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Dose Calculations for the Codsiposal WP of HLW Glass and the Shippingport LWBR SNF

Description: The purpose of this calculation is to determine the surface dose rates of a codisposal waste package (WP) containing an intact seed assembly of the Shippingport light-water breeder reactor (LWBR) spent nuclear fuel (SNF) and the Savannah River Site (SRS) high-level waste (HLW) in glass form. The Shippingport LWBR SNF is loaded in a Department of Energy (DOE) standardized 18-in. canister. The canister is surrounded by five 4.5-m-long Hanford pour canisters containing the HLW glass. Gamma dose rate calculation for the WP containing only the HLW glass is also performed. The results will provide information about the contribution of DOE SNF to the total dose rate on the WP surfaces.
Date: November 5, 1999
Creator: Radulescu, G.
Partner: UNT Libraries Government Documents Department

Volumes, Masses, and Surface Areas for Shippingport LWBR Spent Nuclear Fuel in a DOE SNF Canister

Description: The purpose of this calculation is to estimate volumes, masses, and surface areas associated with (a) an empty Department of Energy (DOE) 18-inch diameter, 15-ft long spent nuclear fuel (SNF) canister, (b) an empty DOE 24-inch diameter, 15-ft long SNF canister, (c) Shippingport Light Water Breeder Reactor (LWBR) SNF, and (d) the internal basket structure for the 18-in. canister that has been designed specifically to accommodate Seed fuel from the Shippingport LWBR. Estimates of volumes, masses, and surface areas are needed as input to structural, thermal, geochemical, nuclear criticality, and radiation shielding calculations to ensure the viability of the proposed disposal configuration.
Date: October 22, 1999
Creator: Davis, J.W.
Partner: UNT Libraries Government Documents Department

Fuel Summary Report: Shippingport Light Water Breeder Reactor

Description: The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.
Date: January 1, 1999
Creator: Illum, D.B.; Olson, G.L. & McCardell, R.K.
Partner: UNT Libraries Government Documents Department

Elevated-Temperature Mechanical Properties of an Advanced Type 316 Stainless Steel

Description: Type 316FR stainless steel is a candidate material for the Japanese Demonstration Fast Breeder Reactor Plant to be built in Japan early in the next century. Like type 316L(N), it is a low-carbon grade of stainless steel with a more closely specified nitrogen content and chemistry optimized to enhance elevated-temperature performance. Early in 1994, under sponsorship of The Japan Atomic Power Company, work was initiated at Oak Ridge National Laboratory (ORNL) aimed at obtaining an elevated-temperature mechanical-properties database on a single heat of this material. The product form was 50-mm plate manufactured by the Nippon Steel Corporation. Data include results from long-term creep-rupture tests conducted at temperatures of 500 to 600 C with test times up to nearly 40,000 h, continuous-cycle strain-controlled fatigue test results over the same temperature range, limited creep-fatigue data at 550 and 600 C, and tensile test properties from room temperature to 650 C. The ORNL data were compared with data obtained from several different heats and product forms of this material obtained at Japanese laboratories. The data were also compared with results from predictive equations developed for this material and with data available for type 316 and type 316L(N) stainless steel.
Date: August 1, 1999
Creator: Brinkman, C.R.
Partner: UNT Libraries Government Documents Department

Shippingport LWBR (Th/U Oxide) Fuel Characteristics for Disposal Criticality Analysis

Description: Department of Energy (DOE)-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into eight characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. The Shippingport Light Water Breeder Reactor (LWBR) fuels incorporate more of the conventional materials (zirconium cladding/heavy metal oxides) and fabrication details (rods and spacers) that make them comparable to a typical commercial fuel assembly. The LWBR seed/blanket configuration tested a light-water breeder concept with Th-232/U-233 binary fuel matrix. Reactor design used several assembly configurations at different locations within the same core . The seed assemblies contain the greatest fissile mass per (displaced) unit volume, but the blanket assemblies actually contain more fissile mass in a larger volume; the atom-densities are comparable.
Date: September 1, 1999
Creator: Taylor, L. L. & Loo, H. H.
Partner: UNT Libraries Government Documents Department

Radiation Damage Calculations for the FUBR and BEATRIX Irradiations of Lithium Compounds in EBR-II and FFTF

Description: The Fusion Breeder Reactor (FUBR) and Breeder Exchange Matrix (BEATRIX) experiments were cooperative efforts by members of the International Energy Agency to investigate the irradiation behavior of solid breeder materials for tritium production to support future fusion reactors. Lithium ceramic materials including Li{sub 2}O, LiAlO{sub 2}, Li{sub 4}SiO{sub 4}, and Li{sub 2}ZrO{sub 3} with varying {sup 6}Li enrichments from 0 to 95% were irradiated in a series of experiments in the Experimental Breeder Reactor (EBR II) and in the Fast Flux Test Facility (FFTF) over a period of about 10 years from 1982 to 1992. These experiments were characterized in terms of the nominal fast neutron fluences and measured {sup 6}Li burnup factors, as determined by either mass spectrometry or helium measurements. Radiation damage in these compounds is caused by both the {sup 6}Li-burnup reaction and by all other possible neutron reactions with the atoms in the compound materials. In this report, displacements per atom (dpa) values have been calculated for each type of material in each of the various irradiations that were conducted. Values up to 11% {sup 6}Li-burnup and 130 dpa are predicted for the longest irradiations. The dpa cross sections were calculated for each compound using the SPECOMP computer code. Details of the dpa calculations are presented in the report. Total dpa factors were determined with the SPECTER computer code by averaging the dpa cross sections over the measured or calculated neutron flux spectra for each series of irradiations. Using these new calculations, previously measured radiation damage effects in these lithium compounds can be compared or correlated with other irradiation data on the basis of the dpa factor as well as {sup 6}Li-burnup.
Date: June 17, 1999
Creator: Greenwood, L. R.
Partner: UNT Libraries Government Documents Department

Noble Gas Measurement and Analysis Technique for Monitoring Reprocessing Facilities

Description: An environmental monitoring technique using analysis of stable noble gas isotopic ratios on-stack at a reprocessing facility was developed. This technique integrates existing technologies to strengthen safeguards at reprocessing facilities. The isotopic ratios are measured using a mass spectrometry system and are compared to a database of calculated isotopic ratios using a Bayesian data analysis method to determine specific fuel parameters (e.g., burnup, fuel type, fuel age, etc.). These inferred parameters can be used by investigators to verify operator declarations. A user-friendly software application (named NOVA) was developed for the application of this technique. NOVA included a Visual Basic user interface coupling a Bayesian data analysis procedure to a reactor physics database (calculated using the Monteburns 3.01 code system). The integrated system (mass spectrometry, reactor modeling, and data analysis) was validated using on-stack measurements during the reprocessing of target fuel from a U.S. production reactor and gas samples from the processing of EBR-II fast breeder reactor driver fuel. These measurements led to an inferred burnup that matched the declared burnup with sufficient accuracy and consistency for most safeguards applications. The NOVA code was also tested using numerous light water reactor measurements from the literature. NOVA was capable of accurately determining spent fuel type, burnup, and fuel age for these experimental results. Work should continue to demonstrate the robustness of this system for production, power, and research reactor fuels.
Date: September 1, 1999
Creator: Charlton, William S
Partner: UNT Libraries Government Documents Department