• 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.

• Nuclear Engineering and Technology
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• 제44권4호
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• pp.355-362
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• 2012

The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict Pressure Tube (PT) diametral creep employing the previously measured PT diameters and operating conditions. There are twelve bundles in a fuel channel, and for each bundle a linear model was developed by using the dependent variables, such as the fast neutron fluences and the bundle coolant temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3, and 4 of the Wolsung nuclear plant in Korea were used to develop the BPLM. The data from the remaining 10 channels were used to test the developed BPLM. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from Units 2, 3, and 4. Two error components for the BPLM, which are the epistemic error and the aleatory error, were generated. The diametral creep prediction and two error components will be used for the generation of the regional overpower trip setpoint at the corresponding effective full power days. The root mean square (RMS) errors were also generated and compared to those from the current prediction method. The RMS errors were found to be less than the previous errors.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.124-129
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• 1998

A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The $^{239}$ Pu and $^{235}$ U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the fled uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%.. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel has shown that it is desirable to increase the $^{239}$ Pu and $^{235}$ U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, il is recommended to have enrichments of 0.45 and 1.00 wt% for $^{239}$ Pu and $^{235}$ U, respectively.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.641-646
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• 1998

In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUIN $T^{™}$ has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA).

• 전자공학회논문지
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• 제49권9호
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• pp.279-286
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• 2012

최근 범용소프트제어기 설계는 원자력발전소의 첨단주제어실에 적용되고 있다. 범용소프트제어기는 고집적 주제어실에서 비안전 기기뿐만이 아니라 안전기기를 제어할 수 있는 소프트웨어 기반의 수동제어 수단이다. 따라서 범용소프트제어기는 신형 주제어실의 단일 워크스테이션 구현을 위한 필수적인 설계특성을 갖고 있다. 전통적인 주제어실은 컴퓨터 기반으로 하는 통합 운전원 인터페이스 체계로 대체되고 있다. 범용소프트제어기의 오작동신호 발생 가능성을 줄이기 위해 어떠한 기기의 조작을 위해서는 2단계의 구분된 운전원 조작을 요구하는 설계를 고려하였다. 범용소프트제어기 오작동 가능성은 매우 낮기 때문에 범용소프트제어기 그자체로 발전소의 트립 가능성을 증가시키지는 않는다. 범용소프트제어기는 원자력발전소의 계측제어분야/인간연계 분야의 혁신을 대표한다. 범용소프트제어기는 인간연계를 기반으로 하는 단일 표시장치에 다양한 디비젼의 제어와 표시기를 통합하고 있다. 범용소프트제어기의 고장으로부터 안전기능 수행의 영향을 막기 위해 안전기기 및 기능에는 공학적 안전설비 신호가 적용된다. 또한 안전등급 수동스위치는 범용소프트제어기의 신호보다 우선한다. 그러므로 범용소프트제어기의 오작동 신호는 안전관련 스위치로부터의 제어신호에 의해 차단되어질 수 있다.

• Nuclear Engineering and Technology
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• 제22권1호
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• pp.12-19
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• 1990

1981년 6월 9일 고리 1호기 원자력발전소에서 발생한 외부 전원 상실사고 자료를 근거로 RELAP5/MOD2코드모델 평가를 하였다. 계산된 주요 열ㆍ수력학 변수를 실측자료와 비교 분석하였으며 증기발생기의 Nodalization 민감도 분석이 수행되었다. 계산된 열ㆍ수력학 변수는 실측치와 비교적 잘 일치하고 있으며, 이러한 유형의 사고 분석에 RELAP5/MOD2가 적합하다는 것을 보였다. 그러나 가압기 압력과 수위변동에서는 상당한 차이를 보였으며 높게 계산되었다. 이러한 사실은 RELAP5의 수직관에서의 층류 열전달 모델에 기인하는 것으로 해당모델의 개선을 요하고 있다는 것을 알았다. 그리고 증기발생기의 Nodalization 연구를 통하여 수위변동을 잘 예측하기 1위해서는 증기발생기 증기 Dome와 Downcomer사이에 압력을 전달시켜주는 유로를 모델링 하여야 한다는 것을 알았다.

• 한국압력기기공학회 논문집
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• 제17권1호
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• pp.18-27
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• 2021

This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.

• Nuclear Engineering and Technology
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• 제27권6호
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• pp.877-887
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• 1995

영광 3호기는 국내에서 노심보호계통으로_ 노심보호연산기 (CPC)를, 노심감시계통으로 노심운전제한 감시계통 (COLSS)을 사용하는 최초의 원자력발전소이다. CPC는 핵비등 이탈율 및 국부 설출력밀도를 실시간으로 계산하여 노심 조건이 설계 제한치를 초과하면 원자로를 정지시키도록 설계되었다. COLSS는 운전원에게 핵비등 이탈율/선형열출력 여유도, 사분출력 경사비, 및 축방향 출력편차에 대한 기술지침서의 운전제한치을 적용하는데 도움을 제공하고 운전제한치를 초과하는 경우, 경보를 제공하도록 설계되었다. 영광3호기 초기 시운전시험 동안, 다양한 노심 조건에서 CPC/COLSS의 성능을 검증하고 최적의 교정 상수를 얻기 위하여 광범위한 CPC /COLSS 관련시험이 수행되었다. 대부분의 시험결과는 시험허용 범위를 만족하였고, 시험허용 범위를 불만족한 경우에는 시험결과를 분석, 평가하여 문제점을 해결하였다. 각 시험결과를 분석, 평가한 결과 영광 3호기에서 CPC/COLSS가 설계된 데로 성공적으로 설치, 운전되는 것을 확인할 수 있었다.