Interactions of Zircaloy Cladding with Gallium: Final Report

Description: The U.S. Department of Energy has established a dual-track approach to the disposition of plutonium arising from the dismantling of nuclear weapons. Both immobilization and reactor-based mixed-oxide (MOX) fuel technologies are being evaluated. The reactor-based MOX fuel option requires assessment of the potential impact of concentrations of gallium (on the order of 1 to 10 ppm), not present in conventional MOX fhel, on cladding material performance. Three previous repmts"3 identified several compatibility issues relating to the presence of gallium in MOX fuel and its possible reaction with fiel cladding. Gallium initially present in weapons-grade (WG) plutonium is largely removed during processing to produce MOX fhel. After blending the plutonium with uranium, only 1 to 10 ppm gallium is expected in the sintered MOX fuel. Gallium present as gallium oxide (G~OJ could be evolved as the suboxide (G~O). Migration of the evolved G~O and diffusion of gallium in the MOX matrix along thermal gradients could lead to locally higher concentrations of G~03. Thus, while an extremely low concentration of gallium in MOX fiel almost ensures a lack of significant interaction of gallium whh Zircaloy fhel cladding, there remains a small probability that corrosion effects will not be negligible. General corrosion in the form of surface alloying resulting from formation of intermetallic compounds between Zircaloy and gallium should be ma& limited and, therefore, superficial because of the expected low ratio of gallium to the surface area or volume of the Zircaloy cladding. Although the expected concentration of gallium is low and there is very limited volubility of gallium in zirconium, especially at temperatures below 700 "C,4 grain boundary penetration and liquid metal embrittlement (LME) are forms of localized corrosion that were also considered. One fuel system darnage mechanism, pellet clad interaction, has led to some failure of the Zircaloy cladding in light-water ...
Date: September 1, 1998
Creator: Wilson, D.F.; Manneschmidt, E.T.; King, J.F.; Strizak, J.P. & DiStefano, J.R.
Partner: UNT Libraries Government Documents Department

Shipping Cask Studies with MOX Fuel

Description: Tasks of nuclear safety assurance for storage and transport of fresh mixed uranium-plutonium fuel of the VVER-1000 reactor are considered in the view of 3 MOX LTAs introduction into the core. The precise code MCU that realizes the Monte Carlo method is used for calculations.
Date: May 17, 2001
Creator: Pavlovichev, A.M.
Partner: UNT Libraries Government Documents Department

Behavior of Zircaloy Cladding in the Presence of Gallium

Description: The U.S. Department of Energy has established a dual-track approach to the disposition of plutonium arising from the dismantling of nuclear weapons. Both immobilization and reactor-based mixed-oxide (MOX) fuel technologies are being evaluated. The reactor-based MOX fuel option requires assessment of the potential impact of concentrations of gallium (on the order of 1 to 10 ppm), not present in conventional MOX fuel, on cladding material performance. An experimental program was designed to evaluate the performance of prototypic Zircaloy cladding materials against (1) liquid gallium, and (2) various concentrations of G~03. Three types of tests were performed: (1) corrosion, (2) liquid metal embrittlement, and (3) corrosion-mechanical. These tests were to determine corrosion mechanisms, thresholds for temperature and concentration of gallium that delineate behavioral regimes, and changes in the mechanical properties of Zircaloy. Results have generally been favorable for the use of weapons-grade (WG) MOX fhel. The Zircaloy cladding does react with gallium to form intermetallic compounds at >3000 C; however, this reaction is limited by the mass of gallium and is therefore not expected to be significant with a low level (parts per million) of gallium in the MOX fuel. Furthermore, no evidence for grain boundary penetration by gallium or liquid metal embrittlement was observed.
Date: September 28, 1998
Creator: DiStefano, J.R.; King, J.F.; Manneschmidt, E.T.; Strizak, J.P. & Wilson, D.F.
Partner: UNT Libraries Government Documents Department

Errors associated with standard nodal diffusion methods as applied to mixed oxide fuel problems

Description: The evaluation of the disposition of plutonium using light water reactors is receiving increased attention. However, mixed-oxide (MOX) fuel assemblies possess much higher absorption and fission cross- sections when compared to standard UO2 assemblies. Those properties yield very high thermal flux gradients at the interfaces between MOX and UO2 assemblies. It has already been reported that standard flux reconstruction methods (that recover the homogeneous intranodal flux shape using the converged nodal solution) yield large errors in the presence of MOX assemblies. In an accompanying paper, we compare diffusion and simplified PN calculations of a mixed-oxide benchmark problem to a reference transport calculation. In this paper, we examine the errors associated with standard nodal diffusion methods when applied to the same benchmark problem. Our results show that a large portion of the error is associated with the quadratic leakage approximation (QLA) that is commonly used in the standard nodal codes.
Date: July 24, 1998
Creator: Brantley, P. S., LLNL
Partner: UNT Libraries Government Documents Department

