CSRL-V ENDF/B-V Library and Thermal Reactor and Criticality Safety Benchmarks

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

OSTI ID:5340325

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

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

CSRL-V, an ENDF/B-V 227-group neutron cross-section library which has recently been expanded to include Bondarenko factor data for unresolved resonance processing, was used to calculate performance parameters for a series of thermal reactor and criticality safety benchmarks. Among the thermal benchmarks calculated were the Babcock and Wilcox lattice critical experiments B and W-XIII and B and W-XX. These two slightly enriched (2.46%) UO₂, water-moderated, tight-pitch lattice experiments were chosen because (a) they have similar U²³⁸ resonance shielding characteristics as power reactor cores, and (b) they provide benchmark results representative of high-leakage and low-leakage lattices, respectively. Among the criticality safety benchmarks calculated were homogeneous, highly enriched (93.2%) uranyl fluoride spheres with hydrogen-to-uranium ratios varying from 76 to 972.

• 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: 5340325
• Report Number(s): CONF-820566-2; ON: DE82015684; TRN: 82-015878
• Resource Relation: Conference: Thermal Reactor Benchmark Calculations, Techniques, Results and Applications Seminar, Upton, NY (United States), 17-18 May 1982; Other Information: Portions of document are illegible
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
97 MATHEMATICS AND COMPUTING
NUCLEAR DATA COLLECTIONS
REACTOR KINETICS
CROSS SECTIONS
KINETICS
Nuclear Criticality Safety Program (NCSP)
Babcock
Wilcox
Lattice Critical Experiment
Evaluated Nuclear Data File (ENDF)
Bondarenko Factor Data
Neutron Cross-Section Library
220100* - Nuclear Reactor Technology- Theory & Calculation