A Monte Carlo Green's function method for three-dimensional neutron transport
- Knolls Atomic Power Lab., Schenectady, NY (United States)
This paper describes a Monte Carlo transport kernel capability, which has recently been incorporated into the RACER continuous-energy Monte Carlo code. The kernels represent a Green's function method for neutron transport from a fixed-source volume out to a particular volume of interest. This method is very powerful transport technique. Also, since kernels are evaluated numerically by Monte Carlo, the problem geometry can be arbitrarily complex, yet exact. This method is intended for problems where an ex-core neutron response must be determined for a variety of reactor conditions. Two examples are ex-core neutron detector response and vessel critical weld fast flux. The response is expressed in terms of neutron transport kernels weighted by a core fission source distribution. In these types of calculations, the response must be computed for hundreds of source distributions, but the kernels only need to be calculated once. The advance described in this paper is that the kernels are generated with a highly accurate three-dimensional Monte Carlo transport calculation instead of an approximate method such as line-of-sight attenuation theory or a synthesized three-dimensional discrete ordinates solution.
- OSTI ID:
- 7057459
- Report Number(s):
- CONF-920606-; CODEN: TANSA
- Journal Information:
- Transactions of the American Nuclear Society; (United States), Vol. 65; Conference: American Nuclear Society annual meeting, Boston, MA (United States), 7-12 Jun 1992; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
REACTOR PHYSICS
MONTE CARLO METHOD
BOLTZMANN EQUATION
DISCRETE ORDINATE METHOD
GREEN FUNCTION
KERNELS
NEUTRON DETECTORS
NEUTRON TRANSPORT
PRESSURE VESSELS
REACTOR CORES
THREE-DIMENSIONAL CALCULATIONS
CONTAINERS
DIFFERENTIAL EQUATIONS
EQUATIONS
FUNCTIONS
MEASURING INSTRUMENTS
NEUTRAL-PARTICLE TRANSPORT
PARTIAL DIFFERENTIAL EQUATIONS
PHYSICS
RADIATION DETECTORS
RADIATION TRANSPORT
REACTOR COMPONENTS
220100* - Nuclear Reactor Technology- Theory & Calculation