Comparing fast neutron transport calculations using code package KATRIN-2.0 for various options of VVER-440 core setup

Abstract

Calculation of radiation exposure for VVER reactor vessel is a key task for the following determination of their burn-up life. This problem is especially topical for VVER-440 reactors of the first generation which now require substantiation report of safe operation over-designed service life. The estimation of radiation exposure of the reactor vessel involves the calculation of fast neutron fluence (E≥0.5 MeV) on welded joints and basis metal of the reactor vessel. The result of neutron fluence calculation and the associated estimation error are affected by the precision of fission neutron source definition in the reactor core. This study examines two approaches to defining fission neutron source in the fuel assembly: specifically, assembly-wise distribution, as well as rod-wise distribution approaches. The main objective of this work is to characterize the difference in calculation results of neutron fluence (E≥0.5 MeV) on the VVER-440 reactor vessel at rod-wise and assembly-wise definition of the fission neutron source in peripheral fuel assemblies, and to estimate the accuracy of fluence calculation for each source definition method, by comparing them with some experimental data. The present study examines the results of two different experiments carried out on the block of No.1 of Kola Nuclear Power Plant (V-230): each involving activity measurement for templates cut out on the internal reactor vessel surface and activity measurement of neutron-activation detectors of Niobium on the external reactor vessel surface.