사용후핵연료 임계도 평가를 위한 전산코드의 유효성 검증에 관한 연구

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nuclear criticality, code
A Study on the Validation of Nuclear Criticality Analysis Codes for Spent Fuel

Gwangbong Kim
Advisor : Prof. woonkwan Chung, Ph.D.
Department of Nuclear Engineering,
Graduate School of Chosun University

Recently, Regulatory requirements for wet and dry storage of LWR fuel have been strengthened. (DSS-ISG-2010-01 Rev.0 and ISG-8 Rev.3)
Thus, the critical experiments and validation of burnable nuclide in Criticality Evaluation has increased in importance.
This study shows the result of validation test for KENO-Va, MCNP-X code that is often used for criticality evaluation and for SERPENT code which was developed in Finland in 2004 and widely used in Europe recently.
To determine the suitability of the test results for enhanced regulatory requirements, the statistical analysis and validation of three codes were carried out in accordance with NUREG / CR-6698 respectively.
Critical safety analysis target to be considered in this calculation is the validation for the light-water reactor spent fuel storage facilities.
Accordingly, criticality experiments using UO₂ fuel rods constituting the assembly form and using water as a moderator were selected to demonstrate the validity of the analysis method.
Total 7 group(01, 02, 09, 13, 16, 17, 42) 115 experiment data that satisfies these conditions was selected from the Handbook Volume -IV "low enriched uranium systems (LEU-COMP- THERM-0XX)" which has been published by ICSBEP as OECD / NEA documents.
Data analysis was carried out for statistical analysis and validation according to NUREG / CR-6698.
In addition, the Student t-test was used to determine statistical significance of the input data.
If the relationship can be determined between the calculated and the independent variables, the one-sided tolerance band can be used.
Bias and bias uncertainty of one-sided tolerance band for the criticality experiment is expressed as follows.
bias = K(fit) - 1.0 : K(fit) 가 1.0 미만인 경우
bias = 0.0 : K(fit) 가 1.0 이상인 경우

bias uncertainty = 본문 식 2-26 참조
Summary of Calculation result (SERPENT) : 본문 Table 3-15 참조
Summary of Bias and Bias uncertainty as a function of enrichment :
본문 Table 4-1 참조
Summary of Bias and Bias uncertainty as a function of fuel diameter :
본문 Table 4-2 참조

Applying the one-sided tolerance band for the fuel enrichment and diameter to criticality evaluation we can calculate the optimal which satisfies the USL in criticality evaluation.
Based on bias and bias uncertainty evaluation results for the three codes , we can find the best estimated bias of the SERPENT code is lowest.
Alternative Title
A Study on the Validation of Nuclear Criticality Analysis Codes for Spent Fuel
Alternative Author(s)
Kim, Gwang Bong
조선대학교 대학원
일반대학원 원자력공학과
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Table Of Contents
목 차


제 1 장 서론···················································1

제 2 장 전산코드 입력자료 생성 및 분석·················3
제 1 절 전산코드······················································3
제 2 절 검증방법(Validation Method)······························5
1. 검증인자의 적용범위 정의 5
2. 임계실험 선정 6
다. LEU-COMP-THERM-009 11
라. LEU-COMP-THERM-013 13
마. LEU-COMP-THERM-016 14
바. LEU-COMP-THERM-017 16
사. LEU-COMP-THERM-042 17
3. 임계실험 모델링 18
제 3 절 데이터 분석 25
1. 정규성 검정 25
2. 편이 및 편이 불확실도 결정 25
3. 데이터 경향성 분석 27
4. 데이터 처리를 위한 통계분석 방법 29
가. 일방공차한계 (one-sided tolerance limit) 29
나. 공차구간 (tolerance band) 30

제 3 장 전산코드 검증결과 33
제1절 정규성 검정(Normality Test) 33
제2절 편이 및 편이 불확실도 결정 35
제3절 경향분석 42
1. SCALE 계산결과 경향 분석 42
가. 농축도 43
나. 연료봉 지름 43
다. 격자 피치 44
라. SCALE 경향분석 요약 45
2. MCNP 계산결과 경향 분석 49
3. SERPENT 계산결과 경향 분석 53

제 4 장 결론················································59

조선대학교 대학원
김광봉. (2016). 사용후핵연료 임계도 평가를 위한 전산코드의 유효성 검증에 관한 연구.
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