Monte Carlo Cross Section Testing for Thermal and Intermediate 235U/238U Critical Assemblies, ENDF/B-V vs ENDF/B-VI
Conference
·
OSTI ID:350945
- Lockheed Martin Corp., Schenectady, NY (United States)
The purpose of this study is to investigate the eigenvalue sensitivity to changes in ENDF/B-V and ENDF/B-VI cross section data sets by comparing RACER vectorized Monte Carlo calculations for several thermal and intermediate spectrum critical experiments. Nineteen Oak Ridge and Rocky Flats thermal solution benchmark critical assemblies that span a range of hydrogen-to-235U (H/U) concentrations (2052 to 27.1) and above-thermal neutron leakage fractions (0.555 to 0.011) were analyzed. In addition, three intermediate spectrum critical assemblies (UH3-UR, UH3-NI, and HISS-HUG) were studied.
- Research Organization:
- Knolls Atomic Power Laboratory (KAPL), Niskayuna, NY (United States)
- Sponsoring Organization:
- USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
- DOE Contract Number:
- AC12-76SN00052
- OSTI ID:
- 350945
- Report Number(s):
- KAPL-P--000242; K--96165; CONF-970607--; ON: DE99002614
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
97 MATHEMATICS AND COMPUTING
99 GENERAL AND MISCELLANEOUS
CROSS SECTIONS
EIGENVALUES
MONTE CARLO METHOD
MULTIPLICATION FACTORS
Monte Carlo Calculations
NUCLEAR DATA COLLECTIONS
Nuclear Criticality Safety Program (NCSP)
R CODES
RACER
REACTOR KINETICS
URANIUM 235
URANIUM 238
ZERO POWER REACTORS
97 MATHEMATICS AND COMPUTING
99 GENERAL AND MISCELLANEOUS
CROSS SECTIONS
EIGENVALUES
MONTE CARLO METHOD
MULTIPLICATION FACTORS
Monte Carlo Calculations
NUCLEAR DATA COLLECTIONS
Nuclear Criticality Safety Program (NCSP)
R CODES
RACER
REACTOR KINETICS
URANIUM 235
URANIUM 238
ZERO POWER REACTORS