SuperMC应用于先进钠冷快堆的适用性评估

摘要

FDS Team(INEST,CAS,China)developed a highly efficient and general purpose code-the SuperMC-which is capable to model and simulate the complex nuclear systems.In order to validate and apply the SuperMC code/program in the area of designing and safety analysis of fast reactors,the high-fidelity core neutronics simulations including transport and bumup calculations were performed and analyzed based on two benchmarked Sodium-cooled Fast-spectrum Reactors(BFS-62-3A and BN-600).The study could essentially be divided in to two major parts:
　　Ⅰ.The benchmarking of SuperMC program for Sodium-cooled Fast Reactors'neutronic analysis with BFS-62-3A which is a benchmarked experiment conducted at BFS-2facility,IPPE,Russia.This work incorporates three distinctive subjects including:
　　a.Validation of SuperMC code for its neutron transport calculation capability.During the validation,various parameters were calculated-The spectral indices,for instance,gave C/E values0.9927and0.9990for F49/F25and F28/F25.The average discrepancies for radial fission rate distributions in the fuel region and control rod worths were found to be2.7％and less than5％respectively.
　　b.The testification of competence of indigenously developed point-wise nuclear data library,the HENDL/MC.The simulations agreed well with the experiment and this fraction of our study hence enabled the HENDL/MC library to be benchmarked.
　　c.A detailed investigation on core's neutronic analysis of BFS-62-3A:including estimation of the impact of reflector density on fission rates,sodium void reactivity effect,sensitivity analysis,the effects of various data libraries on the criticality and fission rates,and code-to-code verification.For reflector's density dependence of reaction rates in peripheral region of the assembly,for instance,a decrease of density by5％was found to be in good agreement with the experiment.For code-to-code verification,control rod worth(CRW)and fission rates were calculated.For CRW,the average deviations are8.19％and4.97％for the Serpent and SuperMC respectively.For fission rates,the average discrepancies between SuperMC and other codes were about1.8％and1.6％for239pu and235U,respectively.
　　Ⅱ.The neutronic analysis of BN-600hybrid-core reactor using SuperMC program.The prototype reactor was reconstructed for criticality studies and fuel burnup analysis.The calculations,keeping the shim rods(SHRs)mid-core inserted and scram rods(SCRs)fully out/withdrawn,involved estimation of change in isotopic concentrations and prediction of burnup effect and burnup reactivity loss during reactor operation.The results so obtained were compared with the reference data available in IAEA's technical document(IAEA-TECDOC-1623)and other published articles.The comparison gave a good agreement of the SuperMC's results with available reference data.
　　It is to conclude that SuperMC code is quite competent and capable to perform the neutronic analysis of fast spectrum sodium-cooled reactors.During the neutronics study of BFS reactor(described in

目录

声明

ABSTRACT

Table of Contents

List of Figures

List of Tables

Nomenclature

Preface

1 Introduction

1．1 Nuclear energy

1．1．1 Challenges with nuclear energy sector

1．1．2 Opportunities and benefits with nuclear energy sector

1．1．4 Fast reactors’ technology

1．3 Some featured work on sodium-cooled fast reactors in the world

1．3．1 China-The CFR-600

1．3．2 Russia-The BN-1200

1．3．3 USA-The PRISM

1．3．4 France-The ASTRID

1．4 Research status of SFRs’ neutronics

1．5 Research Significance

1．6 Objectives of thesis

1．7 Structure of thesis

2 Basics of reactor physics，methodologies and tools for core neutronics simulation

2．1 Basics of reactor physics

2．1．1 Neutronic parameters

2．1．2 Physics behind breeding and transmutation

2．2 Methodologies for coge neutronics simulation

2．2．1 Methods for neutron transport simulation

2．2．2 Methods to solve burnup equations

2．3 Tools for core neutronics simulation

2．3．1 SuperMC code-An overview

2．3．2 Nuclear data libraries

2．4 Summary

3 Neutronic analysis of a BFS test reactor

3．1 Innovative nuclear-related development and nuclear research facilities in Russia

3．1．1 BFS-2

3．1．2 Some featured accomplishments of BFS-2

3．1．4 Core configuration of BFS-62-3A

3．1．5 Objective of the BFS-62-3A experiment

3．1．6 Short description on BFS-62 experiments

3．2 Validation of SuperMC code

3．2．1 Modeling and simulations

3．2．2 Results and discussion

3．3 Benchmarking of HENDL library

3．3．1 The strategy adopted for this work

3．3．2 Monte Cado simulations for benchmarking

3．3．3 Results and discussion

3．4 Detailed neutronic analysis of BFS-62-3A test reactor

3．4．1 Strategy adopted for this work

3．4．2 Main purpose of this work

3．4．3 Simuladons performed

3．4．4 Results and discussion

3．5 Summary

4 Criticality and burnup studies of BN-600 reactor

4．1 BN-600 reactor-An overview

4．2 Simulations performed

4．3 Results and discussion

4．3．1 Effective multiplication factor

4．3．2 Burnup studies

4．4 Summary

5 Conclusions and future directions

5．1 Summary conclusions

5．2 Future directions

Bibliography

Appendix

Acknowledgement

Publications and research achievements