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서미로(M. R. Seo),조하얀(H. Y. Jo),황미정(M. J. Hwang) 한국동력기계공학회 2008 한국동력기계공학회 학술대회 논문집 Vol.2008 No.11
In Korea, MR(Maintenance Rule) programs development for all PWR were completed. However, in case of PHWR (CANDU type, Wolsong Unit 1&2,3,4), the study of MR program was delayed, since the design concepts and operating experiences are different from those of PWR. Additionally, in Canada that designed the CANDU reactor originally, the MR program has not been tried even though the similar programs such as the Reliability Program and Maintenance Program exist. In 2007, KEPRI started a project for developing similar MR program for the domestic PHWR considering the unique design characteristics and operating experience of Wolsong Unit 1,2,3 & 4.This paper describes the characteristics of CANDU design experienced during the function analysis in scoping process that is the first step in MR program development, and the results of scoping process, ^ince the design features of CANDU is different from those of PWR (for example, using D20 as a coolant and moderator, the pressure tube and Calandria, and the fuel changing machine, etc.), the applicability of NUMARC 93-01 Criteria to CANDU had to be investigated and justified in this paper. Next step, the risk importance determination process for functions in scope will be described. The risk importance was determined by mapping important PSA basic event to function in scope and Delphi method. For Delphi evaluation, Delphi evaluation item for CANDU need to be developed because the design and normal operation functions are different from PWR. The final review for risk importance determination was performed by Expert Panel organized by the plant expert engineers of Wolsong Unit 3&4. In this paper, the discussion results are also described, so it will be a great help to understand the functional characteristics ana identity the important function of CANDU plant
김한철,하광순,김성중,서미로,강상호,이두용,송용만,이종성,임희정,조창석,연제원,김성일,조송원,송진호,류용호 한국원자력학회 2017 Nuclear Engineering and Technology Vol.49 No.8
In order to develop a domestic research roadmap for severe accidents, a special committee was establishedby the Korean Nuclear Society. One of the subcommittees discussed the characteristics and therelevant technical issues in the stages of fission product release and physical forms of radionucliderelease and transport. The group members developed a tree to identify fission product release phenomenaby tracing failures of individual defense-in-depth barriers and added possible countermeasuresagainst 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 forcontainment, 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 thecurrent knowledge level in Korea and in foreign countries, looking at the significance and urgency ofissues 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 requiremid-term and long-term research periods
웨스팅하우스형 원전에서 가압기 안전밸브 개방고착 확률 리스크 영향 평가
오해철 (H. C. Oh),하정헌 (J. H. Ha),서미로(M.R. SEO),정백순 (B. S. Chung),김명기 (M. K. Kim) 한국동력기계공학회 2008 한국동력기계공학회 학술대회 논문집 Vol.2008 No.11
In an effort to increase the quality of probabilistic safety assessments (PSAs), it is recommended that the component failure are included the impact of the accident progression phenomena, either in the accident sequence models or in the system models. During various accident sequences modeled in the PSA, the pressurizer safety valves (PSVs) of a pressurized water reactor (PSVs) may cycle numerous times. The PSVs will initially relieve steam. In some longer duration accident sequences, the steam relief will eventually become liquid relief. With each cycle of the PSV, there is a probability that the safety valve will fail to reseat. The PSV failure-to-reseat results in the need for additional mitigative systems to prevent core damage. This paper describes a quantitative assessment to identify risk impacts of the failure probability due to the repeated cycling operation of PSVs (pressurizer safety valves) in westinghouse type PSA.