Extended step characteristic model for quarter-core gamma heating calculations
- 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-; 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
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22 GENERAL STUDIES OF NUCLEAR REACTORS
K REACTOR
COMPUTERIZED SIMULATION
REACTOR CORES
RADIATION HEATING
C CODES
COMPARATIVE EVALUATIONS
LOSS OF COOLANT
MONTE CARLO METHOD
REACTOR PHYSICS
ACCIDENTS
CALCULATION METHODS
COMPUTER CODES
EVALUATION
HEATING
HEAVY WATER MODERATED REACTORS
PHYSICS
PRODUCTION REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SIMULATION
SPECIAL PRODUCTION REACTORS
210400* - Power Reactors
Nonbreeding
Otherwise Moderated or Unmoderated
220100 - Nuclear Reactor Technology- Theory & Calculation