Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 4-Three Mile Island Unit 1 Cycle 5

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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 is to be taken for the reduced reactivity of burned or spent fuel relative to its original ''fresh'' composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using relevant and well-documented critical configurations from commercial pressurized water reactors. The analysis methodology utilized for all calculations in this report is based on the modules and data ... continued below

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DeHart, M.D. January 1, 1995.

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Description

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 is to be taken for the reduced reactivity of burned or spent fuel relative to its original ''fresh'' composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using relevant and well-documented critical configurations from commercial pressurized water reactors. The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Isotopic densities for spent fuel assemblies in the core were calculated using the SCALE-4 SAS2H analytical sequence. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code family was used to extract the necessary isotopic densities from SAS2H results and to provide the data in the format required for SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) for the critical configuration. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all calculations. This volume of the report documents a reactor critical calculation for GPU Nuclear Corporation's Three Mile Island Unit 1 (TMI-1) during hot, zero-power startup testing for the beginning of cycle 5. This unit and cycle were selected because of their relevance in spent fuel benchmark applications: (1) cycle 5 startup occurred after an especially long downtime of 6.6 years; and (2) the core consisted primarily (75%) of burned fuel, with all fresh fuel loaded on the core outer periphery. A k{sub eff} value of 0.9978 {+-} 0.0004 was obtained using two million neutron histories in the KENO V.a model. This result is close to the known critical k{sub eff} of 1.0 for the actual core and is consistent with other mixed-oxide criticality benchmarks. Thus this method is shown to be valid for spent fuel applications in burnup credit analyses.

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  • Report No.: ORNL/TM-12294/V4
  • Grant Number: DE-AC05-00OR22725
  • DOI: 10.2172/885647 | External Link
  • Office of Scientific & Technical Information Report Number: 885647
  • Archival Resource Key: ark:/67531/metadc892713

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  • January 1, 1995

Added to The UNT Digital Library

  • Sept. 23, 2016, 2:42 p.m.

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  • Dec. 1, 2016, 6:39 p.m.

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DeHart, M.D. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 4-Three Mile Island Unit 1 Cycle 5, report, January 1, 1995; [Tennessee]. (digital.library.unt.edu/ark:/67531/metadc892713/: accessed October 20, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.