Validation of SCALE-4 for burnup credit applications

Journal Article · Sat Apr 01 04:00:00 EST 1995 · Nuclear Technology

OSTI ID:46136

Bowman, S M; DeHart, M D; Parks, C V ^[1]

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1. Oak Ridge National Lab., TN (United States). Computing Applications Division

In the past, a criticality analysis of PWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. If credit is allowed for fuel burnup in the design of casks that are used in the transport of spent light water reactor fuel to a repository, the increase in payload can lead to a significant reduction in the cost of transport and a potential reduction in the risk to the public. A portion of the work has been performed at ORNL in support of the US DOE efforts to demonstrate a validation approach for criticality safety methods to be used in burnup credit cask design. To date, the SCALE code system developed at ORNL has been the primary computational tool used by DOE to investigate technical issues related to burnup credit. The ANSI/ANS-8.1 criticality safety standard requires validation and benchmarking of the calculational methods used in evaluating criticality safety limits for applications outside reactors by correlation against critical experiments that are applicable. Numerous critical experiments for fresh PWR-type fuel in storage and transport configurations exist and can be used as part of a validation database. However, there are no critical experiments with burned PWR-type fuel in storage and transport configurations. As an alternative, commercial reactors offer an excellent source of measured critical configurations. The results reported demonstrate the ability of the ORNL SCALE-4 methodology to predict a value of k{sub eff} very close to the known value of 1.0, both for fresh fuel criticals and for the more complex reactor criticals. Beyond these results, additional work in the determination of biases and uncertainties is necessary prior to use in burnup credit applications.

• DOE Contract Number: AC05-84OR21400
• OSTI ID: 46136
• Journal Information: Nuclear Technology, Journal Name: Nuclear Technology Journal Issue: 1 Vol. 110; ISSN 0029-5450; ISSN NUTYBB
• Country of Publication: United States
• Language: English

Related Subjects

05 NUCLEAR FUELS
42 ENGINEERING
99 GENERAL AND MISCELLANEOUS
BURNUP
CRITICALITY
DESIGN
PWR TYPE REACTORS
S CODES
SAFETY ANALYSIS
SPENT FUEL CASKS
SPENT FUEL ELEMENTS
TRANSPORT