Calculations of fast-reactor data-testing assemblies utilizing cross sections from the MC$sup 2$ and SDX codes
Reactor physics calculations were performed with ENDF/B Version-III cross-section data for several of the fast-reactor data-testing assemblies specified by the Cross Section Evaluation Working Group. In these calculations, multigroup cross sections were generated using both the Argonne MC/sup 2/ and SDX codes for comparative purposes. The multigroup cross sections were then used in S/sub n/ transport-theory calculations to obtain k/sub eff/ and central activation rattos, and in perturbation-theory calculations to obtain central worths. Results with MC/sup 2/ and SDX cross sections are in good agrement except whem regions containing large amounts of a structure material are involved. (auth)
- Research Organization:
- Argonne National Lab., IL
- NSA Number:
- NSA-29-017708
- OSTI ID:
- 4376736
- Journal Information:
- Nucl. Sci. Eng., v. 52, no. 4, pp. 486-492, Journal Name: Nucl. Sci. Eng., v. 52, no. 4, pp. 486-492; ISSN NSENA
- Country of Publication:
- Country unknown/Code not available
- Language:
- English
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