Burnup Credit Validation of SCALE-4 Using Light Water Reactor Criticals

Conference · Sun Jun 20 00:00:00 EDT 1993

OSTI ID:7368624

Bowman, S. M. ^[1]; Hermann, O. W. ^[1]; Brady, M. C. ^[2]

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1. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
2. Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

The ANSI/ANS 8.1 criticality safety standard recommends validation and benchmarking of the calculational methods used in evaluating criticality safety limits for away-from-reactor applications. The lack of critical experiments with burned light-water-reactor (LWR) fuel in racks or in casks necessitates the validation of burnup credit methods by comparison to LWR core criticals. These are relevant benchmarks because they test a methodology's ability to predict spent fuel isotopics and to evaluate the reactivity effects of heterogeneities and strong absorbers. Data are available to perform analyses at precise state points. The US Department of Energy Burnup Credit Program has sponsored analysis of selected reactor core critical configurations from commercial pressurized-water-reactors (PWRs) in order to validate an appropriate analysis methodology. The initial methodology used the SCALE-4 code system to analyze a set of reactor critical configurations from Virginia Power's Surry and North Anna reactors. The methodology has since been revised to simplify both the data requirements and the calculational procedure for the criticality analyst. This revised methodology is validated here by comparison to three reactor critical configurations from Tennessee Valley Authority's Sequoyah Unit 2 Cycle 3 and two from Virginia Power's Surry Unit 1 Cycle 2.

• Research Organization: Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• DOE Contract Number: AC05-84OR21400; AC04-76DP00789
• OSTI ID: 7368624
• Report Number(s): CONF-930601-23; ON: DE93015505
• Resource Relation: Conference: American Nuclear Society (ANS) Annual Meeting, San Diego, CA (United States), 20-24 Jun 1993
• Country of Publication: United States
• Language: English

Related Subjects

42 ENGINEERING
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
FUEL ELEMENTS
BURNUP
PWR TYPE REACTORS
SPENT FUELS
BENCHMARKS
CRITICALITY
REACTOR FUELING
SAFETY
SEQUOYAH-2 REACTOR
STANDARDS
SURRY-1 REACTOR
ENERGY SOURCES
ENRICHED URANIUM REACTORS
FUELS
MATERIALS
NUCLEAR FUELS
POWER REACTORS
REACTOR COMPONENTS
REACTOR MATERIALS
REACTORS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
Nuclear Criticality Safety Program (NCSP)
ANSI/ANS 8.1
Light-Water-Reactor (LWR)
420203* - Engineering- Handling Equipment & Procedures
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
220100 - Nuclear Reactor Technology- Theory & Calculation