Extended step characteristic model for quarter-core gamma heating calculations
Conference
·
· Transactions of the American Nuclear Society; (United States)
OSTI ID:5804957
- Westinghouse Savannah River Company, Aiken, SC (United States)
Discrete ordinates codes are seldom used in lattice or core calculation, because of their limitation to simple geometries, which can be represented using an orthogonal mesh in a given coordinate system. Rough geometric approximations are often applied to obtain an estimate for a flux distribution. However, other methods, such as integral transport or Monte Carlo approaches, are generally more suited to irregular geometries. Each of these methods has its own weaknesses: integral transport methods are limited to problems in which the angular variation of the flux is isotropic or linearly anisotropic; Monte Carlo methods can be time consuming. The extended step characteristic (ESC) method has been developed to apply the discrete ordinates approximation to complicated geometries for which other methods provide less satisfactory solutions. The CENTAUR code has been developed to solve the two-dimensional transport equation using the ESC approach. This paper presents results of CENTAUR calculations for a quarter-core gamma redistribution problem for the Savannah River site (SRS) K reactor, under drained tank conditions following a postulated double-ended guillotine break loss-of-coolant accident. The calculations were used to confirm TWOTRAN calculations, which were based on a coarse approximation of the core geometry. A comparison of the results serves to demonstrate the capabilities and efficiency of the ESC approach.
- OSTI ID:
- 5804957
- Report Number(s):
- CONF-930601--
- Conference Information:
- Journal Name: Transactions of the American Nuclear Society; (United States) Journal Volume: 68
- Country of Publication:
- United States
- Language:
- English
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22 GENERAL STUDIES OF NUCLEAR REACTORS
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C CODES
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COMPARATIVE EVALUATIONS
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MONTE CARLO METHOD
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210400* -- Power Reactors
Nonbreeding
Otherwise Moderated or Unmoderated
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220100 -- Nuclear Reactor Technology-- Theory & Calculation
ACCIDENTS
C CODES
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COMPARATIVE EVALUATIONS
COMPUTER CODES
COMPUTERIZED SIMULATION
EVALUATION
HEATING
HEAVY WATER MODERATED REACTORS
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MONTE CARLO METHOD
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PRODUCTION REACTORS
RADIATION HEATING
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORES
REACTOR PHYSICS
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SPECIAL PRODUCTION REACTORS