CSRL-V ENDF/B-V 227-Group Neutron Cross-Section Library and Its Application to Thermal-Reactor and Criticality Safety Benchmarks

Conference · Mon May 17 04:00:00 EDT 1982

OSTI ID:5236157

Ford, III, W. E. ^[1]; Diggs, B. R. ^[1]; Knight, J. R. ^[1]; Greene, N. M. ^[1]; Petrie, L. M. ^[1]; Webster, C. C. ^[1]; Westfall, R. M. ^[1]; Wright, R. Q. ^[1]; Williams, M. L. ^[1]

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1. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

Characteristics and contents of the CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data) 227-neutron-group AMPX master and pointwise cross-section libraries are described. Results obtained in using CSRL-V to calculate performance parameters of selected thermal reactor and criticality safety benchmarks are discussed.

• Research Organization: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP); US Nuclear Regulatory Commission (USNRC), Office of Nuclear Regulatory Research; US Nuclear Regulatory Commission (USNRC), Office of Nuclear Material Safety and Safeguards; Electric Power Research Inst. (EPRI), Palo Alto, CA (United States)
• DOE Contract Number: W-7405-ENG-26
• OSTI ID: 5236157
• Report Number(s): CONF-820566-6; ON: DE82017497
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220100 -- Nuclear Reactor Technology-- Theory & Calculation
220900* -- Nuclear Reactor Technology-- Reactor Safety
97 MATHEMATICS AND COMPUTING
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CROSS SECTIONS
Cross Section Evaluation Working Group (CSEWG)
Evaluated Nuclear Data File (ENDF)
KINETICS
Lattice Critical Experiment
NUCLEAR DATA COLLECTIONS
Neutron Cross-Section Library
Nuclear Criticality Safety Program (NCSP)
REACTOR KINETICS
REACTOR SAFETY
REACTORS
SAFETY
THERMAL REACTORS
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