• Journal of the Korean Society of Safety
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• v.25 no.2
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• pp.24-28
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• 2010

The nuclear power Plant onsite AC electrical power sources are required to supply power to the engineering safety facility buses if the offsite power source is lost. Typically, Diesel Generators are used as the onsite power source. The 125 VAC buses are part of the onsite Class 1E AC and DC electrical power distribution system. The DC power distribution system ensure the availability of DC electrical power for system required to shutdown the reactor and maintain it in a safety condition after an anticipated operational occurrence or a postulated Design Base Accident. Recently, onsite DC power supply system trip occurs the loss of system function. To obtain the performance such as reliability and availability, we analyzed the cause of battery charger trip and described the improvement of DC power supply system reliability. Finally, we provide reliability performance criteria of charger in order to ensure the probabilistic goals for the safety of the nuclear power plants.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.121-130
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• 2021

Cascading failure is the main cause of large blackouts in electrical power systems; this paper analyzes a cascading failure in Hanbit nuclear power plant unit two (2) caused by a circuit breaker (CB) operation failure. This malfunction has been expanded to the loss of offsite power (LOOP). In this study, current practices are reviewed and then the methodologies of how to prevent cascading failures in protection power systems are introduced. An overview on the implementation of IEC61850 GOOSE messaging-based zone selective interlocking (ZSI) scheme as key solution is proposed. In consideration of ZSI blocking time, all influencing factors such as circuit breaker opening time, relay I/O response time and messages travelling time in the communication network should be taken into account. The purpose of this paper is to elaborate on the effect of cascading failure in NPP electrical power protection system and propose preventive actions for this failures. Finally, the expected advantages and challenges are elaborated.

• Nuclear Engineering and Technology
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• v.21 no.1
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• pp.1-11
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• 1989

Numerous losses of decay heat removal capability have occurred at U during stutodwn while its significance to safety is needless to say. A study is carried out as an attempt to assess what could be done to lower the frequency of these events and to mitigate their consequences in the unlikely event that one occurs. The shutdown risk model is developed and analyzed using Event/Fault Tree for the typical pressurized water reactor. The human cognitive reliability (HCR) model, two-stage bayesian approach and staircase function model are used to estimate human reliability, initiating event frequency and offsite power non-recovery probability given loss of offsite power, respectively. The results of this study indicate that the risk of a Pm at shutdown is not much lower than the risk when the plant is operating. By examining the dominant accident sequences obtained, several design deficiencies are identified and it is found that some proposed changes lead to significant reduction in core damage frequency due to loss of cooling events.

• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.175-182
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• 1986

The KNUI (Korea Nuclear Unit 1) loss of offsite power transient as a design-base accident has been simulated using the RELAP5/MOD1/NSC computer code. The analysis is carried out using the best-estimate methodology, but the sequence and its assumptions are based on the evaluation methodology th at emphasizes conservatism. Important thermal-hydraulic parameters such as average temperature, steam generator level and pressurizer water volume are compared with the results in the KNU1 Final Safety Analysis Report (FSAR). The present analysis gives much lower RCS average temperature and pressurizer water volume, and much higher S/G water volume at the turnaround point, which may be considered to be additional improved safety margins. This is expected since the present analysis deals with the best-estimate thermal-hydraulic models as well as the initial conditions on a best-estimate basis. These additional safety margins may contribute to further validate the safety of the KNU1 in this type of accidents(Decrease in Heat Removal by the Secondary System).

• 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.

• 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).

• 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.

• 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 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.