Development of a fresh MOX fuel transport package for disposition of weapons plutonium

Description: The US Department of Energy announced its Record of Decision on January 14, 1997, to embark on a dual-track approach for disposition of surplus weapons-usable plutonium using immobilization in glass or ceramics and burning plutonium as mixed-oxide (MOX) fuel in reactors. In support of the MOX fuel alternative, Oak Ridge National Laboratory initiated development of conceptual designs for a new package for transporting fresh (unirradiated) MOX fuel assemblies between the MOX fabrication facility and existing commercial light-water reactors in the US. This paper summarizes progress made in development of new MOX transport package conceptual designs. The development effort has included documentation of programmatic and technical requirements for the new package and development and analysis of conceptual designs that satisfy these requirements.
Date: November 1, 1998
Creator: Ludwig, S.B.; Pope, R.B.; Shappert, L.B.; Michelhaugh, R.D. & Chae, S.M.
Partner: UNT Libraries Government Documents Department

Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

Description: The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78).
Date: November 1, 1997
Creator: Wachs, G.W.
Partner: UNT Libraries Government Documents Department

Fabrication, Inspection, and Test Plan for the Advanced Test Reactor (ATR) High-Power Mixed-Oxide (MOX) Fuel Irradiation Project

Description: The Department of Energy (DOE) Fissile Disposition Program (FMDP) has announced that reactor irradiation of Mixed-Oxide (MOX) fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The High-Power MOX fuel test will be irradiated in the Advanced Test Reactor (ATR) to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. The purpose of the high-power experiment, in conjunction with the currently ongoing average-power experiment at the ATR, is to contribute new information concerning the response of WG plutonium under more severe irradiation conditions typical of the peak power locations in commercial reactors. In addition, the high-power test will contribute experience with irradiation of gallium-containing fuel to the database required for resolution of generic CLWR fuel design issues. The distinction between "high-power" and "average-power" relates to the position within the nominal CLWR core. The high-power test project is subject to a number of requirements, as discussed in the Fissile Materials Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation High-Power Test Project Plan (ORNL/MD/LTR-125).
Date: September 1, 1998
Creator: Wachs, G. W.
Partner: UNT Libraries Government Documents Department

Pilot-scale equipment development for pyrochemical reduction of spent oxide fuel

Description: Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical conditioning of sodium-bonded spent metal fuel from Experimental Breeder Reactor II, resulting in uranium, ceramic, and metal waste forms. Equipment is being developed at ANL which will precondition irradiated oxide fuel and demonstrate the application of electrometallurgical conditioning to such non-metallic fuels as well. The oxide reduction process preconditions irradiated oxide fuel such that uranium and transuranic (TRU) constituents are chemically reduced into metallic form via a molten Li/LiCl-based reduction system. In this form the spent fuel is further conditioned in an electrorefiner and waste handling equipment, thereby placing the uranium, TRU elements, and fissions products into stable forms suitable for placement in a long-term repository. Development of the Li/LiCl-based oxide reduction process has proceeded at lab- (nominally 50 grams of heavy metal (HM)) and engineering-scale (nominally 10-kg of HM) for unirradiated oxide fuel. The presentation described the process and equipment design for scale-up from lab- and engineering-scale reduction of unirradiated oxide fuel in gloveboxes to pilot-scale (up to 100-kg of HM) reduction of irradiated oxide fuel in a hot cell. [Abstract only.]
Date: July 1, 1998
Creator: Herrmann, S.D.; King, R.W.; Durstine, K.R. & Eberl, C.S.
Partner: UNT Libraries Government Documents Department

A Deterministic Study of the Deficiency of the Wigner-Seitz Approximation for Pu/MOX Fuel Pins

