• Nuclear Engineering and Technology
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• 제41권9호
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• pp.1181-1190
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• 2009

Knowing more about the Local Power Density (LPD) at the hottest part of a nuclear reactor core can provide more important information than knowledge of the LPD at any other position. The LPD at the hottest part needs to be estimated accurately in order to prevent the fuel rod from melting in a nuclear reactor. Support Vector Machines (SVMs) have successfully been applied in classification and regression problems. Therefore, in this paper, the power peaking factor, which is defined as the highest LPD to the average power density in a reactor core, was estimated by SVMs which use numerous measured signals of the reactor coolant system. The SVM models were developed by using a training data set and validated by an independent test data set. The SVM models' uncertainty was analyzed by using 100 sampled training data sets and verification data sets. The prediction intervals were very small, which means that the predicted values were very accurate. The predicted values were then applied to the first fuel cycle of the Yonggwang Nuclear Power Plant Unit 3. The root mean squared error was approximately 0.15%, which is accurate enough for use in LPD monitoring and for core protection that uses LPD estimation.

• Nuclear Engineering and Technology
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• 제30권3호
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• pp.222-234
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• 1998

The Westinghouse OT$\Delta$T DNB protection logic heavily restricts the operation region by applying the same logic for a full range of operating pressure in order to maintain its simplicity. In this work, a fuzzy neural network method is used to estimate the DNB protection limit using the measured average temperature and pressure of a reactor core. Fuzzy system parameters are optimized by a hybrid learning method. This algorithm uses a gradient descent algorithm to optimize the antecedent parameters and a least-squares algorithm to solve the consequent parameters. The proposed method is applied to Yonggwang 3&4 nuclear power plants and the proposed method has 5.99 percent larger thermal margin than the conventional OT$\Delta$T trip logic. This simple algorithm provides a good information for the nuclear power plant operation and diagnosis by estimating the DNB protection limit each time step.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2002년도 추계학술대회 논문집 전력기술부문
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• pp.53-58
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• 2002

Sheath circulating current is increased as the change of sheath mutual impedance which is caused by imbalance of cable system, and different section length between joint box. If excessive current flows in sheath. sheath loss will be increased and then transmission capacity of underground transmission system is reduced. Accordingly, This paper proposed sheath current reduction device using resistor and reactor and proved the reduction effect of that device using EMTP/ATP. And also in this paper, when transients are occurred at the underground system with reduction device by ground fault and lightning surge. we analyzes transient effect of system variously. From this result. authors establish the protection methods of sheath circulating current reduction device.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.795-802
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• 2013

For the improvement of APR1400 Diverse Protection System (DPS) design, the Advanced DPS (ADPS) has recently been developed to enhance the fault tolerance capability of the system. Major fault masking features of the ADPS compared with the APR1400 DPS are the changes to the channel configuration and reactor trip actuation equipment. To minimize the fault occurrences within the ADPS, and to mitigate the consequences of common-cause failures (CCF) within the safety I&C systems, several fault avoidance design features have been applied in the ADPS. The fault avoidance design features include the changes to the system software classification, communication methods, equipment platform, MMI equipment, etc. In addition, the fault detection, location, containment, and recovery processes have been incorporated in the ADPS design. Therefore, it is expected that the ADPS can provide an enhanced fault tolerance capability against the possible faults within the system and its input/output equipment, and the CCF of safety systems.

• 한국화재소방학회논문지
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• 제26권6호
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• pp.38-44
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• 2012

원자력발전소 화재방호의 목표는 화재 시 원자로의 안전정지 상태를 유지하여 환경으로의 방사성물질 누출을 최소화하며, 종사자 인명안전 및 재산을 보호하는데 있다. 소방시설은 발생된 화재를 조기 감지 및 진압하여 화재로 인한 피해를 완화시킬 수 있는 심층방어개념에 입각한 중요한 방어수단의 하나이다. 그러나 소방방재청에서 제시하고 있는 소방시설 설치기준이 원자력발전소에 특화되어 있지 않아 인허가 시 별도의 심의 절차가 요구되고 있다. 또한, 성능위주설계와 같은 규정은 작업자의 인구밀도가 비교적 낮은 원자력발전소에 적용하는데 어려움이 있다. 이 논문에서는 원전 화재방호와 관련된 법령의 상세 검토를 통하여 도출된 근본적인 문제점과 KEPIC FPN의 국내 원전 적용성에 대한 평가를 통하여 소방시설에 대한 기술기준에 대한 개선방향을 제시하였다.

