A Monte Carlo method of evaluating heterogeneous effects in plate-fueled reactors

Conference · Tue Jan 01 00:00:00 EST 1991 · Transactions of the American Nuclear Society; (United States)

OSTI ID:5854787

Thayer, R C; Redmond, II, E L; Ryskamp, J M ^[1]

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

Few-group nuclear cross sections for small plate-fueled, light and heavy water test reactors are frequently generated with unit cell models that contain a homogeneous mixture of fuel, cladding, and water. The heterogeneous unit cells do not need to be represented explicitly for neutronics calculations when the plate and coolant channel thicknesses are small compared with the mean-free-path of neutrons. However, neutron and photon heating calculations were performed with heterogeneous fuel models to predict accurately the heat deposited in the fuel meat, cladding, and coolant. Heat deposited in the coolant channels and outside the fuel elements does not have a direct impact on the peak fuel meat temperature but must be included in the total coolant system heat balance. The results of a heterogeneous Monte Carlo calculation that estimates the heat loads in different fuel regions are presented and the fact that similar homogeneous fuel models can be used for many calculations. The calculations presented here were performed on models of the Advanced Neutron Source (ANS) and the Massachusetts Institute of Technology Reactor 2 (MITR-2). The ANS is a small, 362-MW (fission), plate-fueled, heavy water reactor designed to produce an intense steady-state source of neutrons.

• OSTI ID: 5854787
• Report Number(s): CONF-910603-; CODEN: TANSA
• Journal Information: Transactions of the American Nuclear Society; (United States), Vol. 63; Conference: Annual meeting of the American Nuclear Society (ANS), Orlando, FL (United States), 2-6 Jun 1991; ISSN 0003-018X
• Country of Publication: United States
• Language: English

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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
MITR REACTOR
REACTOR CORES
HETEROGENEOUS EFFECTS
M CODES
RESEARCH AND TEST REACTORS
C CODES
CALCULATION METHODS
CROSS SECTIONS
FUEL ASSEMBLIES
FUEL CANS
FUEL ELEMENTS
FUEL PLATES
HEAVY WATER
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NEUTRON SOURCES
SENSITIVITY ANALYSIS
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ACTINIDE ISO
ACTINIDE NUCLEI
ALPHA DECAY RADIOISOTOPES
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ENRICHED URANIUM REACTORS
EVEN-ODD NUCLEI
HEAVY NUCLEI
HEAVY WATER COOLED REACTORS
HEAVY WATER MODERATED REACTORS
HYDROGEN COMPOUNDS
INTERNAL CONVERSION RADIOISOTOPES
ISOMERIC TRANSITION ISOTOPES
ISOTOPES
MINUTES LIVING RADIOISOTOPES
NUCLEI
OXYGEN COMPOUNDS
PARTICLE SOURCES
RADIATION SOURCES
RADIOISOTOPES
REACTOR COMPONENTS
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TRAINING REACTORS
URANIUM ISOTOPES
220600* - Nuclear Reactor Technology- Research
Test & Experimental Reactors