Published February 1987 | Version v1
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Accuracy estimation for intermediate and low energy neutron transport calculation with Monte Carlo code MCNP

Creators

  • 1. Japan Atomic Energy Research Inst., Tokai, Ibaraki. Tokai Research Establishment

Description

Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)

Availability note (English)

MF available from INIS under the Report Number.

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Additional details

Publishing Information

Imprint Pagination
32 p.
Report number
JAERI-M--87-010

INIS

Country of Publication
Japan
Country of Input or Organization
Japan
INIS RN
18070563
Subject category
S99: GENERAL AND MISCELLANEOUS; S73: NUCLEAR PHYSICS AND RADIATION PHYSICS;
Descriptors DEI
ACCURACY; CONCRETES; EPITHERMAL NEUTRONS; GRAPHITE; JPDR REACTOR; M CODES; MONTE CARLO METHOD; NEUTRON SPECTRA; NEUTRON TRANSPORT; SHIELDING MATERIALS; THERMAL NEUTRONS
Descriptors DEC
BARYONS; BUILDING MATERIALS; BWR TYPE REACTORS; CARBON; COMPUTER CODES; ELEMENTARY PARTICLES; ELEMENTS; ENRICHED URANIUM REACTORS; EXPERIMENTAL REACTORS; FERMIONS; HADRONS; MATERIALS; NEUTRAL-PARTICLE TRANSPORT; NEUTRONS; NONMETALS; NUCLEONS; POWER REACTORS; RADIATION TRANSPORT; REACTORS; RESEARCH AND TEST REACTORS; SPECTRA; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS