보고서 정보
주관연구기관 |
한국원자력안전기술원 Korea Institute of Nuclear Safety |
연구책임자 |
정명조
|
참여연구자 |
강성식
,
고한옥
,
김경조
,
김상현
,
김석훈
,
김선재
,
김선혜
,
김용범
,
김진성
,
김진수
,
김홍기
,
박정순
,
박찬일
,
박희태
,
노윤석
,
송태광
,
송재호
,
신강식
,
유신정
,
이강산
,
이상민
,
이진호
,
정연기
,
정진석
,
최성부
,
최진복
,
홍진기
,
장윤석
,
김윤재
,
허남수
,
송성진
,
김종성
,
박재학
|
보고서유형 | 1단계보고서 |
발행국가 | 대한민국 |
언어 |
한국어
|
발행년월 | 2013-05 |
주관부처 |
원자력안전위원회 Nuclear Safety and Security Commission |
등록번호 |
TRKO202100004604 |
DB 구축일자 |
2021-07-10
|
키워드 |
장기가동원전.경년열화.접근제한용접부.일차수응력부식균열.비파괴검사.원자로내부구조물.주기적 안전성평가.경년열화 통합관리시스템.내진 여유도 평가.배관파손확률.환경피로.Long-term Operating NPPs.Aging.Non-accessible Welds.PWSCC.Reactor Vessel Internals.Periodic Safety Review.Integrated Regulatory Aging(IR-Aging) Management System.Seismic Margin Analysis.Pipe Failure Probability.Environmental Fatigue.
|
초록
▼
• 접근제한설비 경년열화평가 및 핵심기기 열화관리 규제기술 개발
- 접근제한설비 가동중검사의 열화평가 및 문제점 분석
- 접근제한설비 경년열화 완화대책의 규제 적용성 분석 및 평가
- 증기발생기 열화관리에 따른 보수·교체 사례 및 규제요건 분석
- 예방적 재료열화평가에 의한 규제적용 우선순위 분석
• 원자로내부구조물 경년열화관리 및 평가 규제기술 개발
- 원자로내부구조물 규제기준, 설계특성, 경년열화 및 운전경험 분석
- 원자로내부구조물 균열발생 평가 및 지진해석 기술 개발
• 접근제한설비 경년열화평가 및 핵심기기 열화관리 규제기술 개발
- 접근제한설비 가동중검사의 열화평가 및 문제점 분석
- 접근제한설비 경년열화 완화대책의 규제 적용성 분석 및 평가
- 증기발생기 열화관리에 따른 보수·교체 사례 및 규제요건 분석
- 예방적 재료열화평가에 의한 규제적용 우선순위 분석
• 원자로내부구조물 경년열화관리 및 평가 규제기술 개발
- 원자로내부구조물 규제기준, 설계특성, 경년열화 및 운전경험 분석
- 원자로내부구조물 균열발생 평가 및 지진해석 기술 개발
- 원자로내부구조물 유동 평가 기술 개발
- 경년열화 통합관리시스템 개발
• 원전 주요기기 동적영향 및 파손확률 평가 규제기술 개발
- 원전 주요기기 및 배관 성능기반 내진설계 기술 개발
- 유동 및 환경영향을 고려한 배관 건전성평가 기술동향 분석
- 파손확률평가를 위한 건설/가동원전 배관손상사례 분석 및 데이터베이스 구축
(출처 : 보고서 요약서 5p)
Abstract
▼
IV. Research Results
[Development of Regulatory Assessment and Management Technology for Aging of Inaccessible Component]
• Evaluation technique development of performance demonstration for ICW in-service inspection
- NDE accessibility of Reactor vessel welds and head penetration welds we
IV. Research Results
[Development of Regulatory Assessment and Management Technology for Aging of Inaccessible Component]
• Evaluation technique development of performance demonstration for ICW in-service inspection
- NDE accessibility of Reactor vessel welds and head penetration welds were reviewed for the review of ICW as a first stage.
- Current NDE technique and advanced NDE technique was reviewed in order to check the inspectability of ICW.
- Draft version of the regulatory inspection guide for IWC was developed.
- Ultrasonic beam tracing technique program in inaccessible component weld(ICW) and time delay matrix calculation technique are developed. On the other hand, the international RRT(Round Robbin Test) was finished for PARENT(Program for Assess Reliability of Emerging NDE Techniques) and the RRT results were discussed in PARENT 6 meeting which was held by KINS in Korea Seoul.
• Evaluation of preventive mitigation methods on degradation of inaccessible components and weldments
- Technical characteristics of typical preventive mitigation methods on degradation of inaccessible components and weldments, i.e. weld overlay, weld inlay/onlay, mechanical stress improvement process, laser peening and water jet peening, were analyzed and worldwide adoption cases were investigated.
- Weld Overlay method have been used world-widely since 1980’s, and relevant industry technical standard, i.e. ASME Cods Case N-740, was developed. Alternative request shall be submitted to regulatory body, however, to adopt weld overlay method since Code Case N-740-1 is not approved by NRC yet.
