• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2003년도 춘계학술발표대회 요약집
• /
• pp.96-96
• /
• 2003

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1998년도 추계학술발표회요약집
• /
• pp.335-335
• /
• 1998

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 춘계학술발표회
• /
• pp.61-62
• /
• 2005

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 추계학술발표회
• /
• pp.87-88
• /
• 2005

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2001년도 춘계학술발표회요약집
• /
• pp.248.2-248
• /
• 2001

• Nuclear Engineering and Technology
• /
• 제49권8호
• /
• pp.1575-1588
• /
• 2017

In order to develop a domestic research roadmap for severe accidents, a special committee was established by the Korean Nuclear Society. One of the subcommittees discussed the characteristics and the relevant technical issues in the stages of fission product release and physical forms of radionuclide release and transport. The group members developed a tree to identify fission product release phenomena by tracing failures of individual defense-in-depth barriers and added possible countermeasures against failure. For each elemental issue, they searched for technical problems by examining the phenomena, accident management actions, and regulatory aspects relevant to the mitigation features for containment, including mitigation strategies against containment bypass accidents. Regulatory concerns, including the source term and the acceptance criteria for radionuclide release, were also considered. They identified further research needs regarding important technical issues based on the degree of the current knowledge level in Korea and in foreign countries, looking at the significance and urgency of issues and the expected research period required to reach an advanced level of knowledge. As a result, the group identified the 12 most important and urgent issues, most of which were expected to require mid-term and long-term research periods.

• Journal of applied mathematics & informatics
• /
• 제8권2호
• /
• pp.505-518
• /
• 2001

The huge and complicated plants such as nuclear power stations are likely to cause the operators to make mistakes due to a variety of inexplicable reasons and symptoms in case of emergency. That’s why the prevention system assisting the operators is being developed for. First of all. I suggest an improved fuzzy diagnosis. Secondly, I want to demonstrate that a classification system of nuclear plant’s accident investigating the causes of accidents foresees possible problems, and maintains the reliability of the diagnostic reports in spite of improper working in part. In the event of emergency in a nuclear plant, a lot of operational steps enable the operators to find out what caused the problems based on an emergent operating plan. Our system is able to classify their types within twenty to thirty seconds. As so, we expect the system to put down the accidents right after the rapid detection of the damage control-method concerned.

• Nuclear Engineering and Technology
• /
• 제52권4호
• /
• pp.742-754
• /
• 2020

Zircaloy cladding oxidation is an important phenomenon for both design basis accident and severe accidents, because it results in cladding embrittlement and rapid fuel temperature escalation. For this reason during the last decade, many experts have been conducting experiments to identify the oxidation phenomena that occur under design basis accidents and to develop mathematical analysis models. However, since the study of design extension conditions (DEC) is relatively insufficient, it is essential to develop and validate a physical and mathematical model simulating the oxidation of the cladding material at high temperatures. In this study, the QUENCH-05 and -06 experiments were utilized to develop the best-fitted oxidation model and to validate the SPACE code modified with it under the design extension condition. It is found out that the cladding temperature and oxidation thickness predicted by the Cathcart-Pawel oxidation model at low temperature (T < 1853 K) and Urbanic-Heidrick at high temperature (T > 1853 K) were in excellent agreement with the data of the QUENCH experiments. For 'LOCA without SI' (Safety Injection) accidents, which should be considered in design extension conditions, it has been performed the evaluation of the operator action time to prevent core melting for the APR1400 plant using the modified SPACE. For the 'LBLOCA without SI' and 'SBLOCA without SI' accidents, it has been performed that sensitivity analysis for the operator action time in terms of the number of SIT (Safety Injection Tank), the recovery number of the SIP (Safety Injection Pump), and the break sizes for the SBLOCA. Also, with the extended acceptance criteria, it has been evaluated the available operator action time margin and the power margin. It is confirmed that the power can be enabled to uprate about 12% through best-estimate calculations.

• 에너지공학
• /
• 제26권2호
• /
• pp.50-63
• /
• 2017

복잡하고 다양한 계통으로 구성된 원자력발전소에서의 안전문화 저하 징후는 설비의 안전성능 저하를 유발할 수 있으며, 이에 대한 적절한 후속조치가 없을 경우 대형 사고를 발생시킬 수 있는 잠재적 원인이 된다. 따라서 원전의 안전성능에 대한 감시와 더불어 조직 및 관리측면에서의 안전문화를 모니터링 함으로써 안전문화 저하 징후를 파악할 필요가 있다. 따라서 본 논문에서는 원전 문제점 혹은 사건이 발생할 경우, 설비 측면보다는 조직의 안전문화관점에서 사건의 근본원인에 기여한 안전문화 취약요소를 평가하여 개선할 수 있는 사건기반 평가 방법론을 정립하였다. 이를 위해 국내 외 산업계에서 활용되고 있는 평가사례들을 검토하여 국내 원전 실정에 적합한 평가 방법론을 제안하였다.

• Nuclear Engineering and Technology
• /
• 제32권1호
• /
• pp.57-66
• /
• 2000

The RELAP5 code has been developed for best-estimate simulation of transients and accidents for pressurized water reactors and their associated systems, but it has not been fully assessed for those of CANDU reactors. However, a previous study suggested that the RELAP5 code could be applicable to simulate the transients and accidents for CANDU reactors. Nevertheless, it is indicated that there are some works to be resolved, such as modeling of headers and multi-channel simulation for the reactor core, etc. Therefore, this study has been initiated with an aim to identify the code applicability for all the postulated transients and accidents in CANDU reactors. In the present study, the small inlet header break experiment (B8604) in the RD-14 test facility was simulated with RELAP5/MOD3.2 code. The RELAP5 results were also compared with both experimental data and those of CATHENA analyses performed by AECL and the analyses demonstrated the code's capability to predict major . phenomena occurring in the transient with sufficient accuracy for both Qualitative and quantitative viewpoint However, some discrepancies in the depressurization of the primary heat transport system after the break and the consequent time delay of the major phenomena were also observed.