• 한국안전학회지
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• 제32권1호
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• pp.140-144
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• 2017

Fire protection for nuclear power plants should be designed according to the concept of "Defense in Depth" to achieve the reactor safety shutdown. This concept focuses on fire prevention, fire suppression and safe shutdown. Fire prevention is the first line of "Defense in Depth" and the licensee should establish administrative measures to minimize the potential for fire to occur. Administrative measures should include procedures to control handling and use of combustibles. Electrical cables is the major contributor of fire loads in nuclear power plants, therefore electrical cables should be fire retardant. Electrical cables installed in nuclear power plants should pass the flame test in IEEE-383 standard in accordance with NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants". To assure the fire retardant of electrical cables during design life, both aged and unaged cable specimens should be tested in accordance with IEEE-383. It can be generally thought that the flammability of electrical cables has been increased by wearing as time passed, however the results from fire retardant tests performed in U.S.A and Korea indicate the inconsistent tendency of aging and consequential decrease in flammability. In this study, it is expected that the effective methodology for validation of fire retardant performance would be identified through the review of the results from fire retardant tests.

• Journal of Radiation Protection and Research
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• 제11권2호
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• pp.139-145
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• 1986

중성자와 우라늄의 핵반응에 의해 생성된 핵분열생성들의 물리적 특성을 이용하며 원자로 내의 핵연료 상태를 해석하는 모델을 개선하였다. 이 모델에서는 고체 핵연료 내에서 특정핵종의 핵분열 생성물의 생성과 이것이 원자로 냉각재까지 방출되는 과정을 계산하고 추적하여 방사능농도와 결함 핵연료봉의 수를 관계짓는 방정식의 계수들을 결정한다. 핵분열생성들의 거동은 이탈(knock out)과 이동(migration) 두 부분으로 나누어 해석하였으며 트램프 우라늄의 영향을 분리할 수 있도록 하였다. 실측자료로는 가압 경수형 원자로인 고리 원자력발전소 1호기의 1차 냉각재를 분석해서 얻은 I-131과 I-133의 방사능 강도를 이용하였다. 이 실험자료와 위 방정식에서 구한 방사능 강도로부터 구한 결함 핵연료의 수는 제 3 주기에서 $9.34{\pm}1.13$개 제 6 주기에서 $0.294{\pm}0.092$개로 나타났다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2004년도 추계학술대회 논문집 전기설비전문위원
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• pp.83-88
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• 2004

As the electric system is getting larger to meet the increasing demand for electric power, the rating of power apparatus is becoming inevitably higher in its working voltage and larger in its capacity. According to KEPCO reports, power transformers in the KEPCO system have undergone troubles such as winding short insulation breakdowns every year since 1981. the cause of this troubles were high one line grounding fault currents in KEPCO systems that had direct grounding systems. KEPCO has installed the NGR(neutral grounding reactor) to lower this fault current and reduced winding short insulation breakdowns in power transformers. But when a circuit breaker opened a no load bus, some trips of circuit breakers for protecting transformer have occurred by mal-operation of 59GT(overvoltage ground relay) that detect disconnection of NGR. Therefore, in this paper, we analyzed the cause and examined the effect of time delav circuit to prevent wrong operation of 59GT.

• 한국안전학회지
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• 제16권3호
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• pp.125-134
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• 2001

The Advanced Pressurized rater Reactor 1400(APR1400) system is advanced of the successful Korean Nuclear Power Plants(KSNP) design which meets functional needs for safety enhancement reliability improvement, and control in the human-computer monitoring system. Therefore this paper describes the scoring model in order to justify the reliability and safety in APR 1400 under uncertainty. The structure of this paper consists of the human engineering, risk safety, quality function, safety organization management factors of the qualitative factors in chapter 2, and the expectation results of the normalized scoring model in chapter 3. Finally, the proposed reliability model have provided the technical flexibility not only for functional control fields but also for accidents protection systems in APR 1400 under uncertainty.

• 에너지공학
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• 제23권3호
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• pp.96-101
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• 2014

원자력발전소의 원자로보호계통 및 공학적안전설비계통의 계측채널 설정치는 발전소 안전성을 확보하는 데 있으며, 출력조건의 변화시에는 보호계통의 작동이 보장되어야 한다. 본 연구에서는 발전소 운영자료, 시방서 및 운전 매뉴얼등을 사용하여 설정치를 평가하는 데 불확실도의 중요한 요소인 계측기와 process rack drift의 적절성을 확인하고, 설정치 여유가 부족할 경우에 대한 대책을 연구하였다.