• 시스템엔지니어링학술지
• /
• 제17권1호
• /
• pp.85-96
• /
• 2021

Radioactive mixed waste (RMW) means waste mixed with radioactive substances and hazardous substances. In Korea, there are definitions and disposal restrictions on RMW in the Nuclear Safety Management Act, but it is difficult to apply because the contents are insufficient, so this paper proposed applicable management standards. The main RMW generated from nuclear power plants is waste oil, waste asbestos, PCB, and waste fluorescent liquid, and their radiation characteristics are mostly at very low levels and some are estimated at low levels. In addition to nuclear power plants, RMW also occurs in research institutes, industries, and hospitals. The acceptance criteria of all disposal facilities in the world basically prohibit disposal of RMW unless the hazardous substances of RMW are removed or mitigated below the standard value. Cases in Korea, the United States, Japan and Europe were reviewed to propose the RMW management standards in Korea. With reference to the results of the above review, this paper clearly defined RMW and proposed detailed management standards for the separation, storage, treatment and disposal of hazardous substances by applying the Waste Control Act. It also mentioned legislation of management standards, regulatory methods, and acceptance criteria of disposal facility operator.

• Nuclear Engineering and Technology
• /
• 제24권2호
• /
• pp.183-192
• /
• 1992

한국 표준 원전의 부하추종운전 능력 향상을 위하여 노심 축 방향 출력분포 제어에 기존의 regulating bank와는 별개의 heavy bank를 사용하는 노심 제어 기법인 Mode K를 제시하였다 이 경우 heavy bank는 제어할 수 있는 축방향 출력 변화의 범위가 크며, 또한 bank의 움직임과 축 방향 출력 변화가 단조적 관계를 항상 유지할 수 있으므로 운전을 용이하게 할 뿐 아니라 자동제어가 가능하도록 한다. Mode K기법을 이용한 노심 제어 자동화는 부하추종운전에 대한 운전원의 부담을 경감시킬 수 있으며, 특히 주파수 제어 운전등 미리 예측되지 않은 부하변동에 응동할 수 있는 능력을 갖게 한다. 표준 원전의 참조 발전소인 영광 3,4호기의 운전을 모의 계산함으로써 Mode K 의 설계 개념을 평가하였는데 그 결과를 통하여 한국 표준 원전의 실제적 부하추종운전을 위하여 Mode K 기법을 적절하게 사용할 수 있음을 입증하였다.

• Nuclear Engineering and Technology
• /
• 제52권6호
• /
• pp.1180-1187
• /
• 2020

A team of operators is required for nuclear power plant operation, and communication between the operators is an important aspect of the team's ability to successfully carry out tasks. It has been difficult to evaluate the quality of this communication though, and as the relationship between communication quality and team performance has yet to be clarified, it has not been applied to most human reliability analysis (HRA) methodologies. This study investigates the relationship between the quality of communication and team performance using data from a full-scope training simulator of a digital main control room (MCR). Two important characteristics of communication were considered to determine quality: each operator's ability to self-confirm the status of a given task in a digital MCR, and the type of communication, as divided into 1-way, 2-way, and 3-way between operators. To measure team performance, the concept of an unsafe act was employed, which is defined as a human error that has the potential to negatively affect plant safety. Analysis results showed that the communication quality and team performance were related to each other. With this more clearly defined relationship, the results of this study can be applied to related performance shaping factors to improve HRA.

• Nuclear Engineering and Technology
• /
• 제56권3호
• /
• pp.880-885
• /
• 2024

For Safety Assisted Engineering works, real-time simulators have emerged as a mandatory tool among all the key actors involved in the nuclear industry (utilities, designers and safety authorities). EDF, Electricité de France, as the leading worldwide nuclear power plant operator, has a crucial need for efficient and updated simulation tools for training, operating and safety analysis support. This paper will present the work performed at EDF/DT to develop a new generation of engineering simulator to fulfil these tasks. The project is called SiRENE, which is the acronym of Re-hosted Engineering Simulator in French. The project has been economically challenging. Therefore, to benefit from existing tools and experience, the SiRENE project combines: - A part of the process issued from the operating fleet training full-scope simulator. - An improvement of the simulator prediction reliability with the integration of High-Fidelity models, used in Safety Analysis. These High-Fidelity models address Nuclear Steam Supply System code, with CATHARE thermal-hydraulics system code and neutronics, with COCCINELLE code. - And taking advantage of the last generation and improvements of instructor station. The intensive and challenging uses of the new SiRENE engineering simulator are also discussed. The SiRENE simulator has to address different topics such as verification and validation of operating procedures, identification of safety paths, tests of I&C developments or modifications, tests on hydraulics system components (pump, valve etc.), support studies for Probabilistic Safety Analysis (PSA). etc. It also emerges that SiRENE simulator is a valuable tool for self-training of the newcomers in EDF nuclear engineering centers. As a modifiable tool and thanks to a skillful team managing the SiRENE project, specific and adapted modifications can be taken into account very quickly, in order to provide the best answers for our users' specific issues. Finally, the SiRENE simulator, and the associated configurations, has been distributed among the different engineering centers at EDF (DT in Lyon, DIPDE in Marseille and CNEPE in Tours). This distribution highlights a strong synergy and complementarity of the different engineering institutes at EDF, working together for a safer and a more profitable operating fleet.

• 소음진동
• /
• 제5권3호
• /
• pp.361-371
• /
• 1995

The vibration of reactor internals should be monitored and diagnosed for the early detection of the failure of reactor pressure vessel. This can be performed by analyzing the time-history signals from the excore neutron flux detertors. The conventional method is an on-demand system which generates power spectra through Fast Fourier Transform(FFT) algorithm. The operator can make his own decision to detect abnormal vibration using these spectra. This post- processing method, however, requires special expertise in the reactor noise analysis and signal processing for random data. It may mislead the operator into erroneous decision-making, if he is a novice in reactor noise analysis. Hence this study is focused on the automated monitoring and diagnosis procedure for the reactor noise analysis, especially on the Fuzzy algorithm to recognize the pattern of the vibration of Core Suport Barrel. The excore neutron signals of Yonggwang Nuclear Power Plant unit 3 is acquired and analyzed using conventional FFT spectra and tested to adopt the Fuzzy method. An Automated Monitoring and Diagnosis System for CSB Vibration using this Fuzzy method is proposed. Furthermore, vibration data for CSB of Youggwang Nnclear Power Plant unit 3 is presented.

• Transactions on Control, Automation and Systems Engineering
• /
• 제1권2호
• /
• pp.141-146
• /
• 1999

This paper describes the communication method of an auto-startup system. The Auto-Startup system is designed to operate a nuclear power plant automatically during the startup operation . In general , the operations during startup in existing plant have only been manually controlled by the operator. The manual operation caused to the operator mistake. The Auto-Startup system consists of the Distributed Control System(DCS) and G2 (Expert System). Also, Functional Test Facility(FTF) provides the plant's real-data for an Auto-Startup system. So, it is necessary to develop the communication method between these systems. We developed two methods ; one is a network and the other is a hardwire line. To communicate between these systems (DCS-G2 and DCS-FTF) , we developed the communication program. In case of DCS-FTF, we used the TCP/IP and VXI. BUt, in case of DCS-G2 , we , what it called , developed the bridge program using the GSI(G2 Standard Interface). We test to check the function of the important parameter, in time, for analysis of the developed communication method. The results are a good performance when we check the communication time of important parameter. We conclude that Auto-startup system could save heat-up time about at least 5 hours and reduced the change of the reactor operation and trip.

• 방사성폐기물학회지
• /
• 제10권4호
• /
• pp.237-246
• /
• 2012

본 연구에서는 콜로이드와 핵종의 복합이동에 관한 수치모델을 개발하였다. 콜로이드와 핵종의 반응-이동 지배방정식을 풀기 위하여 Operator Splitting Method 중 Strang의 분리 SNI 방식을 수치해석 방법으로 채택하였고 이는 MATLAB을 이용하여 코드화 되었다. 개발된 수치모델은 용질의 이동 및 분산만을 고려한 해석해를 통한 검증과정에서 피어슨 상관계수의 제곱값($r^2$)이 0.99 이상으로 나타나 모델의 정확성이 입증되었다.

