%0 Journal Article
%T Modeling of the ORNL PCA Benchmark Using SCALE6.0 Hybrid Deterministic-Stochastic Methodology
%A Mario Matijevi£¿
%A Dubravko Pevec
%A Kre£¿imir Trontl
%J Science and Technology of Nuclear Installations
%D 2013
%I Hindawi Publishing Corporation
%R 10.1155/2013/252140
%X Revised guidelines with the support of computational benchmarks are needed for the regulation of the allowed neutron irradiation to reactor structures during power plant lifetime. Currently, US NRC Regulatory Guide 1.190 is the effective guideline for reactor dosimetry calculations. A well known international shielding database SINBAD contains large selection of models for benchmarking neutron transport methods. In this paper a PCA benchmark has been chosen from SINBAD for qualification of our methodology for pressure vessel neutron fluence calculations, as required by the Regulatory Guide 1.190. The SCALE6.0 code package, developed at Oak Ridge National Laboratory, was used for modeling of the PCA benchmark. The CSAS6 criticality sequence of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, as well as MAVRIC/Monaco hybrid shielding sequence, was utilized for calculation of equivalent fission fluxes. The shielding analysis was performed using multigroup shielding library v7_200n47g derived from general purpose ENDF/B-VII.0 library. As a source of response functions for reaction rate calculations with MAVRIC we used international reactor dosimetry libraries (IRDF-2002 and IRDF-90.v2) and appropriate cross-sections from transport library v7_200n47g. The comparison of calculational results and benchmark data showed a good agreement of the calculated and measured equivalent fission fluxes. 1. Introduction Calculational methods for determining the neutron fluence are necessary to estimate the fracture toughness of the reactor pressure vessel (RPV) materials, which is one of the key requirements in determining operational limits and lifetime of nuclear power plants. This area of research is of a particular importance in the era of plant lifetime extension demands and possible financial savings which could be achieved by approving the extensions. Any developed or analyzed calculation methodology requires comprehensive verification and validation against evaluated reference data. A large database of benchmarks aimed at validation of computer codes and nuclear data used for radiation transport and shielding problems is ¡°Shielding Integral Benchmark Archive and Database (SINBAD)¡± [1]. One of the most widely used SINBAD benchmarks for qualification of radiation transport methods and evaluation of appropriate nuclear data used for transport as well as for dosimetry calculations in Light Water Reactors (LWR) is the ¡°Pool Critical Assembly Pressure Vessel Facility Benchmark¡± (PCA benchmark) [2]. The purpose of the benchmark was to validate the
%U http://www.hindawi.com/journals/stni/2013/252140/