• 대한기계학회논문집A
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• 제39권3호
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• pp.291-301
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• 2015

제품 규정에는 제조사가 준수해야 하는 제품에 관한 다양한 지침 및 규제사항이 담겨 있다. 이 연구에서는 KS 표준을 활용하여 압력용기의 설계 결과를 검증하는 시스템 프레임워크와 구성 요소들을 제안한다. 그리고 기간 시스템으로부터 설계 템플릿 데이터를 생성하는 방법과 규정 지식베이스를 구축하는 방법을 제시한다. 마지막으로 압력용기 설계 검증 시스템을 구현하고 테스트 데이터를 활용한 실험을 통해 시스템 프레임워크를 검증한 결과를 논의한다.

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.20-24
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• 2013

Reactor Vessel Internals(RVIs), which are installed within the reactor pressure vessel and support the fuel assembly, take responsibility for safety of reactor core. In operating Nuclear Power Plants(NPPs), the RVIs have been exposed to severe conditions such as neutron irradiation, high temperature, high pressure, and high velocity of coolant flow and have degraded by materials aging with long-term operation. Therefore, the effective aging management plan and the appropriate regulatory requirements are necessary to maintain the integrity of RVIs. The purpose of this paper is to provide a review guide for Periodic Safety Review(PSR) of RVIs in presurized water reactor. The review guide is developed based on the revised review guides and reports established from IAEA and USNRC, and the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs in domestic and international NPPs. Consequently, the developed review guide for PSR of RVIs is expected to contribute an overall strategy and standard for the PSR of RVIs.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1726-1746
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• 2022

This paper presents fracture mechanics analysis results for various cracks located at pressurized water reactor pressure vessel nozzle crotch corners taking into consideration constraint effect. Technical documents such as the ASME B&PV Code, Sec.XI were reviewed and then a fracture mechanics analysis procedure was proposed for structural integrity assessment of various nozzle crotch corner cracks under normal operation conditions considering the constraint effect. Linear elastic fracture mechanics analysis was performed by conducting finite element analysis with the proposed analysis procedure. Based on the evaluation results, elastic-plastic fracture mechanics analysis taking into account the constraint effect was performed only for the axial surface crack of the reactor pressure vessel outlet nozzle with cladding. The fracture mechanics analysis result shows that only the axial surface crack in the reactor pressure vessel outlet nozzle has the stress intensity factor exceeding the low bound of upper-shelf fracture toughness irrespectively of considering the constraint effect. It is confirmed that the J-integral for the axial crack of the outlet nozzle does not exceed the ductile crack initiation toughness. Hence, it can be ensured that the structural integrity of all the cracks is maintained during the normal operation.

• Composites Research
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• 제21권6호
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• pp.23-30
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• 2008

본 논문에서는 자연 노화가 필라멘트 와인딩으로 제작된 압력용기의 강도 분포와 구조 사용 수명에 미치는 영향을 연구하였다. 자연 노화에 따라 변화되는 섬유 방향 파괴 변형률을 설계 확률 변수로 하는 확률 강도 해석을 수행하였다. 이때 확률강도 해석은 정확한 파열 압력을 예측하기 위해 연속 파손 모드가 고려되었고, 비선형 한계식의 해를 구하기 위해 FORM방법이 이용되었다. 해석을 통해 노화 시간별 파괴 확률 분포 선도를 구하였다. 복합재 구조물의 특성상 재료 물성 및 제작 공정 변수 영향으로 제품의 성능 변동성이 비교적 크게 나타났고, 노화로 인한 압력용기의 파열 압력 저하 현상은 대부분 10년 이내에서 발생하였다. 임의 적층의 복합재 압력 용기를 모델로 하여 수명을 평가한 결과, 파괴 확률 2.5%와 안전율 1.3을 고려한 설계 압력 3,250psi기준으로 약 13년의 사용 수명이 평가되었다.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2005년도 춘계학술대회 논문집
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• pp.1803-1807
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• 2005

The pipe which it uses from the nuclear power plant or factory by a long period use and a corrosive action the inside defect occurs on the inside. abstract here. The ESPI method is in order to investigate the laser light in the measurement object it will be able to measure the wide territory whole in once, does not receive an effect in direction of defect not to be, has the strong point it will be able to measure a change of place arrowhead real-time defect. It measured a inside defect of pressure vessel by using Out of plane ESPI and In plane of ESPI. It compared a each method result.

