• 한국안전학회지
• /
• 제29권6호
• /
• pp.166-171
• /
• 2014

Nuclear power plants are equipped with the reactor trip system (RTS) and the engineered safety features actuation system (ESFAS) to improve safety on the normal operation. In the event of the design basis accident (DBA), a various of post accident monitor(PAM)systems support to provide important details (e.g. Containment pressure, temperature and pressure of reactor cooling system and core exit temperature) to determine action of main control room (MCR). Operator should be immediately activated for the accident mitigation with the information. Especially, core exit temperature is a critical parameter because the operating mode converts from normal mode to emergency mode when the temperature of core exit reaches $649^{\circ}C$. In this study, uncertainty which was caused by exterior environment, characteristic of thermocouple/connector and accuracy of calibrator/indicator was evaluated in accordance with ANSI-ISA 67.04. The square root of the sum of square (SRSS) methodology for combining uncertainty terms that are random and independent was used in the synthesis. Every uncertainty that may exist in the hardware which is used to measure the core exit temperature was conservatively applied and the associative relation between the elements of uncertainty was considered simultaneously. As a result of uncertainty evaluation, the channel statistical allowance (CSA) of single channel of core exit temperature was +1.042%Span. The range of uncertainty, -0.35%Span ($-4.05^{\circ}C$) ~ +2.08%Span($24.25^{\circ}C$), was obtained as the operating criteria of core exit temperature.

• Nuclear Engineering and Technology
• /
• 제53권10호
• /
• pp.3236-3246
• /
• 2021

Lessons learned from the Fukushima Daiichi nuclear power plant accident directed that multiple failures should be considered more seriously rather than single failure in the licensing bases and safety cases because attempts to take accident management measures could be unsuccessful under the high radiation environment aggravated by multiple failures, such as complete loss of electric power, uncontrollable loss of coolant inventory, failure of essential safety function recovery. In the case of the complete loss of electric power called station blackout (SBO), if there is no mitigation action for recovering safety functions, the reactor core would be overheated, and severe fuel damage could be anticipated due to the failure of the active heat sink. In such a transient condition at CANDU-6 plants, the seal failure of the primary heat transport (PHT) pumps can facilitate a consequent increase in the fuel sheath temperature and eventually lead to degradation of the fuel integrity. Therefore, it is necessary to specify the regulatory guidelines for multiple failures on a licensing basis so that licensees should prepare the accident management measures to prevent or mitigate accident conditions. In order to explore the efficiency of implementing accident management strategies for CANDU-6 plants, this study proposed a realistic accident analysis approach on the SBO transient with multiple-failure sequences such as seal failure of PHT pumps without operator's recovery actions. In this regard, a comparative study for two PHT pump seal failure modes with and without coolant seal leakage was conducted using a best-estimate code to precisely investigate the behaviors of thermal-hydraulic parameters during transient conditions. Moreover, a sensitivity analysis for different PHT pump seal leakage rates was also carried out to examine the effect of leakage rate on the system responses. This study is expected to provide the technical bases to the accident management strategy for unmitigated transient conditions with multiple failures.

• 한국안전학회지
• /
• 제31권5호
• /
• pp.177-186
• /
• 2016

When an event is occurred in a nuclear power plant (NPP), the NPP operator reports it referred by the regulation on reporting and public announcement of accidents and incidents. Some of the events do not need to be reported because they are not included in the reporting criteria of the regulation. However, it is necessary that they should be managed effectively because the accident can be occurred by the recurrence of a lot of them as precursors. Among the events not included in the reporting criteria of the regulation, near miss is the event that is not occurred but can generate a significant consequence. This can provide the cause of the event which does not result an accident. So, it is able to offer insightful knowledges to prevent higher level events about the function and process of NPP. The objective of this study is to analyze the issues of near miss events, prepare the defence against the risk, and improve the management process of NPP. To achieve it, this study performed to analyze the management structure and status of near miss events as well as the accident reporting system of the domestic and foreign regulation bodies. In case of Korea, the status was analyzed by quantitative data, licensee event reports and procedures. Based on these, we could find the causes that near miss events were not managed effectively. Then, systematic alternatives that reflected the perspective of man, technology and organization were drawn.

