Abstract
The estimation of the neutron fluence at the Reactor Pressure Vessel (RPV) is classically carried out by a two-step approach. The first step is to estimate the full core neutron source term whether the second step of the calculation consists in the transport of neutrons from the core (source term) to the RPV using the neutron fission distribution determined in the previous step. For this purpose, the neutron fission distribution is to be accurately determined at the fuel pin level for the assemblies on the border of the core. To achieve this goal, two methods are evaluated in this study. The first method considered is a full core 2D Monte Carlo calculation using the MNCP6 code. The second method is based on a deterministic approach using the CASMO5 multi-segment option, allowing a full 2D transport calculation at the pin level with an expected accuracy similar to a stochastic method. The comparison of the two methods shows an overall good agreement with differences within the statistical uncertainty for different cores: homogeneous UOX core, mixed UOX-MOX loading and the effect of the hafnium rods used in the assemblies in the periphery of the core. The modelling limitation and the associated calculational time are discussed for the comparison of the two approaches.
Highlights
The lifetime of a nuclear reactor is above others related to the ageing of the reactor pressure vessel under neutron irradiation
The second step of the calculation consists in the transport of neutrons from the core to the reactor pressure vessel (RPV) using the neutron fission distribution determined in the previous step, a standard procedure is to combine a stochastic calculation to evaluate the neutron attenuation with a variance reduction technique
This paper evaluates two procedures to estimate the neutron source term at a pin level based on Monte Carlo (MCNP) or deterministic transport method (CASMO)
Summary
The lifetime of a nuclear reactor is above others related to the ageing of the reactor pressure vessel under neutron irradiation. The first step is to estimate the full core neutron source term, either Monte Carlo [1] or deterministic calculation [2] [3] or hybrid methods [4] can be used. Because the fluence at the vessel is strongly related to the fissions occurring at the periphery of the core, mainly the last two rows of assemblies contribute to the neutron fluence at RPV. Inside these two rows of assemblies, the fuel pins near to the core periphery are of higher importance than the pins closer to the center of the core. To achieve this goal, two methods are evaluated in this study. Impact of the multi-group treatment approach will be evaluated in this study
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