Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

A Rais,M Hursin,A Pautz,G Girardin,D Siefman

doi:10.1155/2015/237646

Abstract

In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

Highlights

A large variety of research reactors have been designed and operated during the last 50 years
Due to the diversity of research reactor designs and operating conditions, there is a wide variety of computational tools used in their safety analysis and, nowadays, it is desired to adopt a standard approach for safety analysis of these research reactors [1]
A project financed by swissnuclear was started with the objective of developing methods and models for the coupled neutronics and thermal-hydraulics analysis of the CROCUS reactor at EFPL using advanced and state-of-theart nuclear power plant (NPP) computational tools

Summary

Introduction

A large variety of research reactors have been designed and operated during the last 50 years. Due to the diversity of research reactor designs and operating conditions, there is a wide variety of computational tools used in their safety analysis and, nowadays, it is desired to adopt a standard approach for safety analysis of these research reactors [1]. The coupling of thermal-hydraulic and neutronics codes becomes a fundamental tool for an accurate reactor behavior prediction under transient and accident conditions. Along those lines, a project financed by swissnuclear was started with the objective of developing methods and models for the coupled neutronics and thermal-hydraulics analysis of the CROCUS reactor at EFPL using advanced and state-of-theart NPP computational tools. The fourth section reviews the current thermal-hydraulic modeling of the CROCUS reactor and describes the proposed model

The CROCUS Reactor

Neutronics Modeling

Benchmark Results

Thermal-Hydraulic Modeling

Conclusions