Computer Package for Calculating the Fast-Neutron Fluence in Reactor Pit Based on the Reactor Operational History and the Monte-Carlo Method

Abstract

The systematical calculation method and models as well as the computer package for calculating the fast-neutron fluence in reactor pit based on the reactor operational history and the Monte-Carlo method are developed. The main features of the computer package are as followings: (1) The neutron cross-sections and nuclear density for each types of fuel assemblies and the reflectors are calculated by APOLLO2-F code; (2) The critical boron-concentration searches and the core-depletion calculations are performed by SMART code in each burnup steps from the BOC to EOC under the condition of HFP, EQ.Xe and ARO, and the 3D core neutron-physics distribution parameters vs. the average depletion are obtained with regards to the specific operational history; (3) The homemade CORIST code and COROST code are used to post-process the SMART results so as to obtain the 3D neutron-source strength-density distribution required as the MICROMC (a Monte-Carlo code) input data for the inner 37 fuel assemblies and for the peripheral 10 assemblies (precise to fuel rods) of the 1/4 core configuration of CPR1000 reactor, respectively; (4) The nuclear density of the fissionable nuclides in fuel assemblies and the nuclear density of the other non-fissionable materials are post-processed by the homemade PROCDEN1 code and PROCDEN2 code, respectively; (5) The parallel calculation of MICROMC code is performed for the calculation model of an 1/4 “core-RPV-reactor pit” configuration and aborts while the specific criteria are satisfied; (6) The homemade POSTRES code is used to post-process the MICROMC results, to count the operational time and radiation time, and to calculate the fast-neutron fluence in the steel structure of the RPV and in the concrete structure of the reactor pit, respectively. The verification calculations are also illustrated.