• Journal of the Earthquake Engineering Society of Korea
• /
• v.19 no.5
• /
• pp.207-217
• /
• 2015

In order to increase seismic performance of nuclear power plant (NPP) in strong seismic zone, lead-rubber bearing (LRB) can be applied to seismic isolation system of NPP structures. Simple equivalent linear model as structural analysis model of LRB is more widely used in initial design process of LRB than a bilinear model. Seismic responses for seismically isolated NPP containment structures subjected to earthquakes categorized into 5 different soil-site classes are calculated by both of the equivalent linear- and bilinear- LRB models and compared each others. It can be observed that the maximum displacements of LRB and shear forces of containment in the case of the equivalent linear LRB model are larger than those in the case of bilinear LRB model. From the seismic fragility curves of NPP containment structures isolated by LRB, it can be observed that seismic fragility in the case of equivalent linear LRB model are about 5~30 % larger than those in the case of bilinear LRB model.

• Proceedings of the Korea Water Resources Association Conference
• /
• 2011.05a
• /
• pp.402-402
• /
• 2011

식생의 성장은 수문순환과 밀접히 연관되어 있으며 특히 강수량은 식생의 연간변동성의 주된 요인이다. 하지만 국내의 경우 이러한 식생과 강수량과의 관계를 알아보기 위한 정량적인 접근은 시도되지 않은 실정이다. 순일차생산량(NPP)은 광합성으로부터 생성된 유기물질에서 성장에 필요한 호흡량을 뺀 것이며, 이는 식생량을 산정할 때 주로 이용된다. 본 연구에서는 우리나라 전역의 50 지역의 5년간의 NPP와 연강수량이 분석에 사용되었다. 비록 자료기간이 관계를 정립하기에는 다소 부족하며 NPP자료 자체의 불확실성이 내포되어 있지만 각 지점별 NPP와 강수량과의 상관성이 매우 높은 지역들이 있는 것으로( ��0.6) 나타났다. 하지만 모든 지역의 자료를 동시에 고려하였을 경우에는 결정계수가 크게 감소하였다($R^2$=0.310). 추가적으로 강우이용효율과 연강수량의 관계를 분석하였으나 결정계수가 0.022로 기존의 문헌과는 달리 이들 사이에 뚜렷한 관계가 나타나지는 않았다. 본 연구를 통하여 확인된 사실은 우리나라의 식생과 강수량 관계에서는 우리나라 전역을 전반적으로 설명할 수 있는 일반적인 선형 관계를 찾기는 힘들다는 것이며, 다만 지역별로 살펴볼 경우 식생-강수량 관계가 비교적 잘 맞는 지역들이 다수 존재한다는 것이다. 따라서 우리나라를 대상으로 한 생태적요소가 고려된 수문모형을 개발할 경우 지역적 특성들이 모형 구조에 반영되어야 할 것이다.

• Nuclear Engineering and Technology
• /
• v.37 no.3
• /
• pp.207-220
• /
• 2005

Nuclear power plants (NPP) are complex socio-technological systems that rely on the success of both hardware and human components. Empirical studies of plant operating experience show that human errors are important contributors to accidents and incidents, and that organizational factors play an important role in creating contexts for human errors. Current probabilistic safety assessments (PSA) do not explicitly model the systematic contribution of organizational factors to safety. As some countries, like the United States, are moving towards increased use of risk information in the regulation and operation of nuclear facilities, PSA quality has been identified as an area for improvement. The modeling of human errors, and underlying organizational weaknesses at the root of these errors, are important sources of uncertainty in existing PSAs and areas of on-going research. This paper presents a review of research into the following questions: Is there evidence that organizational factors are important to NPP safety? How do organizations contribute to safety in NPP operations? And how can these organizational contributions be captured more explicitly in PSA? We present a few past incidents that illustrate the potential safety implications of organizational deficiencies, some mechanisms by which organizational factors contribute to NPP risk, and some of the methods proposed in the literature for performing root-cause analyses and including organizational factors in PSA.

• Journal of the Earthquake Engineering Society of Korea
• /
• v.19 no.1
• /
• pp.1-11
• /
• 2015

In order to perform a soil-isolation-structure interaction analysis of seismically isolated nuclear power plant (NPP) structures, the nonlinear behavior of a seismic isolation system may be converted to an equivalent linear model used in frequency domain analysis. Seismic responses for seismically isolated NPP containment structures subjected to a simple artificial acceleration history and different site class earthquakes are evaluated for the equivalent-linear and nonlinear models that have been applied to lead-rubber bearing (LRB) modeling. It can be observed that the maximum displacements of the equivalent linear model are larger than that of the nonlinear model. From the floor response spectrum analysis for the top of NPP containment structures, it can be observed that the spectral acceleration of an equivalent linear model at about 0.5 Hz frequency is about 2~3 times larger than that of a nonlinear model.

• Journal of the Earthquake Engineering Society of Korea
• /
• v.17 no.2
• /
• pp.53-59
• /
• 2013

Several researches have been studied to enhance the seismic performance of nuclear power plants (NPPs) by application of seismic isolation. If a seismic base isolation system is applied to NPPs, seismic performance of nuclear power plants should be reevaluated considering the soil-structure interaction effect. The seismic fragility analysis method has been used as a quantitative seismic safety evaluation method for the NPP structures and equipment. In this study, the seismic performance of an isolated NPP is evaluated by seismic fragility curves considering the soil-structure interaction effect. The designed seismic isolation is introduced to a containment building of Shin-Kori NPP which is KSNP (Korean Standard Nuclear Power Plant), to improve its seismic performance. The seismic analysis is performed considering the soil-structure interaction effect by using the linearized model of seismic isolation with SASSI (System for Analysis of Soil-Structure Interaction) program. Finally, the seismic fragility is evaluated based on soil-isolation-structure interaction analysis results.

