Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 2-Sequoyah Unit 2 Cycle 3

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The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations reported herein is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in ... continued below

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Bowman, S.M. January 1, 1995.

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The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations reported herein is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of three reactor critical configurations for the Sequoyah Unit 2 Cycle 3. This unit and cycle were chosen because of the relevance in spent fuel benchmark applications: (1) the unit had a significantly long downtime of 2.7 years during the middle of cycle (MOC) 3, and (2) the core consisted entirely of burned fuel at the MOC restart. The first benchmark critical calculation was the MOC restart at hot, full-power (HFP) critical conditions. The other two benchmark critical calculations were the beginning-of-cycle (BOC) startup at both hot, zero-power (HZP) and HFP critical conditions. These latter calculations were used to check for consistency in the calculated results for different burnups and downtimes. The k{sub eff} results were in the range of 1.00014 to 1.00259 with a standard deviation of less than 0.001.

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  • Report No.: ORNL/TM-12294/V2
  • Grant Number: DE-AC05-00OR22725
  • DOI: 10.2172/885839 | External Link
  • Office of Scientific & Technical Information Report Number: 885839
  • Archival Resource Key: ark:/67531/metadc876240

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  • January 1, 1995

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  • Sept. 21, 2016, 2:29 a.m.

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  • Dec. 1, 2016, 6:39 p.m.

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Bowman, S.M. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 2-Sequoyah Unit 2 Cycle 3, report, January 1, 1995; [Tennessee]. (digital.library.unt.edu/ark:/67531/metadc876240/: accessed August 23, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.