• 한국지진공학회:학술대회논문집
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• 한국지진공학회 2003년도 춘계 학술발표회논문집
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• pp.110-117
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• 2003

The result of recent seismic hazard analysis indicates that the ground motion response spectra for Korean nuclear power plant site have relatively large high frequency acceleration contents. In the ordinary seismic fragility analysis of nuclear power plant structures and equipments, the safety margin of design ground response spectrum is directly used as a response spectrum shape factor. The effects of input response spectrum shape on the floor response spectrum were investigated by performing the direct generation of floor response spectrum from the ground response spectrum. The safety margin included in the design ground response spectrum should be considered as a floor response spectrum shape factor for the seismic fragility analysis of the equipments located in a building.

• 한국안전학회지
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• 제26권5호
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• pp.41-45
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• 2011

The fire event occurred in fire proof zone often causes serious electrical problems such as shorts, ground faults, or open circuits in nuclear power plants. These would be directed to the loss of safe shutdown capabilities performed by safety related systems and equipments The fire event can treat the basic design principle that safety systems should keep their functions with redundancy and independency. In case of a cable fire, operators can not perform their mission properly and can misjudge the situation because of spurious operation, wrong indication or instrument. These would deteriorate the plant capabilities of safety shutdown and make disastrous conditions. In this paper, the cables of the representative nuclear power plant in korea is selected and the cable functional failure temperature by exposed fire using Cable Response to Live Fire(CAROLFIRE) is studied. It is expected that the results are very useful to know the cable failure temperature by exposed fire. We confirmed the safety and integrity of the cable by exposed fire and those results will use the based data of cable exposed fire characteristics.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2007년도 정기 학술대회 논문집
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• pp.217-222
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• 2007

An Emergency Diesel Generator (EDG) is one of the safety related equipments of a Nuclear Power Plant. The seismic capacity of an EDG in nuclear power plants influences the seismic safety of the plants significantly. A recent study showed that the increase of the seismic capacity of the EDG could reduce the core damage frequency (CDF) remarkably. It is known that the major failure mode of the EDG is a concrete coning failure due to a pulling out of the anchor bolts. The use of base isolators instead of anchor bolts can increase the seismic capacity of the EDG without any major problems. This study introduces a seismic risk analysis method and presents sample results about the seismically isolated and conventional EDG system.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 1997년도 춘계 학술발표회 논문집 Proceedings of EESK Conference-Fall 1997
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• pp.217-224
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• 1997

The safety related equipments for use in nuclear power plants should be subjected to the seismic qualification in order to insure the safety of the nuclear power plant. This paper summarizes the seismic qualification test on the Low Voltage Motor Control Centers(MCC's) for use in Wolsong Nuclear Power Plants, Units 2, 3 and 4. The seismic qualification test was performed on the two prototype MCC's(a two-bay wide unit for Phase #1 Test and a five-bay wide unit for Phase #2 Test). The specimens were electrically powered and monitored during the test process. It was demonstrated that the specimens possessed sufficient structural and electrical integrity to withstand the required seismic conditions.

• 동력기계공학회지
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• 제7권3호
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• pp.23-28
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• 2003

In heavy nuclear power plant, high energy through main steam line is provided to turbine that generate the electric power. Since plant had generated power, high noise has been occurred. Noise make equipments and work environment worse. For finding out the location and the cause of making noise, noise was measured along main steam line at open/close test of Main Steam Isolation Valve (MSIV hereafter). As the result, it was identified that the vortex shedding in the cavity of MSIV is main noise source. The profile change of MSIV seat ring was proposed as the method of noise reduction. After filletting MSIV seat ring, the noise level reduced $10{\sim}20dB$ compared before the change of profile.

• 한국지진공학회논문집
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• 제7권4호
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• pp.23-30
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• 2003

최근 수행된 우리나라 원전 부지에 대한 지진재해도 해석 결과 작성된 등재해도 스펙트럼에서 고진동수 성분의 지진동이 매우 우세하게 나타나고 있다. 일반적으로 지진취약도 해석에서는 설계 스펙트럼에 내재된 보수성을 평가하기 위해 스펙트럼 형상계수가 사용된다. 본 연구에서는 입력지반운동 스펙트럼의 형상이 변화함에 따른 층응답스펙트럼의 형상 변화를 분석하였다. 이때 입력 스펙트럼으로부터 직접 층응답스펙트럼을 작성할 수 있는 직접법을 사용하였다. 본 연구 결과 건물 내부에 설치된 기기의 취약도해석에서는 입력스펙트럼에 내재된 보수성을 구조물의 고유진동수에 대한 스펙트럼 형상계수가 아닌 기기의 고유진동수에 따른 층응답스펙트럼 형상계수로 고려하는 것이 정확한 취약도해석 결과를 주는 것으로 나타났다.

