Whole-Core Neutronics Modeling of a TRIGA Reactor using Integral Transport Theory
An innovative analysis approach for performing criticality calculations of various misload configurations for a TRIGA reactor has been employed recently at the Westinghouse Hanford Company. A deterministic transport theory model with sufficient geometric complexity to evaluate asymmetric loading patterns was used. Calculations of this complexity have been performed in the past using Monte Carlo simulation, such as the MCNP code. However, the Monte Carlo calculations are more difficult to prepare and require more computer time. On the Hanford site CRAY XMP-18 computer, the new methods required less than one-third of the central processing unit time per calculation as compared with an MCNP calculation using 100,000 neutron histories. The calculations were validated to actual core measurements, and very good agreement was achieved. Currently, the Hanford site NRF TRIGA reactor is being put into a nonoperational mode. One of the requirements accompanying this decision was to show computationally that the proposed downloaded core configuration should be substantially subcritical. To accomplish this, a series of criticality computations were undertaken using the latest version of the British neutron transport theory code, WIMS.
- Sponsoring Organization:
- USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
- OSTI ID:
- 6932490
- Report Number(s):
- CONF-901101-; CODEN: TANSAO
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 62; Conference: American Nuclear Society (ANS) Winter Meeting , Washington, DC (United States), 11-16 Nov 1990; ISSN 0003-018X
- Publisher:
- American Nuclear Society
- Country of Publication:
- United States
- Language:
- English
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FUEL MANAGEMENT
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COMPUTER CODES
COMPUTERIZED SIMULATION
CONTROL ELEMENTS
CRITICALITY
M CODES
MONTE CARLO METHOD
NEUTRON DIFFUSION EQUATION
NEUTRON RADIOGRAPHY
NEUTRON TRANSPORT THEORY
REACTOR CORES
TWO-DIMENSIONAL CALCULATIONS
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DIFFERENTIAL EQUATIONS
ENRICHED URANIUM REACTORS
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INDUSTRIAL RADIOGRAPHY
IRRADIATION REACTORS
MATERIALS TESTING REACTORS
PHYSICS
REACTOR COMPONENTS
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RESEARCH AND TEST REACTORS
SIMULATION
SOLID HOMOGENEOUS REACTORS
TRANSPORT THEORY
TRIGA TYPE REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
Nuclear Criticality Safety Program (NCSP)
Analysis Approach
Criticality Calculations
Misload Configurations for a TRIGA Reactor
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