• 시스템엔지니어링학술지
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• 제10권1호
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• pp.33-41
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• 2014

New design concept of Reactor Coolant System (RCS) including a reactor assembly for the SMR is introduced in this work. An exploration of new type of reactor that is advanced from proposed SMRs is performed by using systems engineering approach. In this point of view project structured on three main phases; needs analysis (NA), concept exploration (CE), and concept definition (CD). Main objectives as an output of the CE stage are a small size, low cost, shortening the schedule, and enhancing safety. The SMRs usually have a small size requirement. In order to meet the size requirement and to achieve a productivity, in other words, easiness to manufacture, this paper suggests an integrated PWR design concept through researching predecessors. Although the integrated PWR concept provides many advantages, it has disadvantages that composite of maintenance and a low availability problem. Therefore, this paper comes up with a run-to-fail design concept based on modular design to address the maintenance problem and to maximize the availability of SMRs as well as to be compatible with the overall-SMRs including Barge Mounted(BM)type.

• Journal of Radiation Protection and Research
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• 제14권1호
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• pp.34-45
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• 1989

프랑스의 1300 MWe 급(級) 표준(標準) P'4형 PWR 원전(原電)의 일차냉각재상실사고(一次冷却材喪失事故)(LOCA)시(時) 원전(原電) 주제어가내(主制御家內) 운전원(運轉員)에 대한 고사선(故射線) 피습선량(被濕線量)을 계산하여 주제어실(主制御室)의 체류안전성(滯留安全性)을 평가(評價)하였다. 본(本) 평가(評價)에서 사용(使用)된 제가정(諸假定)은 프랑스의 표준안전성분석보고서(漂準安全性分析報告書)에 따랐다. 본(本) 평가(評價)를 위하여 LOCA 사고시(事故時) 원자로건물외(原子爐建物外)로 방출(放出)되는 방사핵종(放射核種)의 방사능(放射能), 주제어실(主制御室)에서의 체적인자(體積因子) 및 제어실내(制御室內) 운전원(運轉員)의 전신(全身) 및 갑상선(甲狀膳) 피폭선량(被爆線量)을 사고발생후(事故發生後) 30일까지 전산(電算)할 수 있는 간단한 전산(電算)프로그램, COREX를 개발(開發)하였다. 본(本) 연구(硏究)에서 얻어진 계산결과(計算結果)는 대체적으로 프랑스의 EDF(불란서 전력주식회사(電力株式會社) 에서 제안(提案)한 결과(結果)와 대체적으로 잘 일치(一致)하였으나, 전신외부피폭선량(全身外部被爆線量)의 값은 일부(一部) 체적인자(體積因子) 값의 차이로 인(因)하여 일부 편차(偏差)를 보였다.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.33-44
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• 2006

This paper presents a review of small-to-medium-sized, pressurized-water-cooled nuclear power reactors whose major primary coolant systems are integrated into a reactor pressure vessel, the concepts categorized as Integrated Primary System Nuclear Reactors (IPSRs). Typical examples of these proposals of interest in this review are CAREM, SMART, IRIS and IMR, all of which are being aimed at the near term deployment. Emphasis is placed on thermal hydraulic aspects. A brief characterization of the IPSR concepts is made and comparisons of plant key parameters are shown. Discussions will follow for the core cooling under rated power conditions and natural circulation heat removal on the basis of the design data available in the public domain.

