Benchmarking of a Generic CANDU Reactor with PARCS, MCNP and RFSP (NUREG/IA-0453)

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Publication Information

Manuscript Completed: June 2015
Date Published: September 2015

Prepared by:
B. Arsenault, O. Shaikh, T. Downar*, D. Jabaay*, A. Ward*, Y. Xu*,

AMEC-NSS,
4th Floor, 700 University Ave, Toronto, Ontario M5G1X6, Canada

*University of Michigan, Nuclear Eng. and Rad. Science
1934 Cooley, Ann Arbor, MI 48109-2104, USA

K. Tien, NRC Project Manager

Division of Systems Analysis
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Prepared as part of:
The Agreement on Research Participation and Technical Exchange
Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by:
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Availability Notice

Abstract

The purpose of this project is to benchmark a generic CANDU reactor with the computer codes PARCS, MCNP and RFSP. RFSP is a diffusion code used by the Canadian nuclear industry for fuel management and for safety analyses. A number of changes were made to PARCS and GenPMAXS, the utility used to generate the fuel tables for PARCS. The changes were required to allow the modeling of specific aspects related to CANDU reactors.

More specifically, the benchmarking exercise involves an assessment of the Finite Difference Method and the hybrid NEM/ANM solvers available in PARCS against the results of Monte Carlo simulations performed with MCNP. The results of the simulations calculated with the Finite Difference Methods in PARCS and RFSP are compared to determine possible areas of improvement in RFSP.

This work is an in-kind contribution and has been funded by Amec Foster Wheeler.