The Experts below are selected from a list of 240 Experts worldwide ranked by ideXlab platform
P.s. Tam - One of the best experts on this subject based on the ideXlab platform.
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Safety Evaluation Report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). Supplement No. 18
1995Co-Authors: P.s. TamAbstract:In June 1982, the Nuclear Regulatory Commission staff (NRC staff or staff) issued a Safety Evaluation Report, NUREG-0847, regarding the application by the Tennessee Valley Authority (TVA or the applicant) for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2. Each of the following sections and appendices of this supplement is numbered the same as the section or appendix of the SER that is being updated, and the discussions are supplementary to, and not in lieu of, the discussion in the SER, unless otherwise noted. Accordingly, Appendix A continues the chronology of the Safety review. Appendix E lists principal contributors to this supplement. Appendix FF is added in this supplement. The other appendices are not changed by this supplement.
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Safety Evaluation Report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). Supplement No. 17
1995Co-Authors: P.s. TamAbstract:This Report supplements the Safety Evaluation Report (SER), NUREG-0847 (June 1982), Supplement No. 1 (September 1982), Supplement No. 2 (January 1984), Supplement No. 3 (January 1985), Supplement No. 4 (March 1985), Supplement No. 5 (November 1990), Supplement No. 6 (April 1991), Supplement No. 7 (September 1991), 1991), Supplement No. 8 (January 1992), Supplement No. 9 (June 1992), Supplement No. 10 (October 1992), Supplement No. 11 (April.1993), Supplement No. 12 (October 1993), Supplement No. 13 (April 1994), Supplement No. 14 (December 1994), Supplement No. 15 (June 1995), and Supplement No. 16 (September 1995) issued by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50--390 and 50--391). The facility is located in Rhea county, Tennessee, near the Watts Bar Dam on the Tennessee River. In this supplement, NRC examines the significant problems of construction quality and quality assurance effectiveness that led TVA to withdraw its certification in 1985 that Watts Bar Unit I was ready to load fuel. Also discussed are the extensive corrective actions performed by TVA according to its nuclear performance plans and other supplemental programs, and NRC`s extensive oversight to determine whether the Watts Bar Unit 1 construction quality and TVA`s operational readiness and quality assurance effectiveness are adequate for a low-power operating license to be issued. SSER 17 does not address Watts Bar Unit 2, except for the systems which are necessary to support Unit 1 operation.
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Safety Evaluation Report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). Supplement No. 15
1995Co-Authors: P.s. TamAbstract:This Report supplements the Safety Evaluation Report (SER), NUREG-0847 (June 1982), Supplement No. 1 (September 1982), Supplement No. 2 (January 1984), Supplement No. 3 (January 1985), Supplement No. 4 (March 1985), Supplement No. 5 (November 1990), Supplement No. 6 (April 1991), Supplement No. 7 (September 1991), Supplement No. 8 (January 1992), Supplement No. 9 (June 1992), Supplement No. 10 (October 1992), Supplement No. 11 (April 1993), Supplement No. 12 (October 1993), Supplement No. 13 (April 1994), and Supplement No. 14 (December 1994) issued by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). The facility is located in Rhea County, Tennessee, near the Watts Bar Dam on the Tennessee River. This supplement provides recent information regarding resolution of some of the outstanding and confirmatory items, and proposed license conditions identified in the SER.
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Safety Evaluation Report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). Supplement No. 14
1994Co-Authors: P.s. TamAbstract:Supplement No. 14 to the Safety Evaluation Report for the application filed by the Tennessee Valley Authority for license to operate Watts Bar Nuclear Plant, Units 1 and 2, Docket Nos. 50-390 and 50-391, located in Rhea County, Tennessee, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation with additional information submitted by the applicant since Supplement No. 13 was issued, and matters that the staff had under review when Supplement No. 13 was issued.
S.f. Kessler - One of the best experts on this subject based on the ideXlab platform.
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Criticality Safety Evaluation Report for the Cold Vacuum Drying (CVD) Facilities Process Water Handling System
2000Co-Authors: S.f. KesslerAbstract:This Report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility. The controls and limitations on equipment design and operations to control potential criticality occurrences are identified.
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Criticality Safety Evaluation Report CSER-96-019 for Spent Nuclear Fuel (SNF) Processing and Storage Facilities Multi Canister Overpack (MCO)
1999Co-Authors: S.f. KesslerAbstract:This criticality Evaluation is for Spent N Reactor fuel unloaded from the existing canisters in both KE and KW Basins, and loaded into multiple canister overpack (MCO) containers with specially built baskets containing a maximum of either 54 Mark IV or 48 Mark IA fuel assemblies. The criticality Evaluations include loading baskets into the cask-MCO, operation at the Cold Vacuum Drying Facility,a nd storage in the Canister Storage Building. Many conservatisms have been built into this analysis, the primary one being the selection of the K{sub eff} = 0.95 criticality Safety limit. This revision incorporates the analyses for the sampling/weld station in the Canister Storage Building and additional analysis of the MCO during the draining at CVDF. Additional discussion of the scrap basket model was added to show why the addition of copper divider plates was not included in the models.
