Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 1-Summary
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 critical configurations from commercial pressurized- water reactors (PWR). 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 SAS2H analytical sequence in SCALE-4. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code sequence was used to extract the necessary isotopic densities from SAS2H results and to provide the data in the format required for SCALE-4 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 analysis of each critical configuration. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the Cycle 2 core; Volume 4 documents the calculations performed based on GPU Nuclear Corporation's Three Mile Island Unit 1 Cycle 5 core; and, lastly, Volume 5 describes the analysis of Virginia Power's North Anna Unit 1 Cycle 5 core. Each of the reactor-specific volumes provides the details of calculations performed to determine the effective multiplication factor for each reactor core for one or more critical configurations using the SCALE-4 system; these results are summarized in this volume. Differences between the core designs and their possible impact on the criticality calculations are also discussed. Finally, results are presented for additional analyses performed to verify that solutions were sufficiently converged. All calculations show the ability to predict a k{sub eff} value very close to 1.0 for various conditions and cooling times. Thus, the methodology applied is shown to be a valid approach for calculating the value of k{sub eff} for systems with spent PWR fuel.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Organization:
- US Department of Energy (US)
- DOE Contract Number:
- AC05-00OR22725
- OSTI ID:
- 814373
- Report Number(s):
- ORNL/TM-12294/V1; TRN: US0304203
- Resource Relation:
- Other Information: PBD: 1 Jan 1995
- Country of Publication:
- United States
- Language:
- English
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Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 3-Surry Unit 1 Cycle 2
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
ACTINIDES
BENCHMARKS
COOLING TIME
CRITICALITY
CROSS SECTIONS
FISSION
MULTIPLICATION FACTORS
PWR TYPE REACTORS
REACTIVITY WORTHS
REACTOR CORES
RESONANCE
SPENT FUELS
WATER