• 한국전산구조공학회논문집
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• 제27권2호
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• pp.103-111
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• 2014

본 연구에서는 비선형 유한요소 해석 기법을 적용한 격납건물의 내압취약도 평가를 수행하였다. 대상 구조물은 국내 대표적인 가압경수로형 원전 격납건물 중 하나로 하였다. 비선형 극한내압 해석을 위해 대규모 개구부를 고려한 격납건물의 3차원 유한요소 모델을 도출하였다. 재료 특성 및 구조적 성능에 내포된 불확실성을 고려하기 위하여 각 변수들의 변동성에 대한 극한내압 성능의 민감도 해석을 수행하였다. 민감도 해석 결과를 통해 확률론적 내압 취약도 평가를 위한 불확실성 변수 및 분포 특성을 도출하였다. 현재의 텐던 긴장력 상태를 고려하기 위하여 가동 중 검사 보고서에 기록된 텐던 긴장력 값을 중앙값으로 적용하였다. 누설(leak)과 파단(rupture)을 파괴모드로 정의하고, 각각에 대한 극한내압 취약도 평가를 위하여 한계상태를 정의하였다. 각 파괴모드에 대한 대상 격납건물의 내압취약도를 내압 성능 중앙값, 고신뢰도 저파괴확률 성능값, 신뢰도 수준에 따른 취약도 곡선을 통하여 제시하였다. 누설 및 파단 파괴모드에 대한 고신뢰도 저파괴확률값은 각각 0.7991 MPa, 0.8691 MPa로 평가되었다.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2427-2444
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• 2021

The lack of plant-side energy storage analysis to support nuclear power plants (NPP), has setup this research endeavor to understand the characteristics and role of specific storage technologies and the integration to an NPP. The paper provides a qualitative review of a wide range of configurations for integrating the energy storage system (ESS) to an operating NPP with pressurized water reactor (PWR). The role of ESS technologies most suitable for large-scale storage are evaluated, including thermal energy storage, compressed gas energy storage, and liquid air energy storage. The methods of integration to the NPP steam cycle are introduced and categorized as electrical, mechanical, and thermal, with a review on developments in the integration of ESS with an operating PWR. By adopting simplified off-design modeling for the steam turbines and heat exchangers, the results show the performance of the PWR steam cycle changes with respect to steam bypass rate for thermal and mechanical storage integration options. Analysis of the integrated system characteristics of proposed concepts for three different ESS suggests that certain storage technologies could support steady operation of an NPP. After having reviewed what have been accomplished through the years, the research team presents a list of possible future works.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.148-164
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• 2017

The focus on the issues surrounding spent nuclear fuel and lifetime extension of old nuclear power plants continues to grow nowadays. A transparent decision-making process to identify the best suitable nuclear fuel cycle (NFC) is considered to be the key task in the current situation. Through this study, an attempt is made to develop an equilibrium model for the NFC to calculate the material flows based on 1 TWh of electricity production, and to perform integrated multicriteria decision-making method analyses via the analytic hierarchy process technique for order of preference by similarity to ideal solution, preference ranking organization method for enrichment evaluation, and multiattribute utility theory methods. This comparative study is aimed at screening and ranking the three selected NFC options against five aspects: sustainability, environmental friendliness, economics, proliferation resistance, and technical feasibility. The selected fuel cycle options include pressurized water reactor (PWR) once-through cycle, PWR mixed oxide cycle, or pyroprocessing sodium-cooled fast reactor cycle. A sensitivity analysis was performed to prove the robustness of the results and explore the influence of criteria on the obtained ranking. As a result of the comparative analysis, the pyroprocessing sodium-cooled fast reactor cycle is determined to be the most competitive option among the NFC scenarios.

• 대한인간공학회:학술대회논문집
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• 대한인간공학회 1993년도 추계학술대회논문집
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• pp.21-30
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• 1993

This paper describes the development of a simulation model of nuclear power plant operators including cognitive aspects by using a network modeling soft ware, Micro-SAINT (System Analysis of Integrated Networks of Tasks) for the analysis of operator performance. Network model description based on Micro-SAINT includes tasks, resources, precedence relations among tasks, flow of information and PSFs (Performance Shaping Factors) on task performance. We have tried to evaluate the performance with several performance measures such as the number of tasks allocated, relative time presure among operators within a shift, for the selected test accident scenarior; small-break LOCA (Loss of Coolant Accident) in a PWR (Pressurized Water Reactor) type nuclear power plant.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1339-1345
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• 2017