• Structural Engineering and Mechanics
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• 제37권3호
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• pp.257-265
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• 2011

Plant-specific analyses of 5 types of domestic reactors in Korea are performed to assure the structural integrity of the reactor pressure vessel (RPV) during transients which are expected to initiate pressurized thermal shock (PTS) events. The failure probability of the RPV due to PTS is obtained by performing probabilistic fracture mechanics analysis. The through-wall cracking frequency is calculated and compared to the acceptance criterion. Considering the fluence at the end of life expected by surveillance test, the sufficient safety margin is expected for the structural integrity of all reactor pressure vessels except for the oldest one during the pressurized thermal shock events. If the flaw with aspect ratio of 1/12 is considered to eliminate the conservatism, the acceptance criteria is not exceeded for all plants until the fluence level of $8{\times}10^{19}\;n/cm^2$, generating sufficient margin beyond the design life.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2005년도 추계학술대회 논문집
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• pp.910-913
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• 2005

The pipe which it uses from the nuclear power plant or factory by a long period use and a corrosive action the inside defect occurs on the inside. abstract here. The ESPI method is in order to investigate the laser light in the measurement object it will be able to measure the wide territory whole in once, does not receive an effect in direction of defect not to be. has the strong point it will be able to measure a change of place arrowhead real-time defect. It measured a inside defect of pressure vessel by using ESPI and FEM. It compared a each method result.

• 한국공작기계학회:학술대회논문집
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• 한국공작기계학회 2005년도 춘계학술대회 논문집
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• pp.228-233
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• 2005

The damage of the pressure courage by degradation can become the reason of unexpected break down or failure accident and it is very important because safety accident, the production loss, environmental pollution, social problems are occur. Consequently The result to investigat of failure accident for domestic pressure vessel, the factor of degradation is SCC, Sorrosion, Cavity, Crack.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.401-413
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• 2022

This paper provides a simple method by which to estimate the cross-section stress profiles for nozzles designed according to ASME Code Section III. Further, this method validates the effectiveness of earlier work performed by the authors on standard nozzles. The method requires only the geometric information of the pressure vessel and the attached nozzle. A PWR direct vessel injection nozzle, a PWR outlet nozzle, a PWR inlet nozzle and a BWR recirculation outlet nozzle are selected based on their corresponding specific designs, e.g., a varying nozzle radius, a varying nozzle thickness and an outlet nozzle boss. A cross-section stress profile comparison shows that the estimates are in good agreement with the finite element analysis results. Differences in stress intensity factors calculated in accordance with ASME BPVC Section XI Appendix G are discussed. In addition, a change in the dimensions of an alternate nozzle design relative to the standard values is discussed, focusing on the stress concentration factors of the nozzle inside corner.

• 한국압력기기공학회 논문집
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• 제19권1호
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• pp.36-43
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• 2023

Lots of researches have been conducted on in-vessel retention (IVR) to prevent or mitigate severe accident in nuclear power plants. Various methodologies were proposed and the external reactor vessel cooling was selected as a part of promising IVR strategy. In this study, the strategy is strengthened by enhancing the natural circulation performance through the adoption of insulation in the reactor cavity. A thermal analysis was carried out based on an assumed accident scenario and its results were used as boundary conditions for subsequent seven flow analysis cases. By comparing the natural circulation performance, effects of annular gaps and insulation shapes on the mass flow rate and flow velocity were quantified. The improvement in cooling performance can be reflected in actual design via detailed assessment.