This report deals with radiation transport calculations supporting safety analysis of nuclear reactors. The calculations employ Monte Carlo with the MCNP code and use the DSA variance reduction methodology. The first part of the report contains an analysis of ex-core responses, in particular pressure vessel damage, in a Gen-III PWR. Results from two approaches are compared. The first approach involves a decoupling of the problem. An analog simulation of the eigenvalue problem is performed and a fission source is created. This source is then used in a fixed source calculation with variance reduction parameters generated with the DSA. The second approach calculates the ex-core responses within the eigenvalue problem, again employing the DSA. The second part of the report deals with the PCA-Replica benchmark (SINBAD archive NEA-1517/93) concerning integral experiments performed at the NESTOR research reactor (Winfrith, UK). A fission plate irradiates a mock-up reproducing key materials for LWRs. The results are neutron spectra and dosimeter responses. Various nuclear data libraries are compared. The results show a satisfactory agreement between calculation and experiment.
|Titolo:||Development and Application of Monte Carlo Neutronics Methodologies for Safety Studies of Current Operating Reactors|
|Data di pubblicazione:||8-nov-2017|
|Appare nelle tipologie:||5.2 Documento in Garanzia della Qualità|