SCALE-4 analysis of pressurized water reactor critical configurations. Volume 3: Surry Unit 1 Cycle 2
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 selected critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations in this report is based on the codes and data provided in the SCALE-4 code system. This volume of the report documents the SCALE system analysis of two reactor critical configurations for Surry Unit 1 Cycle 2. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted a direct comparison of criticality calculations using the utility-calculated isotopics, with those using, the isotopics generated by the SCALE-4 SAS2H sequence. These reactor critical benchmarks have been reanalyzed using the methodology described above. The two benchmark critical calculations were the beginning-of-cycle (BOC) startup at hot, zero-power (HZP) and an end-of-cycle (EOC) critical at hot, full-power (HFP) critical conditions. These calculations were used to check for consistency in the calculated results for different burnup, downtime, temperature, xenon, and boron conditions. The k{sub eff} results were 1.0014 and 1.0113, respectively, with a standard deviation of 0.0005.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Organization:
- USDOE, Washington, DC (United States)
- DOE Contract Number:
- AC05-84OR21400
- OSTI ID:
- 71346
- Report Number(s):
- ORNL/TM-12294/V.3; ON: DE95010428; TRN: 95:015200
- Resource Relation:
- Other Information: PBD: Mar 1995
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
05 NUCLEAR FUELS
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS
SURRY-1 REACTOR
CRITICALITY
SPENT FUELS
BENCHMARKS
TRANSPORT
SPENT FUEL STORAGE
SPENT FUEL CASKS
SAFETY ANALYSIS
CALCULATION METHODS
S CODES
MULTIPLICATION FACTORS
BURNUP
REACTOR CORES
THEORETICAL DATA
FUEL RACKS