PWR and BWR Pressure Vessel Fluence Calculation Benchmark Problems and Solutions (NUREG/CR-6115)
On this page:
Download complete document
Manuscript Completed: September 2001
Date Published: September 2001
J.F. Carew, K. Hu, A. Aronson, A. Prince, G. Zamonsky
Brookhaven National Laboratory
Upton, NY 11973-5000
W.R. Jones, NRC Project Manager
Division of Engineering Technology
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
NRC Job Code W6968
In order to provide the high confidence required in the determination of pressure vessel embrittlement and the evaluation of the Pressurized Thermal Shock (PTS) screening criteria, Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," recommends detailed benchmarking of pressure vessel fluence calculations. In response to this recommendation, this report provides the detailed specification and corresponding numerical solutions for a set of PWR and BWR pressure vessel fluence benchmark problems. PWR benchmark problems have been specified for (1) a standard core loading pattern, (2) a low-leakage core loading pattern and (3) a partial length shield assembly core design. Since BWR fuel loading patterns are presently not being designed for vessel fluence reduction, only a single BWR problem is specified. These benchmark problems allow a detailed assessment of the numerical procedures, code implementation, and the various modeling approximations against state-of-the-art solutions for representative operating configurations.
The problems have been solved using the DORT discrete ordinates transport code, the MESH source processing code, and the BUGLE-93 ENDF/B-VI nuclear data. Detailed neutron flux solutions are provided at selected pressure vessel azimuthal, radial, and axial locations. A pin-wise core neutron source is provided which includes the fuel burnup dependence of both the magnitude and energy dependence of the neutron source. Dosimeter responses are calculated for a typical set of LWR fast neutron dosimeter materials. In addition, MCNP Monte Carlo calculations have been performed for the PWR standard core and partial length shield assembly core loadings and the BWR problem. Comparisons of the MCNP and DORT vessel fluence calculations are presented.