ENDF/B dosimetry cross section file benchmark neutron flux-spectral uncertainties

Conference · Wed Oct 01 04:00:00 EDT 1975 · Natl. Bur. Stand. (U.S.), Spec. Publ.; (United States)

OSTI ID:7365019

McElroy, W N

An ENDF/B file of evaluated energy dependent cross sections for dosimetry applications has been established. The fission and most reliable non-fission reaction cross sections on this file are used with current recommended sets of evaluated measured reaction rates for several benchmark spectra to establish multiple foil derived flux-spectra with Monte Carlo uncertainties for comparison with spectrometry and calculated spectra. It is concluded that integral data testing of cross sections on the ENDF/B-IV file is presently limited to the +-5 to 10 percent (1 sigma) range because of uncertainties in the benchmark flux-spectra. 1 figure, 1 table. (auth)

• Research Organization: Hanford Engineering Development Lab., Richland, WA
• OSTI ID: 7365019
• Conference Information: Journal Name: Natl. Bur. Stand. (U.S.), Spec. Publ.; (United States) Journal Volume: 1
• Country of Publication: United States
• Language: English

Related Subjects

651000* -- Nuclear Physics
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
BARYON REACTIONS
CROSS SECTIONS
ERRORS
FISSION
HADRON REACTIONS
KINETICS
MONTE CARLO METHOD
NEUTRON FLUX
NEUTRON REACTIONS
NEUTRON SPECTRA
NUCLEAR DATA COLLECTIONS
NUCLEAR REACTION KINETICS
NUCLEAR REACTIONS
NUCLEON REACTIONS
RADIATION FLUX
REACTION KINETICS
SPECTRA