• New & Renewable Energy
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• v.7 no.4
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• pp.37-41
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• 2011

This paper presents a Combined solar-wave Power Generation (CPG) concept that the CPG unit is maintained as buoyant at the level of sea water and it is also supported by a submerged tunnel, with the aim of supplying emergency electric power during the station blackout events of nuclear power plants. The CPG concept has been motivated from the 2011 Fukushima-Daiichi Accidents due to the loss of both offsite AC power and emergency diesel power caused by natural hazards such as earthquake and tsunami. The CPG is conceptualized by applying different types and different sites for emergency power generation, in order to reduce common cause failures of emergency power suppliers due to natural hazards. Thus, the CPG can provide a new mean for supplying emergency electric power during station blackout events of nuclear power plants. For this application, the CPG requirements are described with a typical configuration at the ocean side of a submerged tunnel.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1587-1596
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• 2020

The location selection for nuclear power plants (NPP) is a strategic decision, which has significant impact operation of the plant and sustainable development of the region. Further, the ranking of the alternative locations and selection of the most suitable and efficient locations for NPPs is an important multi-criteria decision-making problem. In this paper, the non-sequential Monte Carlo probabilistic method and the Latin hypercube sampling probabilistic method are used to evaluate and select the optimal locations for NPP. These locations are identified by the power plant's onsite loads and the average of the lowest number of relay protection after the NPP's trip, based on electricity considerations. The results obtained from the proposed method indicate that in selecting the optimal location for an NPP after a power plant trip with the purpose of internal onsite loads of the power plant and the average of the lowest number of relay protection power system, on the IEEE RTS 24-bus system network given. This paper provides an effective and systematic study of the decision-making process for evaluating and selecting optimal locations for an NPP.

• Proceedings of the KIEE Conference
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• 1993.07a
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• pp.140-143
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• 1993

Station Blackout(SBO) event means the complete loss of alternating current(AC) electrical power to the essential & nonessential switchgear buses in a nuclear power plants. Since many safety systems in nuclear power plants depend upon alternative current power, the SBO event could be an important contributor to damage of reactor core. Therefore, the SBO events have been considered as a very important safety issues in a nuclear power plants. In this paper, as evaluating the design characteristics of offsite & emergency power systems, an acceptable minimum SBO duration is calculated. And it is presented that the design method for alternative AC(AAC) sources to cope with the SBO events.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.62 no.3
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• pp.300-305
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• 2013

LOOP(Loss of Offsite Power) and SBO(Station Blackout) events have been occurring in nuclear power plants should be reviewed and be controlled on important electrical equipments by professional engineer to prevent and to safety improvement from safety assessment and reliability analysis report. LOOP and SBO occasionally happened by internal or external causes. This paper contained that LOOP frequency in the United States NPPs and in the domestic NPPs have compared and analyzed data by the past lessons and probabilistic statistics. Additionally will be installed MG(Mobile Generator) according to the lessons of Fukushima nuclear accident in Japan, which CDF(Core Damage Frequency) and LOOP frequency have reconsidered. And this paper proposed to reduce reliability criteria using PSA(Probabilistic Safety Analysis).

• Journal of the Korean Society of Safety
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• v.26 no.5
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• pp.105-109
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• 2011

Since a severe accident, which happens in low frequency, can cause serious damages, the interests in off-site consequence analysis for a nuclear power plant have been increased after Chernobyl, TMI and Fukushima accidents. Consequences, which are the effects on health and environment caused by released radioisotopes, are evaluated using MACCS II code based on the method of Level 3 PSA. To perform a consequence analysis for the reference plants, the input data of the code were generated such as meteorological data, population distribution, release fractions, and so on. Using these input data, acute and lifetime dose as an organ, CCDF for early fatalities and latent cancer fatalities, and average individual risk were analyzed by using MACCS II code in this study. These results might contribute to establishing accident management plan and quantitative health object.