Description: The Wigner-Seitz pin-cell approximation has long been applied as a modeling approximation in analysis of UO2 lattice fuel cells. In the past, this approximation has been appropriate for such fuel. However, with increasing attention drawn to mixed-oxide (MOX) fuels with significant plutonium content, it is important to understand the implications of the approximation in a uranium-plutonium matrix. The special geometric capabilities of the deterministic NEWT computer code have been used to assess the adequacy of the Wigner-Seitz cell in such an environment, as part of a larger study of computational aspects of MOX fuel modeling. Results of calculations using various approximations and boundary conditions are presented, and are validated by comparison to results obtained using KENO V.a and XSDRNPM.
Date: September 27, 1999
Creator: DeHart, M.D.
Partner: UNT Libraries Government Documents Department

Criticality Safety Scoping Study for the Transport of Weapons-Grade Mixed-Oxide Fuel Using the MO-1 Shipping Package

Description: This report provides the criticality safety information needed for obtaining certification of the shipment of mixed-oxide (MOX) fuel using the MO-1 [USA/9069/B()F] shipping package. Specifically, this report addresses the shipment of non-weapons-grade MOX fuel as certified under Certificate of Compliance 9069, Revision 10. The report further addresses the shipment of weapons-grade MOX fuel using a possible Westinghouse fuel design. Criticality safety analysis information is provided to demonstrate that the requirements of 10 CFR S 71.55 and 71.59 are satisfied for the MO-1 package. Using NUREG/CR-5661 as a guide, a transport index (TI) for criticality control is determined for the shipment of non-weapons-grade MOX fuel as specified in Certificate of Compliance 9069, Revision 10. A TI for criticality control is also determined for the shipment of weapons-grade MOX fuel. Since the possible weapons-grade fuel design is preliminary in nature, this report is considered to be a scoping evaluation and is not intended as a substitute for the final criticality safety analysis of the MO-1 shipping package. However, the criticality safety evaluation information that is presented in this report does demonstrate the feasibility of obtaining certification for the transport of weapons-grade MOX lead test fuel using the MO-1 shipping package.
Date: May 1, 1999
Creator: Dunn, M.E. & Fox, P.B.
Partner: UNT Libraries Government Documents Department

Licensing issues associated with the use of mixed-oxide fuel in US commercial nuclear reactors

Description: On January 14, 1997, the Department of Energy, as part of its Record of Decision on the storage and disposition of surplus nuclear weapons materials, committed to pursue the use of excess weapons-usable plutonium in the fabrication of mixed-oxide (MOX) fuel for consumption in existing commercial nuclear power plants. Domestic use of MOX fuel has been deferred since the late 1970s, principally due to nuclear proliferation concerns. This report documents a review of past and present literature (i.e., correspondence, reports, etc.) on the domestic use of MOX fuel and provides discussion on the technical and regulatory issues that must be addressed by DOE (and the utility/consortia selected by DOE to effect the MOX fuel consumption strategy) in obtaining approval from the Nuclear Regulatory Commission to use MOX fuel in one or a group of existing commercial nuclear power plants.
Date: April 1, 1997
Creator: Williams, D.L. Jr.
Partner: UNT Libraries Government Documents Department

Cost estimate and economic issues associated with the MOX option (prior to DOE`s record of decision)

Description: Before the January 1997 Record of Decision (ROD), the U.S. Department of Energy Office of Fissile Materials Disposition (DOE-MD) evaluated three technologies for the disposition of {approximately}50 MT of surplus plutonium from defense-related programs-reactors, immobilization, and deep boreholes. As part of the process supporting the ROD, and comprehensive assessment of technical viability, cost, and schedule was conducted by DOE-MD and its national laboratory contractors. Oak Ridge National Laboratory managed and coordinated the life-cycle cost (LCC) assessment effort for this program. This report discusses the economic analysis methodology and the results for the reactor options considered prior to ROD. A secondary intent of the report is to discuss major technical and economic issues that impact cost and schedule. To evaluate the economics of the reactor option and other technologies on an equitable basis, a set of cost-estimating guidelines and a common cost-estimating format were utilized by all three technology teams. This report includes the major economic analysis assumptions and the comparative constant-dollar and discounted-dollar LCCs for all nine reactor scenarios.
Date: April 1, 1997
Creator: Reid, R.L. & Miller, J.W.
Partner: UNT Libraries Government Documents Department

Application of Sensitivity and Uncertainty Analysis Methods to a Validation Study for Weapons-Grade Mixed-Oxide Fuel