• Proceedings of the Korea Concrete Institute Conference
• /
• 2006.11a
• /
• pp.237-240
• /
• 2006

Almost 30 years have been passed since the first nuclear power plant was operated in Korea. Many studies have been actively conducted from the early 1990's in order to develop the deterioration management system for concrete structures in NPPs(Nuclear Power Plants) accordingly. Base on these studies, a systematic deterioration management program has developed and operated since 1997. According to this program, systematic inspections to provide database and evaluation were periodically performed (every overhaul at intervals of $12{\sim}18$ month and every five years). Accumulated deterioration database was usefully utilized for the NPP PSR (Periodic Safety Review). In this paper, the long-term durability and integrity of Kori 1,2 NPP concrete structures which are the oldest ones in Korea were evaluated based on the precise inspection database and regulatory inspection results including compressive strength, depth of carbonation, amount of chlorination and spontaneous potential of reinforcing bar, etc. It was noted that Kori 1,2 NPP structures have not any serious durability problems.

• Nuclear Engineering and Technology
• /
• v.28 no.3
• /
• pp.290-298
• /
• 1996

Direct Containment Heating, DCH, is supposed to occur during a core melt-down accident if the primary system pressure is still high at the time of vessel breach in a Nuclear Power Plant (NPP). In this case, DCH is considered to be one of very important severe phenomena during postulated severe accident scenario because of the fast heat transfer rate to atmosphere and the sharp pressure increase in a containment. To reduce the effect of this DCH phenomena, the capture volume wes designed at Ulchin NPP units 3 and 4. But, the effect of this has not been studied extensively. This work consists of experimental and numerical analyses of the effects of capture volume in the cavity on DCH phenomena. The experimental model is a 1/30 scaled-down model of Ulchin NPP units 3 and 4. We used three types of capture volumes to investigate the effect of size. Numerical analysis using CONTAIN 1.2 is performed with the correlation for the dispersed fraction of molten corium from the cavity into the containment derived from the experimental data to examine the effect of capture volume on DCH phenomena in full scale of Ulchin NPP units 3 and 4.

• Nuclear Engineering and Technology
• /
• v.50 no.6
• /
• pp.860-868
• /
• 2018

This article presents a new design of earthquake instrumentation that is suitable for quick decision-making after the seismic event at the nuclear power plant (NPP). The main objective of this work is to ensure more availability of the NPP by expediting walk-down period when the seismic wave is incident. In general, the decision-making to restart the NPP after the seismic event requires more than 1 month if an earthquake exceeds operating basis earthquake level. It affects to the plant availability significantly. Unnecessary shutdown can be skipped through quick assessments of operating basis earthquake, safe shutdown earthquake events, and damage status to structure, system, and components. Multidecision parameters such as cumulative absolute velocity, peak ground acceleration, Modified Mercalli Intensity Scale, floor response spectrum, and cumulative fatigue are discussed. The implementation scope on the field-programmable gate array platform of this work is limited to cumulative absolute velocity, peak ground acceleration, and Modified Mercalli Intensity. It can ensure better availability of the plant through integrated decision-making process by automatic assessment of NPP structure, system, and components.

• Nuclear Engineering and Technology
• /
• v.40 no.4
• /
• pp.269-276
• /
• 2008

Operational license of WWER-440/213 units at Paks NPP, Hungary is limited to the design lifetime of 30 years. Prolongation by additional 20 years of the operational lifetime is feasible. Moreover, enhancement of the reactor thermal power by 8% will increase both the net power output and the competitiveness of the plant. Paks NPP is a pioneer considering the power up-rate and preparation of long-term operation of WWER-440/213 design. Systematic preparatory work for long-term operation of Paks NPP has been started in 2000. A regulatory framework and a comprehensive engineering practice have been developed. According to the authors view, creation of a gapless engineering system via consequent application of best practices, and feed-back of experiences together with proper consideration of WWER-440/V213 features are the decisive elements of ensuring the safety of long-term operation. That systematic engineering approach is in the focus of recent paper. Key elements of justification and measures for ensuring the safety of long-term operation of Paks NPP WWER-440/213 units are identified and discussed. These are the assessment of plant condition and review of adequacy of ageing management programmes, also the review, validation and reconstitution of time limited ageing analyses as core tasks of licence renewal.

• Nuclear Engineering and Technology
• /
• v.53 no.9
• /
• pp.2888-2898
• /
• 2021

Vital area identification (VAI) is an essential procedure for the design of physical protection systems (PPSs) for nuclear power plants (NPPs). The purpose of PPS design is to protect vital areas. VAI has been improved continuously to overcome the shortcomings of previous VAI generations. In first-generation VAI, a sabotage fault tree was developed directly without reusing probabilistic safety assessment (PSA) results or information. In second-generation VAI, VAI model was constructed from all PSA event trees and fault trees. While in third-generation VAI, it was developed from the simplified PSA event trees and fault trees. While VAIs have been performed for NPPs in full-power operations, VAI for NPPs in low-power and shutdown (LPSD) operations has not been studied and performed, even though NPPs in LPSD operations are very vulnerable to sabotage due to the very crowded nature of NPP maintenance. This study is the first to research and apply VAI to LPSD operation of NPP. Here, the third-generation VAI method for full-power operation of NPP was adapted to the VAI of LPSD operation. In this study, LPSD VAI for a few plant operational states (POSs) was performed. Furthermore, the operation strategy of vital areas for both full-power and LPSD operations was discussed. The LPSD VAI method discussed in this paper can be easily applied to all POSs. The method and insights in this study can be important for future LPSD VAI that reflects various LPSD operational states. Regulatory bodies and electric utilities can take advantage of this LPSD VAI method.