Contribution to the validation of MCNP neutronics design of the advanced neutron source reactor

Conference · Fri Jan 01 00:00:00 EST 1993 · Transactions of the American Nuclear Society; (United States)

OSTI ID:5802904

Rubio, G A; Ougouag, A M; Wemple, C A; Ryskamp, J M ^[1]

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1. Idaho National Engineering Lab., Idaho Falls (United States)

In the research and development plan of the advanced neutron source (ANS) reactor, one of the steps planned for neutronics design methods validation is to compare computational model results and experimental results for a critical facility exhibiting similarities with the ANS reactor. One such facility is the FOEHN experiment. In this paper, an MCNP model of the FOEHN experiment is developed, and its results are compared with experimental data from the literature. The MCNP models reproduces measured quantities of interest with a high level of agreement.

• OSTI ID: 5802904
• Report Number(s): CONF-930601-; CODEN: TANSAO
• Journal Information: Transactions of the American Nuclear Society; (United States), Vol. 68; Conference: American Nuclear Society (ANS) annual meeting, San Diego, CA (United States), 20-24 Jun 1993; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
REACTOR PHYSICS
MATHEMATICAL MODELS
REACTORS
COMPUTERIZED SIMULATION
DESIGN
MONTE CARLO METHOD
NEUTRON SOURCES
REACTOR CORES
VALIDATION
CALCULATION METHODS
PARTICLE SOURCES
PHYSICS
RADIATION SOURCES
REACTOR COMPONENTS
SIMULATION
TESTING
220600* - Nuclear Reactor Technology- Research
Test & Experimental Reactors
220100 - Nuclear Reactor Technology- Theory & Calculation