Description: At the Oak Ridge National Laboratory (ORNL), sensitivity and uncertainty (S/U) analysis methods and a Generalized Linear Least-Squares Methodology (GLLSM) have been developed to quantitatively determine the similarity or lack thereof between critical benchmark experiments and an application of interest. The S/U and GLLSM methods provide a mathematical approach, which is less judgment based relative to traditional validation procedures, to assess system similarity and estimate the calculational bias and uncertainty for an application of interest. The objective of this paper is to gain experience with the S/U and GLLSM methods by revisiting a criticality safety evaluation and associated traditional validation for the shipment of weapons-grade (WG) MOX fuel in the MO-1 transportation package. In the original validation, critical experiments were selected based on a qualitative assessment of the MO-1 and MOX contents relative to the available experiments. Subsequently, traditional trending analyses were used to estimate the {Delta}k bias and associated uncertainty. In this paper, the S/U and GLLSM procedures are used to re-evaluate the suite of critical experiments associated with the original MO-1 evaluation. Using the S/U procedures developed at ORNL, critical experiments that are similar to the undamaged and damaged MO-1 package are identified based on sensitivity and uncertainty analyses of the criticals and the MO-1 package configurations. Based on the trending analyses developed for the S/U and GLLSM procedures, the {Delta}k bias and uncertainty for the most reactive MO-1 package configurations are estimated and used to calculate an upper subcritical limit (USL) for the MO-1 evaluation. The calculated bias and uncertainty from the S/U and GLLSM analyses lead to a calculational USL that supports the original validation study for the MO-1.
Date: July 20, 2001
Creator: Dunn, M.E.
Partner: UNT Libraries Government Documents Department

Review of the Analyses of the Doppler-Effect Measurements in SEFOR (Southwest Experimental Fast Oxide Reactor)

Description: The SEFOR experimental results and the three original analyses are reviewed and discussed. The emphasis of the review is placed on aspects that are pertinent to a possible modern re-analysis of the experimental results. Looking at the analysis results in terms of zero and first order effects shows that the zero order effects, the Doppler constant of the two SEFOR cores, are obtained by the three analyses in satisfactory agreement. But the first order effects, but temperature variation of this Doppler-constant quantity, cannot be determined with any informative accuracy. Since this is likely due to limitations in the experiments, a re-analysis - except for methodological reasons - does not appear to be fruitful.
Date: July 1987
Creator: Ott, Karl O.
Partner: UNT Libraries Government Documents Department

Fission-Product Releases to the Primary System of EBR-II, January 1974-March 1975

Description: Seven releases of fission products occurred in EBR-II from January 1974 to March 1975 - - five from mixed-oxide elements and two from sodium-bonded driver-fuel elements. Four releases were from elements that contained a xenon tag, which aided considerably in locating three of the elements; data from the fourth element allowed estimation of changes of tag composition with reactor exposure. Rapid release of fission from two breached mixed-oxide elements caused the reactor to trip because of increased delayed-neutron activity, the first time such behavior has been observed. Identification of a subassembly of Mark-1A driver-fuel elements was complicated by multiple failure of its untagged elements during the diagnosis period. Several of these elements had some exposed fuel in the core, which was the likely cause of increasing delayed-neutron signals from the subassembly.
Date: October 1976
Creator: So, B. Y. C.; Lambert, J. D. B.; Johnson, D. L.; Ebersole, E. R. & Brunson, G. S.
Partner: UNT Libraries Government Documents Department

Examination and Evaluation of in-Reactor Fracture of Shroud Tubes in Mixed-Oxide Fuel Experiment X159

Description: During disassembly and subsequent visual examinations of X159 (a Mark A-19A type of subassembly containing mixed-oxide fuel), 11 of 19 shroud tubes were found fully or partially severed. Several of the capsules within their shroud tubes were distinctly kinked at axial locations near those at which the shroud tubes were severed. The examination also disclosed that the shroud-tube fractures occurred during reactor operations.
Date: June 1977
Creator: Flinn, J. E.
Partner: UNT Libraries Government Documents Department

A Portable Calorimeter System for Nondestructive Assay of Mixed-Oxide Fuels

Description: Calorimetric assay provides a precise, nondestructive method to determine sample plutonium content based on the heat emitted by decaying radionuclides. This measurement, in combination with a gamma-spectrometer analysis of sample isotopic content, yields the total sample plutonium mass. The technique is applicable to sealed containers and is essentially independent of sample matrix configuration and elemental composition. Conventional calorimeter designs employ large water-bath heat sinks and lack the portability needed by inspection personnel. The ANL air-chamber isothermal calorimeters are low-thermal-capacitance devices which eliminate the need for large constant-temperature heat sinks. These instruments are designed to use a feedback system that applies power to maintain the sample chamber at a constant electrical resistance and, therefore, at a constant temperature. The applied-power difference between a plutonium-containing sample and a blank determines the radioactive-decay power. The operating characteristics of a calorimeter designed for assaying mixed-oxide powders, fuel pellets, and plutonium-containing solutions are discussed. This device consists of the calorimeter, sample pre-heater, and a microprocessor-controlled data-acquisition system. The small-sample device weighs 18 kg and has a measurement cycle of 20 min, with a precision of 0.1% at 10 mW. A 100-min gamma-ray measurement gives the specific power with a precision of better than 1% for samples containing 1 to 2 g of plutonium.
Date: 1978
Creator: Roche, C. T.; Perry, R. B.; Lewis, R. N.; Jung, E. A. & Haumann, J. R.
Partner: UNT Libraries Government Documents Department

Enhanced Thermal Conductivity Oxide Fuels

Description: the purpose of this project was to investigate the feasibility of increasing the thermal conductivity of oxide fuels by adding small fractions of a high conductivity solid phase.
Date: January 17, 2006
Creator: Solomon, Alvin; Revankar, Shripad & McCoy, J. Kevin
Partner: UNT Libraries Government Documents Department

Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint US/Russian Progress Report for Fiscal Year 1997, Volume 4, part 4-ESADA Plutonium Program Critical Experiments: Single-Region Core Configurations

Description: The purpose of this study is to simulate and assess the findings from selected ESADA experiments. It is presented in the format prescribed by the Nuclear Energy Agency Nuclear Science Committee for material to be included in the International Handbook of Evaluated Criticality Safety Benchmark Experiments.
Date: May 1, 1999
Creator: Akkurt, H. & Abdurrahman, N.M.
Partner: UNT Libraries Government Documents Department

Unique QA/QC requirements for analytical chemistry at LANL

Description: One of the missions of group NMT-1 (Nuclear Materials Technology Division/Analytical Chemistry) at Los Alamos National Laboratory (LANL) is to provide analysis of both radioactive and nonradioactive materials to address the stockpile stewardship needs of the US Department of Energy (DOE). Trace to high levels of various constituents are measured using traditional analytical methods and state-of-the-art instrumental methods. Capabilities include Pu and U assay, wet chemistry, plasma spectroscopy, mass spectrometry radiochemistry, x-ray fluorescence, anion and cation analysis, special standards preparation, surface analysis, and gas analysis. The authors are currently developing and implementing a plan to independently assess the quality of the analytical data produced by NMT-1. Nuclear materials of a matrix similar to the client`s samples but having different concentration levels of analytes that are representative of the client`s samples will be used. Well-characterized, stable, homogeneous materials have been identified as possible candidates for single-blind quality control (QC) samples. These materials include Pu metal, Pu oxide, uranium metal, uranium oxide, and uranium-plutonium mixed oxide with varying degrees of purity. These single-blind samples will be periodically distributed along with regular client samples to be analyzed by the above mentioned analytical methods.
Date: December 31, 1998
Creator: Tandon, L.; Gautier, M.A.; Hammond, C.F. & Porterfield, D.R.
Partner: UNT Libraries Government Documents Department

Nuclear fuels technologies fiscal year 1996 research and research development test results

Description: During fiscal year 1996, the Department of Energy`s Office of Fissile Materials Disposition (OFMD) funded Los Alamos National Laboratory (LANL) to investigate issues associated with the fabrication of plutonium from dismantled weapons into mixed-oxide (MOX) nuclear fuel for disposition in nuclear power reactors. These issues can be divided into two main categories: issues associated with the fact that the plutonium from dismantled weapons contains gallium, and issues associated with the unique characteristics of the PuO{sub 2} produced by the dry conversion process that OFMD is proposing to convert the weapons material. Initial descriptions of the experimental work performed in fiscal year 1996 to address these issues can be found in Nuclear Fuels Technologies Fiscal Year 1996 Research and Development Test Matrices`. However, in some instances the change in programmatic emphasis towards the Parallex program either altered the manner in which some of these experiments were performed (i.e., the work was done as part of the Parallex fabrication development and not as individual separate-effects tests as originally envisioned) or delayed the experiments into Fiscal Year 1997. This report reviews the experiments that were conducted and presents the results. 7 figs., 14 tabs.
Date: November 8, 1996
Creator: Beard, C.A.; Blair, H.T.; Buksa, J.J. & Butt, D.P.
Partner: UNT Libraries Government Documents Department