- ASME Code Case N-766, industry technical standard for Weld Inlay/Onlay method, was developed in 2010. NRC has been negative in approving Code Case N-766 since NRC understand that crack growth evaluation on Alloy 52 material was not fully proved.
- Mechanical Stress Improvement Process(MSIP) has been used in BWR plants successively since it was developed in 1982. The applicability in PWR plants was verified by NRC and MSIP has been proved effective method to mitigate degradation of inaccessible components and weldments.
- Water Jet Peening(WJP) and Laser Peening(LP) has been used in Japan nuclear power plant successively since 1999 and US has no experience for WJP and LP in nuclear power plant. US industry, however, submitted topical report for WJP and LP in 2012 and the relevant document, MRP-335, is under review by NRC.
- Existing integrity evaluation method, i.e. ASME, R6, and API579, and recently suggested method, i.e. Univeral Weight Function Method, were verified and new evaluation method, which is not sensitive to the result of stress-fitting but accurate, was developed
• Analysis of regulatory requirements and cases of repair/replacement activity for nuclear major components
- Regulatory requirements(NSSC act of in-service inspection, KEPIC MI, NRC inspection manual and so on) of repair/replacement activity for nuclear major components were reviewed
- Industry experience cases of repair/replacement activity for nuclear major components were reviewed
- An interim technical inspection guidance for the repair/replacement activity of steam generator was developed
• Development of assessment technique for PMMD(Proactive Management for Material Degradation) of aging sensitive materials
- For the proactive material aging management, the evaluation group which is composed by Korean experts was utilized to decide the rank of material degradation importance. This result will be utilized for the setting up the plan of safety regulatory research.
- MIRRE System was developed by aging management program database. Specially, the regulatory inspection guideline of buried pipe and tank which is high rank of material degradation was developed.
[Development of Regulatory Assessment and Management Technologies for Aging of Reactor Vessel Internals]
• Analysis of categories, aging mechanisms, and operating experience of RVIs
- Regulatory standards and industrial codes & standards applicable to RVIs are investigated and analyzed. Overseas industrial codes and standards for RVIs include ASME B&PV Code Section III NG, France RCC-M and Germany KTA 3204 for design, and ASME B&PV Code Section XI, Japan JEAC-4205 and Germany KTA 3204 for inservice inspection. Domestic regulatory requirements regarding RVIs can be categorized as the ones applicable to construction and operational stage of NPPs. For the construction stage, regulatory requirements for design, construction and preservice inspection are provided. For the operational stage, regulatory requirements concerning inservice inspection, PSR, continued operation are provided.
- Typical design characteristics of the Westinghouse(WH)-type and Combustion Engineering(CE)-type RVIs are investigated and provided. In addition, design characteristics of the domestic RVIs, which include WH-, CE - and Framatome-type, are also analyzed. In this study, design characteristics of RVIs include detailed components of RVIs and their functions, materials, design loads, and design criteria.
- Operating conditions such as neutron dose, water chemistry, and operating temperature for RVIs are analyzed. Considering those, 8 possible aging mechanisms(stress corrosion cracking, wear, fatigue, thermal aging embrittlement, irradiation assisted stress corrosion cracking, irradiation embrittlement, void swelling, stress relaxation/creep) and their thresholds or screening criteria are investigated. For aging mechanisms related to the irradiation, more data should be obtained under the pressurized water reactor condition. In addition, degradation and repair/replacement experiences of RVIs are provided.
• Development of the safety review guide for the PSR(Periodic Safety Review) of RVIs and IR-Aging(Integrated Regulatory Aging management system)
- The draft PSR safety review guide for RVIs is developed based on the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs. The draft PSR safety review guide is divided into six major chapters and one appendix. Six chapters are (1) Areas of Review, (2) Acceptance Criteria, (3) Review Procedures, (4) Evaluation Findings, (5) Implementations, (6) References.
- IR-Aging system is developed for the systematic regulatory management of aging in main components of NPPs for long-term operation. The web-based IR-Aging is composed of four sub-modules: Regulatory Requirement, Aging Database, AMP/TLAA, Audit calculation program modules. In addition, integrated database for RVIs is also developed within the IR-Aging system and it includes all documents necessary to aging assessment and management of RVIs.
• Development of the crack initiation prediction technique for the RVIs
- The crack initiation prediction technique of the reactor internals considering fatigue and IASCC was developed, and then applied to BFA, LCP, and UCP alignment pin.
- As a result of the application, the calculated CUF, IASCC initiation damage, and combined damage don't exceed 1 during design lifetime. Therefore, it is found that structural integrity of the BFA, LCP, and UCP alignment pin will be maintained during the design lifetime in the viewpoint of fatigue and IASCC.
• Development of seismic analysis techniques for RVIs
- Seismic or dynamic analysis for statically determinate beams subjected to in-phase, multi-support motions is performed theoretically and numerically.
- The theoretical analysis shows the following facts: The large mass model of the statically determinate beams is approximately equivalent to the effective force model when an appropriate large mass ratio is employed for large mass model. This fact holds true for both of differential equation of motion and the finite element equation of motion.