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Criticality Safety Evaluation Report for the multi-canister overpack
1999Co-Authors: S.f. KesslerAbstract:This criticality Evaluation is for Spent N Reactor fuel unloaded from the existing canisters in both KE and KW Basins, and loaded into multiple canister overpack (MCO) containers with specially built baskets containing a maximum of either 54 Mark 1V or 48 Mark IA fuel assemblies. The criticality Evaluations include loading baskets into the cask-MCO, operations at the Cold Vacuum Drying Facility, and storage in the Canister Storage Building. Many conservatisms have been built into this analysis, the primary one being the selection of the k{sub eff} = 0.95 criticality Safety limit. Additional analyses in this revision include partial fuel basket loadings, loading 26.1 inch Mark IA fuel assemblies into Mark IV fuel baskets, and the revised fuel and scrap basket designs. The MCO MCNP model was revised to include the shield plug assembly.
K.n. Schwinkendorf - One of the best experts on this subject based on the ideXlab platform.
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Criticality Safety Evaluation Report for spent nuclear fuelprocessing and storage facilities
1997Co-Authors: K.n. SchwinkendorfAbstract:This criticality Evaluation is for Spent N Reactor fuel unloaded from the existing canisters in both KE and KW Basins, and loaded into multiple canister overpack (MCO) containers with specially- built baskets containing either 54 Mark IV or 48 Mark IA fuel assemblies. The criticality Evaluations include loading baskets into the MCO/Cask, operations at the Cold Vacuum Drying Facility (CVDF), and storage in the Canister Storage Building (CSB). Many conservatisms have been built into this analysis, the primary one being the selection of the k{sub eff} @ 0.95 criticality Safety limit.
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Criticality Safety Evaluation Report for K Basin filter cartridges
1995Co-Authors: K.n. SchwinkendorfAbstract:A criticality Safety Evaluation of the K Basin filter cartridge assemblies has been completed to support operations without a criticality alarm system. The results show that for normal operation, the filter cartridge assembly is far below the Safety limit of k{sub eff} = 0.95, which is applied to plutonium systems at the Hanford Site. During normal operating conditions, uranium, plutonium, and fission and corrosion products in solution are continually accumulating in the available void spaces inside the filter cartridge medium. Currently, filter cartridge assemblies are scheduled to be replaced at six month intervals in KE Basin, and at one year intervals in KW Basin. According to available plutonium concentration data for KE Basin and data for the U/Pu ratio, it will take many times the six-month replacement time for sufficient fissionable material accumulation to take place to exceed the Safety limit of k{sub eff} = 0.95, especially given the conservative assumption that the presence of fission and corrosion products is ignored. Accumulation of sludge with a composition typical of that measured in the sand filter backwash pit will not lead to a k{sub eff} = 0.95 value. For off-normal scenarios, it would require at least two unlikely, independent, and concurrentmore » events to take place before the k{sub eff} = 0.95 limit was exceeded. Contingencies considered include failure to replace the filter cartridge assemblies at the scheduled time resulting in additional buildup of fissionable material, the loss of geometry control from the filter cartridge assembly breaking apart and releasing the individual filter cartridges into an optimal configuration, and concentrations of plutonium at U/Pu ratios less than measured data for KE Basin, typically close to 400 according to extensive measurements in the sand filter backwash pit and plutonium production information.« less
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Criticality Safety Evaluation Report for 300 Area N Reactor fuel fabrication and storage facility
1993Co-Authors: K.n. SchwinkendorfAbstract:Criticality Safety support calculations for the 300 Area Fuels Manufacturing Facility have been performed to update values currently found in the criticality prevention specifications. The old safe masses and associated dimensions table has been updated with new values, which were produced using more modern computer codes that comply with Software Quality Assurance (SQA) requirements. Monte Carlo calculations, using MCNP (Carter 1991), were performed to spot check these results, obtained using the WIMS-E code (Gubbins et. al. 1982). In addition, the criticality considerations associated with certain accident, or upset, conditions were analyzed. These scenarios include fire, the bringing together of multiple safe masses into one neutronically coupled system, misstacking, and accidental interspersed moderation.
J.v. Nelson - One of the best experts on this subject based on the ideXlab platform.
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Criticality Safety Evaluation Report for the cold vacuum drying facility's process water handling system
1999Co-Authors: J.v. NelsonAbstract:This Report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility. The controls and limitations on equipment design and operations to control potential criticality occurrences are identified.
S.d. Roblyer - One of the best experts on this subject based on the ideXlab platform.
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Criticality Safety Evaluation Report for the Cold Vacuum Drying Facility`s process water handling system
1998Co-Authors: S.d. RoblyerAbstract:This Report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility (CVDF). The controls and limitations on equipment design and operations to control potential criticality occurrences are identified. The effectiveness of equipment design and operation controls in preventing criticality occurrences during normal and abnormal conditions is evaluated and documented in this Report. Spent nuclear fuel (SNF) is removed from existing canisters in both the K East and K West Basins and loaded into a multicanister overpack (MCO) in the K Basin pool. The MCO is housed in a shipping cask surrounded by clean water in the annulus between the exterior of the MCO and the interior of the shipping cask. The fuel consists of spent N Reactor and some single pass reactor fuel. The MCO is transported to the CVDF near the K Basins to remove process water from the MCO interior and from the shipping cask annulus. After the bulk water is removed from the MCO, any remaining free liquid is removed by drawing a vacuum on the MCO`s interior. After cold vacuum drying is completed, the MCO is filled with an inert cover gas, the lid is replaced on the shipping cask,more » and the MCO is transported to the Canister Storage Building. The process water removed from the MCO contains fissionable materials from metallic uranium corrosion. The process water from the MCO is first collected in a geometrically safe process water conditioning receiver tank. The process water in the process water conditioning receiver tank is tested, then filtered, demineralized, and collected in the storage tank. The process water is finally removed from the storage tank and transported from the CVDF by truck.« less
