Whole-Core Neutronics Modeling of a TRIGA Reactor using Integral Transport Theory

Conference · Sun Nov 11 04:00:00 EST 1990 · Transactions of the American Nuclear Society

OSTI ID:6932490

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--
• Conference Information: Journal Name: Transactions of the American Nuclear Society Journal Volume: 62
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
220600* -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
97 MATHEMATICS AND COMPUTING
ACCURACY
Analysis Approach
CALCULATION METHODS
COMPUTER CODES
COMPUTERIZED SIMULATION
CONTROL ELEMENTS
CRITICALITY
Criticality Calculations
DIFFERENTIAL EQUATIONS
ENRICHED URANIUM REACTORS
EQUATIONS
FUEL MANAGEMENT
HOMOGENEOUS REACTORS
HYDRIDE MODERATED REACTORS
INDUSTRIAL RADIOGRAPHY
IRRADIATION REACTORS
M CODES
MATERIALS TESTING REACTORS
MONTE CARLO METHOD
Misload Configurations for a TRIGA Reactor
NEUTRON DIFFUSION EQUATION
NEUTRON RADIOGRAPHY
NEUTRON TRANSPORT THEORY
Nuclear Criticality Safety Program (NCSP)
PHYSICS
REACTOR COMPONENTS
REACTOR CORES
REACTOR PHYSICS
REACTORS
RESEARCH AND TEST REACTORS
SIMULATION
SOLID HOMOGENEOUS REACTORS
TRANSPORT THEORY
TRIGA TYPE REACTORS
TRIGA-1-HANFORD REACTOR
TWO-DIMENSIONAL CALCULATIONS
W CODES
WATER COOLED REACTORS
WATER MODERATED REACTORS