• Nuclear Engineering and Technology
• /
• 제41권7호
• /
• pp.945-952
• /
• 2009

The Three Mile Island Unit 2 (TMI-2) accident provides unique full scale data, thus providing opportunities to check the capability of codes to model overall plant behavior and to perform a spectrum of sensitivity and uncertainty calculations. As part of the TMI-2 analysis benchmark exercise sponsored by the Organization for Economic Cooperation and Development Nuclear Energy Agency (OECD NEA), several member countries are continuing to improve their system analysis codes using the TMI-2 data. The Republic of Korea joined this benchmark exercise in November 2005. Seoul National University has analyzed the TMI-2 accident as well as the currently proposed alternative scenario along with a sensitivity study using the Modular Accident Analysis Program Version 4.03 (MAAP4.03) code in collaboration with the Korea Hydro and Nuclear Power Company. Two input files are required to simulate the TMI-2 accident with MAAP4: the parameter file and an input deck. The user inputs various parameters, such as volumes or masses, for each component. The parameter file contains the information on TMI-2 relevant to the plant geometry, system performance, controls, and initial conditions used to perform these benchmark calculations. The input deck defines the operator actions and boundary conditions during the course of the accident. The TMI-2 accident analysis provided good estimates of the accident output data compared with the OECD TMI-2 standard reference. The alternative scenario has proposed the initial event as a loss of main feed water and a small break on the hot leg. Analysis is in progress along with a sensitivity study concerning the break size and elevation.

• Nuclear Engineering and Technology
• /
• 제44권5호
• /
• pp.507-522
• /
• 2012

When decommissioning a nuclear power plant, it is difficult to make an appropriate plan to ensure sufficient space for temporary placement and conveyance operations of dismantling targets. This paper describes a system to support temporary placement and conveyance operations using augmented reality (AR). The system employs a laser range scanner to measure the three-dimensional (3D) information of the environment and a dismantling target to produce 3D surface polygon models. Then, the operator simulates temporary placement and conveyance operations using the system by manipulating the obtained 3D model of the dismantling target in the work field. Referring to the obtained 3D model of the environment, a possible collision between the dismantling target and the environment is detectable. Using AR, the collision position is presented intuitively. After field workers evaluated this system, the authors concluded that the system is feasible and acceptable to verify whether spaces for passage and temporary storage are sufficient for temporary placement and conveyance operations. For practical use in the future, some new functions must be added to improve the system. For example, it must be possible for multiple workers to use the system simultaneously by sharing the view of dismantling work.

• Nuclear Engineering and Technology
• /
• 제18권2호
• /
• pp.78-84
• /
• 1986

일차섭동이론을 이용하여 물질밀도 감도 계수의 표현식을 유도하였다. Super-Phenix I 평형노심의 초기상태를 기준계로 택했으며 유효중배계수를 계의 응답함수로 정의했다. 볼츠만 연산자의 구성요 소인 물질밀도로 표현되는 핵연료의 농축도와 실효밀도를 입력변화로 선정했다. 위 계산을 수행하는데 전산코드시스템 (KAERI-26군 단면적 library/1DX/2DB/PERT-V)가 사용되었다. 핵연료 농축도의 유효증배계수에 대한 감도계수는 4.576로 계산되었으며, 핵연료 실효밀도의 감도 계수는 0.0756으로 계산되었다. 본 연구는 감도해석법이 대형전산코드를 이용한 직접반복계산법에 비해 계산시간의 단축과 아울러 많은 정보를 준다는 것을 보여준다.

