Issues Related to Criticality Safety Analysis for Burnup Credit Applications

Conference · Sun Sep 17 00:00:00 EDT 1995

OSTI ID:187231

DeHart, M. D. ^[1]; Parks, C. V. ^[1]

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1. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States). Computational Physics and Enginering Division

Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh fuel loading assumption. Parametric analyses are required to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models are evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. This paper discusses the results of studies to determine the effect of two important modeling assumptions on the criticality analysis of pressurized-water reactor (PWR) spent fuel: (1) the effect of assumed burnup history (i.e., specific power during and time-dependent variations in operational power) during depletion calculations, and (2) the effect of axial burnup distributions on the neutron multiplication factor calculated for a three-dimensional (3-D) conceptual cask design.

• Research Organization: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States). Computational Physics and Enginering Division
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 187231
• Report Number(s): CONF-9509100-35; ON: DE96003018
• 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

Related Subjects

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
SPENT FUELS
BURNUP
CRITICALITY
PWR TYPE REACTORS
REACTIVITY
SAFETY ANALYSIS
SPENT FUEL CASKS
DESIGN
FUEL ASSEMBLIES
Nuclear Criticality Safety Program (NCSP)
Spent Fuel Transportation
Pressurized-Water-Reactor (PWR) Spent Fuel
Three-Dimensional (3-D) Conceptual Cask Design
SCALE 4.2
SAS2H
CSAS
Light-Water-Reactor (LWR) Fuel Assembly
3-D Monte Carlo Calculations
Evaluated Nuclear Data File (ENDF)