Validation of a Method for Prediction of Isotopic Concentrations in Burnup Credit Applications
- Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States). Computational Physics and Engineering Division
Unlike fresh fuel assumptions typically employed in the criticality safety analysis of spent fuel configurations, burnup credit applications rely on depletion and decay calculations to predict the isotopic composition of spent fuel. These isotopics are used in subsequent criticality calculations to assess the reduced worth of the spent fuel. To validate the codes and data used in depletion approaches, experimental measurements are compared with numerical predictions for relevant spent fuel samples. This paper describes a set of experimentally characterized pressurized-water-reactor (PWR) fuel samples and provides a comparison to results of SCALE-4 depletion calculations. An approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations.
- Research Organization:
- Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States). Computational Physics and Engineering Division
- Sponsoring Organization:
- USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
- DOE Contract Number:
- AC05-84OR21400
- OSTI ID:
- 103524
- Report Number(s):
- CONF-9509100-12; ON: DE95017431; TRN: 95:020513
- Resource Relation:
- Conference: ICNC '95: 5. International Conference on Nuclear Criticality Safety, Albuquerque, NM (United States), 17-22 Sep 1995; Other Information: PBD: [1995]
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES
97 MATHEMATICS AND COMPUTING
SPENT FUELS
ISOTOPE RATIO
BURNUP
CRITICALITY
PWR TYPE REACTORS
VERIFICATION
Nuclear Criticality Safety Program (NCSP)
Fresh Fuel Assumptions
Spent Fuel Configurations
Pressurized-Water-Reactor (PWR)
SCALE Code System
Light-Water-Reactor (LWR)
Evaluated Nuclear Data File (ENDF)
SAS2H