Naslov
Modeling of the ORNL PCA benchmark using SCALE6.0 hybrid deterministic-stochastic methodology

Autori
Matijević, Mario ; Pevec, Dubravko ; Trontl, Krešimir

Izvornik
Science and Technology of Nuclear Installations (1687-6075) 2013 (2013); 252140-1

Vrsta, podvrsta i kategorija rada
Radovi u časopisima, članak, znanstveni

Ključne riječi
PCA benchmark ; neutron transport ; SCALE6.0 ; reactor pressure vessel ; Monte Carlo ; variance reduction ; adjoint flux ; hybrid method ; multigroup cross sections

Sažetak
Radiation induced damage to the reactor pressure vessel (RPV) by fast neutron fluence is a main factor when considering lifetime extension for nuclear power plants. 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, U.S. Nuclear Regulatory Commission (U.S.NRC) Regulatory Guide 1.190 is the effective guideline for reactor dosimetry calculations, i.e. for correlation of in-core and ex-core experimental measurements with the computational results from computer codes. A well known international shielding database SINBAD contains large selection of models for that purpose, for benchmarking neutron transport methods via stochastic and deterministic computer codes. In this paper a Pool Critical Assembly Pressure Vessel Facility Benchmark (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 (Standardized Computer Analysis for Licensing Evaluation) 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 used 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.

Izvorni jezik
Engleski

Znanstvena područja
Elektrotehnika

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