• Bulletin of the Korean Chemical Society
• /
• 제14권1호
• /
• pp.21-29
• /
• 1993

For the study of relaxation processes in complex spin system, a general master equation, which can be used to simulate a vast range of pulse experiments, has been formulated using the Liouville representation of quantum mechanics. The state of a nonequilibrium spin system in magnetic field is described by a density vector in Liouville space and the time evolution of the system is followed by the application of a linear master operator to the density vector in this Liouville space. In this master equation the nuclear spin relaxation due to intramolecular dipolar interaction or randomly fluctuating field interaction is explicitly implemented as a relaxation supermatrix for a strong coupled two-spin (1/2) system. The whole dynamic information inherent in the spin system is thus contained in the density vector and the master operator. The radiofrequency pulses are applied in the same space by corresponding unitary rotational supertransformations of the density vector. If the resulting FID is analytically Fourier transformed, it is possible to represent the final nonstationary spectrum using a frequency dependent spectral vector and intensity determining shape vector. The overall algorithm including relaxation interactions is then translated into an ANSIFORTRAN computer program, which can simulate a variety of two dimensional spectra. Furthermore a new strategy is tested by simulation of multiple quantum signals to differentiate the two relaxation interaction types.

• 방사성폐기물학회지
• /
• 제21권1호
• /
• pp.175-182
• /
• 2023

APro, a modularized process-based total system performance assessment framework, was developed at the Korea Atomic Energy Research Institute (KAERI) to simulate radionuclide transport considering coupled thermal-hydraulic-mechanical-chemical processes occurring in a geological disposal system. For reactive transport simulation considering geochemical reactions, COMSOL and PHREEQC are coupled with MATLAB in APro using an operator splitting scheme. Conventionally, coupling is performed within a MATLAB interface so that COMSOL stops the calculation to deliver the solution to PHREEQC and restarts to continue the simulation after receiving the solution from PHREEQC at every time step. This is inefficient when the solution is frequently interchanged because restarting the simulation in COMSOL requires an unnecessary setup process. To overcome this issue, a coupling scheme that calls PHREEQC inside COMSOL was developed. In this technique, PHREEQC is called through the "MATLAB function" feature, and PHREEQC results are updated using the COMSOL "Pointwise Constraint" feature. For the one-dimensional advection-reaction-dispersion problem, the proposed coupling technique was verified by comparison with the conventional coupling technique, and it improved the computation time for all test cases. Specifically, the more frequent the link between COMSOL and PHREEQC, the more pronounced was the performance improvement using the proposed technique.

• Nuclear Engineering and Technology
• /
• 제27권2호
• /
• pp.203-215
• /
• 1995

소형냉각재 상실사고시 원자로냉각재펌프( RCP )의 지속적인 운전은 원자로냉각재의 불필요한 누출을 초래하여 심각한 노심노출 및 이에따른 핵연료 손상을 야기시킬 수 있다. TMI 사고 후 미국 NRC의 요구에 따라 CE형 발전소 사용자 단체에서는 “T2/L2”라는 RCP 트립전략을 개발하여 CE형 발전소에 적용 가능토록 일반비상운전지침서에 반영하였다. 상기 T2/L2 RCP 트립전략은 사고후 원자로 냉각재 계통의 압력이 감소하여 RCP 트립설정치에 도달하면 처음 두대의 RCP를 우선 정지시키고, 사고가 LOCA임이 확인되면 나머지 두대의 RCP를 정지시키는 방식을 채택하고 있다. 본 논문에서는 영광3, 4호기의 RCP 트립설정치를 분석, 선정하고 T2/L2 전략의 안전운전양상을 입증하였다 분석결과, 최악의 파단크기로 밝혀진 0.15 ft$^2$의 고온관 파단 LOCA 영광3, 4호기 RCP 트립설정치는 가압기 압력 1775 psia로 나타났으며, 운전원이 마지막 두대의 RCP를 트립시키지 못하였을 경우 혹은 최악의 시점에서 정지시켰을 경우에도 영광3, 4호기의 노심냉각능력은 확보될 수 있음이 확인되었다. 또한 영광3, 4호기의 RCP 트립전략은 미국 NRC가 요구하는 최대 핵연료피복재온도 관점에서의 10 CFR 50.46 요구조건과 운전원 조치시간 관점에서의 ANSI 58.8 요구조건도 충분히 만족함이 판명되었다 따라서, 1775 psia의 RCP 트립설정치를 사용한 영광3, 4호기의 T2/L2 RCP 트립전략은 사고시 운전원에게 향상된 운전지침을 제공할 수 있을 것으로 판단된다.