• 대한핵의학회지
• /
• 제33권2호
• /
• pp.131-142
• /
• 1999

목적: 휴식 부하 심근관류 SPECT로 관상동맥질환을 진단하고 관상동맥협착을 찾을 때 게이트 SPECT 방법과 감쇠보정 후 관류 SPECT를 사용하면 특이도가 올라가서 진단 성능이 향상된다는 보고가 있다. 이 연구는 임상적으로 중간 정도의 관상동맥질환 유병 가능성을 보이는 환자에서 게이트 SPECT가 진단 성능을 향상시키는지 게이트 감쇠보정 SPECT를 시행하여 조사하였다. 대상 및 방법: 휴식기 T1-201 디피리다몰 부하 Tc-99m-MIBI SPECT를 할 때 둘 다 감쇠보정 영상을 얻고 Tc-99m-MIBI SPECT는 게이트 SPECT로 얻어 검사성능을 비교하였다. 혈관조영술로 진단된 단일 혈관질환 13명, 두 혈관질환 18명, 세 혈관질환 8명과 정상임을 확인한 29명을 합한 모두 68명의 환자에서 세 판독자가 독립적으로 각 동맥의 협착 유무와 질환 유무를 5 등급으로 점수화하여 수신자 특성 곡선을 그렸다. 결과: Hanley와 McNeil의 방법으로 곡선 아래 면적을 구하고 비교하여 유의한 차이가 있는지 보았으나 판독자나 어느 동맥영역인지에 상관없이 유의한 차이를 찾지 못하였다. 등급 3보다 큰 등급을 지정한 경우 검사 양성으로 보아 계산한 예민도와 특이도도 유의하게 차이 없었다. 결론: 우리는 이 결과를 보고 관상동맥질환의 검사 전 가능성이 중간 정도인 환자에서는 판독자나 동맥에 상관없이 게이트 SPECT를 더하거나 감쇠보정 SPECT를 더하여 보아도 진단 성능이 향상되지 않는다고 생각하였다.

• 대한전기학회:학술대회논문집
• /
• 대한전기학회 2005년도 심포지엄 논문집 정보 및 제어부문
• /
• pp.84-86
• /
• 2005

The digital systems such as PLC or DCS have been applied to non-safety systems of nuclear power plants because of many difficulties in using analog systems. Nowadays, digital systems have been applied to safety systems of the plants such as reactor protection system. One of the main advantages of digital systems is applicability of automatic testing methods to the systems. The protection system requires high-reliability and high-availability because it shall minimize the propagation of abnormal or accident conditions of nuclear power plants. The calculation of reliability and availability of systems depends on the maintenance period of the system. In general, the maintenance period of the protection system is one-month in case of the manual test. However, the cycle of test can be shortened in several hours by using automatic periodic testing. The reliability and availability of the system is better when test period is shortened because the reliability and availability is inverse proportion to the test period. In this research, we developed the automatic periodic testing method for KNICS Reactor Protection System, which can test the system automatically without an operator or a tester. The automatic testing contained all functions of reaction protection systems from analog-to-digital conversion function of the bistable Processor to the coincident trip function of the coincident processor. By applying the automatic periodic testing to reaction system, the maintenance cost can be cut down and the reliability can be increased.

• Nuclear Engineering and Technology
• /
• 제33권3호
• /
• pp.339-348
• /
• 2001

NSSS Control Systems automatically mitigate transient conditions and leads to a stable plant condition without operator actions when a transient occurs during normal power operation. In this paper, the function and performance of NSSS control systems were examined and evaluated by comparing the predicted results with the measured data for the selected events. Loss of a Main Feedwater Pump and Load Rejection to House Load Operation events were selected for the evaluation among the transient tests peformed during the Power Ascension Test (PAT) of UCN unit 4. The overall schematic control actions of NSSS control systems can be evaluated easily through the observation of these two typical events. The selected events were analyzed by the KISPAC computer code[l] which had been used in developing the control logic and determining the control setpoints during the plant design. Additionally, the performance of FWCS during low power operation was evaluated. The result of evaluation showed that the NSSS control systems were designed properly and the performance of the NSSS control systems was excellent and also the computer code had a good prediction capability.

• 한국콘텐츠학회논문지
• /
• 제10권11호
• /
• pp.341-351
• /
• 2010

경보시스템은 원자력발전소의 안전을 위해 매우 중요하다. 운전원은 경보발생시 경보와 그 원인 간의 논리적 관계를 파악하기 위해 논리도면을 참조한다. 이 논문은 전산화된 월성원자력발전소 3, 4호기 경보 논리도면에서 경보-원인 경로를 추적하는 시스템을 제안한다. 또한 논리도면의 전산화를 위하여 논리도면을 2차원 문자배열로 표시하고 이를 검증할 수 있는 문법을 제시였다. 그리고 추적된 경보-원인 간 논리경로와 논리 상태를 표시하기위하여 ND 및 DC 연산을 제안하였다. 이 시스템은 현재 월성원자력발전소에서 운영 중에 있다.