• 한국전산구조공학회논문집
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• 제34권5호
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• pp.319-325
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• 2021

후쿠시마 원전사고 발생으로 다수기의 지진안전성에 관한 연구의 필요성이 부각되었다. 한 부지에 건설된 원자력발전소의 경우 유사한 지진응답을 보이기 때문에 적게나마 원자력발전소 SSCs간의 지진손상에 대하여 상관성이 존재하므로 합리적 지진안전성 평가를 위하여 지진손상 상관성을 고려하여야 한다. 본 연구에서는 쌍둥이 호기의 필수전원상실사건에 대하여 확률론적 지진안전성 평가를 수행하였다. 적절한 지진손상 상관계수를 도출하기 위하여 확률론적 지진응답해석을 수행하여 적용하였다. External Event Mensuration System 프로그램을 활용하여 다수기의 필수전원상실사건의 고장수목을 구성하여 지진취약도 및 지진리스크를 분석하였다. 또한 SSCs간의 지진손상 상관성을 완전독립 및 완전종속으로 고려하여 비교 분석을 수행하였다.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1297-1306
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• 2019

Probabilistic safety assessments (PSA) have been used for several decades to visualize the risk level of commercial nuclear power plants (NPPs). Since the role of a human reliability analysis (HRA) is to provide human error probabilities for safety critical tasks to support PSA, PSA quality is strongly affected by HRA quality. Therefore, it is important to understand the underlying limitations or problems of HRA techniques. For this reason, this study conducted a survey among 14 subject matter experts who represent the HRA community of domestic Korean NPPs. As a result, five significant HRA issues were identified: (1) providing a technical basis for the K-HRA (Korean HRA) method, and developing dedicated HRA methods applicable to (2) diverse external events to support Level 1 PSA, (3) digital environments, (4) mobile equipment, and (5) severe accident management guideline tasks to support Level 2 PSA. In addition, an HRA method to support multi-unit PSA was emphasized because it plays an important role in the evaluation of site risk, which is one of the hottest current issues. It is believed that creating such a catalog of prioritized issues will be a good indication of research direction to improve HRA and therefore PSA quality.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1146-1156
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• 2021

Single-unit probabilistic safety assessment (SUPSA) has complex Boolean logic equations for accident sequences. Multi-unit probabilistic safety assessment (MUPSA) model is developed by revising and combining SUPSA models in order to reflect plant state combinations (PSCs). These PSCs represent combinations of core damage and non-core damage states of nuclear power plants (NPPs). Since all these Boolean logic equations have complemented gates (not gates), it is not easy to generate exact Boolean solutions. Delete-term approximation method (DTAM) has been widely applied for generating approximate minimal cut sets (MCSs) from the complex Boolean logic equations with complemented gates. By applying DTAM, approximate conditional core damage probability (CCDP) has been calculated in SUPSA and MUPSA. It was found that CCDP calculated by DTAM was overestimated when complemented gates have non-rare events. Especially, the CCDP overestimation drastically increases if seismic SUPSA or MUPSA has complemented gates with many non-rare events. The objective of this study is to suggest a new quantification method named probability subtraction method (PSM) that replaces DTAM. The PSM calculates accurate CCDP even when SUPSA or MUPSA has complemented gates with many non-rare events. In this paper, the PSM is explained, and the accuracy of the PSM is validated by its applications to a few MUPSAs.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1289-1297
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• 2018

Preliminary surface soil Derived Concentration Guideline Levels (DCGLs) were derived conforming to the Multi-Agency Radiation Site Survey and Investigation Manual (MARSSIM) procedure for the site release and reuse of Kori Unit 1 in Korea. Based on the decommissioning experiences of the U.S. nuclear power plants, a suite of residual radionuclides was determined, and uncertainties contributed to the resultant dose by the input parameters were quantified via the sensitivity analysis of parameters. The peak of the mean dose was obtained via the probabilistic analysis of the RESRAD (RESidual RADioactivity)-ONSITE code. Consequently, $DCGL_w$ of Kori Unit 1 in accordance with two scenarios, industrial worker and residential farmer scenario, were derived and the results were compared respectively with other NPPs. It could be used as a basic guideline for establishing regulatory standards for reuse planning, designing the site characterization surveys and implementing final status survey (FSS).