- The finite element analysis based on the large mass model shows the following facts: The accuracy of solutions of LM model is very accurate when the large mass ratio is chosen within an appropriate range. The range of the large mass ratio is 103~1011. However, 107~108 is recommended because the range would be slightly varied depending on commercial programs and machines.
• Development of thermo-hydraulic analysis technique for RVIs
- Design documents for one of domestic RVIs are analyzed. As a result, analysis models are developed for six main structures of RVIs(core support barrel, upper guide structure, core shroud assembly, lower support structure, guide structure support system, control element assembly shroud) to perform thermo-hydraulic and structural analysis. In addition, virtual reality models for RVIs are also developed to understand the geometrical characteristics with the aid of the visualization technique.
- The detailed analysis models are optimized for the cost-effective simulation and used in the thermo-hydraulic and structural analyses.
- Thermo-hydraulic and structural analyses for RVIs are performed considering periodic pump pulsation pressure and irregular loadings due to turbulent, respectively. As a result, the maximum stress value is found at the location of cold leg where the constraint condition was applied. Analysis results considering turbulent condition showed higher stress values than the ones considering periodic pump pulsation, while maximum stress values are found at the same location for both of the cases.
[Development of Regulatory Technology for the Evaluation of Dynamic Effects and Failure Probability of Nuclear Power Plant Components]
• Seismic design technologies analysis related nuclear power plant components and development of seismic evaluation technical guidelines (draft)
- Domestic/foreign regulatory guidelines and industrial standards for seismic design were investigated and seismic design technologies of nuclear power plant such as seismic design evaluation codes, seismic analysis guidelines, detailed procedures of seismic qualification, seismic analysis methods were analysed. Based on the results of the technical analysis, seismic qualification procedures and evaluation guidelines(draft) for seismic safety evaluation of nuclear power plant components were developed.
• Technologies analysis related pipe integrity assessment and environmental fatigue analysis
- The literature survey for recent flow analysis technologies together with the OECD/NEA IBE-1(the first OECD/NEA International Benchmark Exercise) were conducted, and evaluation procedures of thermal fatigue due to flow mixing and thermal stratification and environmental fatigue assessment methods using the environmental correction factor were analyzed. The proposed method was applied to resolve sample problems presented in the EPRI guidelines and calculated environmental correction factor were compared with existing results. Based on the literature survey and validation results, technical guidelines(draft) for nuclear piping integrity assessment considering environmental effects were developed.
- In order to analyze techniques for flow analysis and thermal fatigue evaluation methods, thermal fatigue damage cases due to flow mixing in domestic and foreign nuclear power plants were investigated. The generation mechanism of thermal fatigue, the general theory of numerical evaluation, assessment procedures, limitations were analyzed. In addition, the thermal fatigue evaluation of shutdown cooling heat exchanger downstream junction applying numerical evaluation method were carried out. The optimal evaluation method was derived by comparing and analyzing the evaluation results.
• Nuclear piping damage case analysis of construction/operating plants and database construction for domestic/foreign nuclear piping damage
- For identifying technology trend of nuclear piping failure probability assessment, example analysis was performed using the PRO-LOCA program and effects of recently applied equations and input variables were analyzed. Improvements of pre-developed P-PIE program were derived, the improved P-PIE program was developed to predict accurately the failure probability of nuclear piping.
- For analyzing nuclear piping damage cases of construction and operating plants, nuclear piping damage cases of safety and non-safety related pipes of domestic and foreign plants in OPDE database were investigated. By analysing the database structure and items of OPDE and CODAP database, DB construction draft was proposed. In addition, by investigating repair welding part of Korean construction nuclear power plants(SWN Units 1&2 and SKN Units 3&4), repair welding database that was composed of the welding defect types, welding defect position, repair types, and non-destructive testing results was developed.
(출처 : SUMMARY 17p)
목차 Contents
- 표지 ... 1
- 제 출 문 ... 3
- 보고서 요약서 ... 5
- 요 약 문 ... 7
- SUMMARY ... 15
- Contents ... 25
- 목차 ... 27
- 제 1 장 연구개발과제의 개요 ... 29
- 제 2 장 국내외 기술개발 현황 ... 35
- 제 1 절 국내 기술개발 현황 ... 35
- 제 2 절 국외 기술개발 현황 ... 37
- 제 3 절 현 기술상태의 취약성 ... 39
- 제 3 장 연구개발수행 내용 및 결과 ... 43
- 제 4 장 목표달성도 및 관련분야에의 기여도 ... 45
- 제 5 장 연구개발결과의 활용계획 ... 47
- 제 6 장 연구개발과정에서 수집한 해외과학기술정보 ... 49
- 제 7 장 참고문헌 ... 51
- 부록 1 접근제한설비 경년열화평가 및 핵심기기 열화관리 규제기술개발 ... 69
- 부록 2 원자로내부구조물 경년열화관리 및 평가 규제기술개발 ... 247
- 부록 3 원전 주요기기 동적영향 및 파손확률 평가 규제기술 개발 ... 631
- 끝페이지 ... 960
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