• Journal of the Korean Society of Safety
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• v.26 no.5
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• pp.99-104
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• 2011

An ILRT Interval for a nuclear power plant in Korea was extended from once in five years to once in ten years. Therefore, it is necessary to evaluate risk impact for ILRT interval extensions. In this paper, input data were generated for the reference plants, KSNP, using raw data such as meteorological data, population distribution data and source term data. And, using MACCS II code the risk impact assessment was performed based on the two methodologies of NUREG-1493 and NEI Interim Report. The risk impact derived from an ILRT interval extension was identified not to be significant. It is considered to apply this study and results to making an accident management plan and safety goal, and to the field of public acceptance.

• Nuclear Engineering and Technology
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• v.18 no.2
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• pp.97-106
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• 1986

System thermal-hydraulic parameters and simulated, using the best-estimate system code(RELAPS/MODl/NSC), based upon the sequence of events for the KNU1 (Korea Nuclear Unit 1) loss of offsite power transient at 77.5% power which occurred on June 9,1981. The results are compared with the actual plant transient data and show good agreements. After the flow coastdown following the trips of both reactor coolant pumps, the establishment of natural circulation by the temperature difference between the hot and the cold legs is confirmed. The calculated reactor coolant flowrate closely approximates the plant data indicating the validity of relevant thermal-hydraulic models in the RELAP5/MOD1/NSC. Results also show that the sufficient heat removal capability is secured by the appropriate supply of the auxiliary feedwater without the operation of S/G PORVs. In addition, a scenario accident at full power, based upon the same sequence of events described above, is also analysed and the results confirmed that the safety of KNU1 is secured by the appropriate operation of the S/G PORVs coupled with the supply of auxiliary feedwater which ensures sufficient heat removal capability. The characteristics of the non-safety related components such as the turbine stop valve closing time, S/G PORV settings etc. are recognized to be important in the transient analyses on a bestestimate basis.

• Proceedings of the KIEE Conference
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• 2000.11a
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• pp.26-28
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• 2000

국내 원전 소외 전력 계통과 실계통 대상과도 안정도 해석은 대형 해석프로그램인 PSS/E 등에 의존하여 해석해왔으나, 규제 관점에서 보수적인 과도 안정도를 해석할 수 있는 간이 프로그램을 개발하여, 규제 요건 적합성 여부를 확인할 수 있게 하였다. 본 프로그램은 원전 전력계통에 적용되는 품질요건을 적용하지 않는 제어기에 대하여 신뢰를 두지 않았으며, 송전망 보호 계통-보호계전기 및 차단기-에 의존하여 3상 평형 단락 고장 조건 과도 안정도를 해석한 결과, 과도적인 안정도 판별에 적합함을 확인하였다. 가혹한 송전망 조건 아래 원전 발전기의 고장제거 지연시간에 대한 민감도 분석을 통하여 취약한 원전 부지 및 원전 발전기를 확인할 수 있다.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.562-567
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• 1995

The concept of VLDT (Variable Low DNBR Trip), a new CPC trip function, was proposed and applied to the events of increase in secondary heat removal, such as an excess feedwater event anti an IOSGADV (Inadvertent Opening S/G Atmospheric Dump Valve). Major assumption used in this study was no time delay to LOOP (Loss of Offsite Power) after turbine trip. In case of using this VLDT function, safety criterion of DNB would not be violated under the same condition as previous analysis without any change in thermal margin.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.116-124
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• 2019

$M^2LFR-1000$ is a medium-power modular lead-cooled fast reactor, developed by University of Science and Technology of China (USTC), aiming at achieving a reactor design fulfilling the Gen IV nuclear system requirements and meanwhile emphasizing the optimum safety and economics. In order to evaluate the safety performance of $M^2LFR-1000$ reactor core, three typical transients are selected from initiating events, which are unprotected transient overpower (UTOP), unprotected loss of offsite power (ULOHS+ULOF) and increase of feedwater flowrate with primary pumps trip (IFW+PLOF). These three transients presented and discussed in this paper are performed with the code Analysis of THermal-hydraulics of LEaks and Transients (ATHLET), which is developed by Gesellschaft $f{\ddot{u}}r$ Anlagen-und Reaktorsicherheit gGmbH (GRS). The results indicate that the $M^2LFR$ is safe enough under these three transients due to the good inherent safety features of the reactor, without human intervention, the reactor will reach a new steady state under UTOP condition.