In the framework of the EU funded project NURESAFE, the subchannel code CTF and the neutronics code DYN3D were integrated and coupled on the NURESIM platform. The developments achieved during this 3-year project include assembly-level and pin-by-pin multiphysics thermal hydraulics/neutron kinetics coupling. In order to test this coupling, a PWR rod ejection transient was simulated on a MOX/UOX minicore. The transient is simulated using two different models of the minicore. In the first simulation, both codes model the core with an assembly-wise resolution. In the second simulation, a pin-by-pin fuel-centered model is used in CTF for the central assembly, and a pin power reconstruction method is applied in DYN3D. The analysis shows the influence of the different models on global parameters, such as the power and the average fuel temperature, but also on local parameters such as the maximum fuel temperature.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 춘계 학술대회
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• pp.426-427
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• 2005

Advanced Spent Fuel Conditioning Process (ACP) is a pyrochemical process in which the spent fuel of PWR is transformed into the uranic metal ingot. Through this process, which has been developed in KAERI since 1998, the radioactivity, the radiotoxicity, the heat and the volume of the PWR spent fuel are reduced by a quarter of the original. To demonstrate a lab-scale process and extract the data for the later pilot-scale process, a demonstration facility of ACP (ACPF) is under construction and the lab-scale demonstration is slated for 2006. To establish the safeguardability of ACPF, a safeguards system including a neutron counter based on non-destructive assay, which is named as ACP Safeguards Neutron Counter (ASNC), the ACP Safeguards Surveillance System (ASSS) which consists of two neutron monitors and five IAEA cameras, and Laser Induced Breakdown System (LIBS) have been developed and are ready to be installed at ACPF. The target materials of ACP to assay with ASNC are categorized into three types among which the first is the uranic metal ingot, the second is the salt waste and the last is $UO_2$ and $U_{3}O_8$ powders, rod cuts and hulls. The Pu content of process nuclear materials can be accounted with ASNC. The ASSS is integrated in the ACP Intelligent Surveillance Software (AISS) in which the IAEA camera images and background signals at the rear doors of ACPF are displayed. The composition of special nuclear materials of ACP can be measured with LIBS which can be a supporting measurement tool for ASNC. The conceptual picture of safeguards system of ACPF is shown in Fig. 1.

• Nuclear Engineering and Technology
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• 제36권4호
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• pp.346-356
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• 2004

14 Pressurized Water Reactors (PWR) in Korea use a remote monitoring system (RMS), which have been used in Korea since 1998. A Memorandum of Understanding on Remote Monitoring, based on Enhanced Cooperation on PWRs, was signed at the 10th Safeguards Review Meeting in October 2001 between the International Atomic Energy Agency (IAEA) and Ministry Of Science and Technology (MOST). Thereafter, all PWR power plants applied for remote monitoring systems. However, the existing method is high cost (involving expensive telephone costs). So, it was eventually applied to an Internet system for Remote Monitoring. According to the Internet-based Virtual Private Network (VPN) applied to Remote Monitoring, the Korea Atomic Energy Research Institute (KAERI) came to an agreement with the IAEA, using a Member State Support Program (MSSP). Phase I is a Lab test. Phase II is to apply it to a target power plant. Phase III is to apply it to all the power plants. This paper reports on the penetration testing of Phase I. Phase I involved both domestic testing and international testing. The target of the testing consisted of a Surveillance Digital Integrated System (SDIS) Server, IAEA Server and TCNC (Technology Center for Nuclear Control) Server. In each system, Virtual Private Network (VPN) system hardware was installed. The penetration of the three systems and the three VPNs was tested. The domestic test involved two hacking scenarios: hacking from the outside and hacking from the inside. The international test involved one scenario from the outside. The results of tests demonstrated that the VPN hardware provided a good defense against hacking. We verified that there was no invasion of the system (SDIS Server and VPN; TCNC Server and VPN; and IAEA Server and VPN) via penetration testing.