TRACG Statistical Method for Analysis of Anticipated Operational Occurrences

Conference · Mon Jul 01 04:00:00 EDT 2002

OSTI ID:21167848

Bolger, Francis T; Heck, Charles L; Andersen, Jens G.M. ^[1]

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1. GE Nuclear Energy, M/C A33, Castle Hayne Road, Wilmington, NC, 28402 (United States)

This paper discusses the application of TRACG, the General Electric (GE) proprietary version of the Transient Reactor Analysis Code, to analyses of Anticipated Operational Occurrences (AOOs) for boiling water reactors (BWR). Realistic calculations with TRACG can be used together with statistical quantification of uncertainties to support licensing evaluations for these transient events. The approach follows the code scaling applicability and uncertainty (CSAU) methodology consistent with current regulatory guidelines. TRACG applications offer the benefit of more accurate simulations of BWR events and improved operating margins. TRACG uses advanced realistic one-dimensional and three-dimensional methods to model the phenomena that are important in evaluating the operation of BWRs. TRACG has a multi-dimensional, two-fluid model for the reactor thermal hydraulics and a three-dimensional reactor kinetics model consistent with the GE 3D core simulator PANACEA. These features allow for detailed, realistic simulation of a wide range of BWR phenomena. The models have been used to simulate a large variety of test and reactor configurations. Some of the key benchmark comparisons of TRACG with test data will be presented. AOO events are analyzed to establish the reactor system response, including the calculation of the Operating Limit Minimum Critical Power Ratio (OLMCPR). The application of TRACG is also used to demonstrate that acceptable fuel design limits and reactor coolant pressure boundary design conditions are not exceeded during an AOO. This paper describes the method of quantification of uncertainties as applied to the realistic nominal TRACG analyses. The uncertainties and biases in the models and plant parameters are evaluated using a Monte Carlo method to predict a statistical limit for the critical safety parameters. The uncertainties and biases considered include the model uncertainties, experimental uncertainties, plant uncertainties, process measurement errors, and manufacturing tolerances. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 21167848
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
42 ENGINEERING
BWR TYPE REACTORS
COMPARATIVE EVALUATIONS
COMPUTERIZED SIMULATION
DESIGN
MONTE CARLO METHOD
ONE-DIMENSIONAL CALCULATIONS
REACTOR KINETICS
REACTOR SIMULATORS
THERMAL HYDRAULICS
THREE-DIMENSIONAL CALCULATIONS