• Journal of Radiation Protection and Research
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• 제37권3호
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• pp.129-137
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• 2012

국제방사선방호위원회(ICRP)는 2007년 발행된 ICRP 103 권고를 통해, 행위와 개입으로 대변되는 방사선방호 지침을 각 피폭상황 별로 적용하도록 변경하여 권고하였다. 이 지침에는 계획피폭상황에서 방사선방호 최적화의 수단으로 방사선작업종사자와 일반인에 대해 선원중심의 선량제약치(dose constraint)를 설정하여 운영하도록 권고하고 있다. 이 논문에서는 계획피폭상황에서 일반인 선량제약치를 설정하는데 필요한 국내 원전의 방사성물질의 배출량과 이에 따른 주변주민의 피폭방사선량 평가 결과를 분석하였다. 이를 바탕으로 국내 원전의 동일부지 내 복수호기 원전의 운영을 고려한 선량제약치 설정 방안을 제시하였다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.2606-2611
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• 2008

The high energy fluid leakage from the high temperature and high differential pressure drop system of NPPs (Nuclear Power Plants) decreases efficiency and consequently leads to considerable economic loss due to less power production. Also, the leakage possibly damages critical parts of components such as valve and trim with the effect of cavitation, flashing, and erosion, etc. and deteriorates its performance. Thus, in this study, we diagnosed the as-is leakage for four (4) main steam drain valves and two (2) steam traps of Yonggwang 1,2 units during normal operation by using multi-measuring technique and observed the occurrence of fine leakage. In the course of measuring fluid leakage, the sign of fine leakage is estimated to be the leakage from orifice. By converting the leakage to energy loss, it is equivalent to the amount of several hundred thousand won per each unit, which supports the basis for the justification of fine leakage.

• 한국안전학회지
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• 제38권3호
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• pp.110-120
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• 2023

Following the Fukushima accident, portable equipment employed as accident mitigating systems have been installed and operated to reduce core damage and large early release frequencies. In addition, the establishment of an accident management strategy has gained importance. This study investigated the current status of portable equipment including the international portable equipment FLEX (diverse and flexible coping strategies), and domestic portable equipment multi-barrier accident coping strategy (MACST). Research on optimal utilization of MACST remains insufficient. As a preliminary study for establishing an optimal strategy, sensitivity studies were conducted to facilitate the priority of use on portable equipment, number of portable equipment, and dependency of operator actions based on a multi-unit probabilistic safety assessment model. The results revealed the conditions that reduced the multi-unit and site conditional core damage probabilities, indicating the optimal strategy of MACST. The results of this study can be used as a reference for establishing an optimal strategy that utilizes domestic safety equipment in the future.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.3458-3463
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• 2007

Leakage would happen because of the damage of high temperature and high-pressure valve in nuclear power plant. condition based prevention maintenance is essential by using the suitable method based on local condition. Energy loss prevention can prevent from an accurate test, Local actually and ability. The methods of test for high energy fluid leakage at present are analysis of ${\Delta}$T, AE(Acoustic Emission) analysis, and thermal image. The result for test of AC (Main steam) system in YNG 2 Unit reveals that the AE occurred clearly in leakage situation, but thermal image didn't occur. It is identified that leakage is occurred when the orifice located front and back of valve operates. It shows that making a impatient judgment by using the single method if it is leakage is containing uncertainty. So I think that using the Multi-Measuring method is more sound judgment than Single-Measuring method.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.981-988
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• 2023

A seismic event caused an accident at the Fukushima Nuclear Power Plant, which further resulted in simultaneous accidents at several units. Consequently, this incident has aroused great interest in the safety of nuclear power plants worldwide. A reasonable safety evaluation of such an external event should appropriately consider the correlation between SSCs (structures, systems, and components) and the probability of failure. However, a probabilistic safety assessment in current nuclear industries is performed conservatively, assuming that the failure correlation between SSCs is independent or completely dependent. This is an extreme assumption; a reasonable risk can be calculated, or risk-based decision-making can be conducted only when the appropriate failure correlation between SSCs is considered. Thus, this study analyzed the effect of the failure correlation of SSCs on the safety of the system to realize rational safety assessment and decision-making. Consequently, the impact on the system differs according to the size of the failure probability of the SSCs and the AND and OR conditions.

• Nuclear Engineering and Technology
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• 제26권4호
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• pp.555-561
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• 1994

핵연료 건전성 점검을 위하여 다중채널분석기로 얻은 감마선 스펙트럼을 자동으로 빨리 분석하는 프로그램을 개발하였다. 핵연료의 건전성은 실시간 감시와 주기적인 시료 분석을 통한 원자로냉각재내의 방사선준위로 확인된다. 영광 3·4 호기의 경우, 실시간 감시 계통인 프로세스 방사선 감시 계통(PRMS)이 핵연료의 건전성을 확인한다. 현재, PRMS의 스펙트로미터 채널의 신호처리기는 단일채널 분석기이어서 오직 하나의 방사성핵종만을 파악할 수 있다. 따라서 PRMS를 개선하기 위해서는 단일채널분석기를 다중채널분석기로 대치하여야 한다. 이 프로그램은 실시간 모드와 수동모드로 실행되며, 모든 과정을 자동으로 수행한다. 미 국가표준국의 혼합 표준 선원에 대한 시험 결과는 상용 다중채널분석기인 Canberra System 100의 결과와 잘 일치하였다. 결론적으로 개발된 프로그램은 원자력발전소의 감마선 감시에 사용할 수 있을 것으로 보인다.