• 조명전기설비학회논문지
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• 제23권7호
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• pp.67-74
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• 2009

This paper describes the speed control functions of the typical steam turbine speed controllers and the test results of generator load rejection simulations. The goal of the test is to verify the speed controller's ability to limit the steam turbine's peak speed within a predetermined level in the event of generator load loss. During normal operations, the balance between the driving force of the steam turbine and the braking force of the generator load is maintained and the speed of the turbine-generator is constant. Upon the generator's load loss, in other word, the load rejection, the turbine speed would rapidly increase up to the peak speed at a fast acceleration rate. It is required that the speed controller has the ability to limit the peak speed below the overspeed trip point, which is typically 110[%] of rated speed. If an actual load rejection occurs, a substantial amount of stresses will be applied to the turbine as well as other equipments, In order to avoid this unwanted situation, not an actual test but the other method is necessary. We are currently developing the turbine control system for another nuclear power plant and have plan to do the simulation suggested in this paper.

• 전력전자학회:학술대회논문집
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• 전력전자학회 2018년도 전력전자학술대회
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• pp.177-179
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• 2018

After Fukushima nuclear accident at 2011, nuclear industrial has been focused on operation and maintenance phase, not design and construction phase. Continued good operating performance of nuclear power plants has been the best critical issue to nuclear utilities. Replacement for complete components as well as parts of components is being procured because nuclear utilities must maintain safety and reliability of operating nuclear power plants. However, many suppliers and manufacturers are giving up a nuclear quality assurance program under reduction in new construction of nuclear power plants. It is able to be increased difficulty in procuring spare parts to support operations and maintenance of nuclear power plants. Over 20% of nuclear power plant equipment in some countries is obsolete. Owing to obsolescence of nuclear safety-related items and/or withdrawing a nuclear quality assurance program of suppliers and manufactures, some replacement item and part might be procured to the item not covered by appendix B to USNRC 10 CFR Part 50. Under various methods of the nuclear repair and replacement methodology, utilities are supposed to establish a typical program for a repair and replacement of an electrical equipment and its parts in conjunction with a nuclear quality assurance. Concerning this typical program, this study suggests the repair and replacement methodology of electrical equipments used in nuclear power plants by procurement of a power supply, based on nuclear regulations, codes, standards, guidelines, specific and general technical information, etc..

• 한국화재소방학회논문지
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• 제26권1호
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• pp.10-15
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• 2012

원전 방화지역에서 화재가 발생할 경우 안전정지 케이블은 단락, 접지, 단선 등 전기적 고장으로 케이블에 고장이 발생하여 그 결과로 안전정지계통의 설비와 기기는 안전정지 기능을 상실할 수 있다. 따라서 원자력 발전소 안전정지계통의 설비와 기능은 다중설계, 계열간 독립 및 분리 원칙에 따라 운영되고 있다. 케이블 화재의 경우 기기 오동작과 오지시 또는 잘못된 계측을 유발할 수 있으며 운전원의 상황 판단에 오류를 일으켜 안전정지 기능을 저해할 수 있다. 본 논문에서는 원자력발전소 케이블 화재 조사 및 원인 분석, 케이블 화재온도 및 기능상실 기준 및 국내 원자력발전소에서 사용중인 케이블을 선정하여 노출 화재 시 케이블 기능상실 온도를 분석하여 케이블 안전성을 확인하였다.

• 한국시뮬레이션학회논문지
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• 제24권4호
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• pp.1-9
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• 2015

원자력 분야에서 안전관련(safety-related) 소프트웨어의 활용이 점차 확대됨에 따라서, 그에 상응하는 소프트웨어 안전과 신뢰도 향상을 위한 방안 연구가 지난 10여년 전부터 활발히 진행되고 있다. 원전 계측제어시스템(MMIS)은 원자력 발전소의 두뇌와 신경망에 해당하는 기능을 수행하고 있고 첨단 디지털 장비들로 구성된다. 따라서 원전 계측제어시스템의 소프트웨어 오류는 원자력 발전소 운전에 지장을 초래할 수 있고, 오동작으로 인한 발전소 정지로 경제적 손실을 초래할 수 있다. 소프트웨어 확인 및 검증(verification and validation, V&V)은 소프트웨어 품질을 향상시킬 수 있는 소프트웨어 공학의 분야로 알려져 있고, 원자력 산업계에서는 소프트웨어 생명주기에 따른 철저한 V&V 활동을 이행하고 준수할 것을 법규로 규정하고 있다. V&V 활동은 소프트웨어 전 생명주기에 따라 분석과 시험 활동들의 조합으로 다른 품질관련 공학 업무를 보완하는 역할을 한다. 본 논문에서는 명세 평가, 요건 추적, 소스코드 리뷰, 및 소프트웨어 시험을 통한 최적화된 안전관련 소프트웨어 V&V 방법론에 기반한 소프트웨어 품질 향상 방안과 단계별로 적합한 도구를 활용하여 효율성을 확보할 수 있는 방안을 제시하고자 한다. 제안된 방법론은 실제 신한울 1,2호기 원자력발전소 MMIS 시스템에 적용되어 입증되었다.