• 대한전기학회:학술대회논문집
• /
• 대한전기학회 2007년도 심포지엄 논문집 정보 및 제어부문
• /
• pp.215-216
• /
• 2007

This paper presents a real-time scheduler for nuclear safety system. According to constraints and requirements of nuclear safety system, scheduler design analysis is done and algorithms are developed for implementation. Using DSP based hardware, a real-time scheduler is realized. Consequently, this paper shows the performance of periodical software through the monitoring program.

• 방사성폐기물학회지
• /
• 제4권4호
• /
• pp.401-409
• /
• 2006

우리나라를 포함한 일본 및 미국과 독일, 영국, 프랑스 등 유럽국가의 원자력발전소 해체에 관한 규제제도 조사를 수행하였다. 각국의 해체에 관한 규제제도에 관하여 규제정책, 법규, 인허가 절차, 검사, 대중참여 등의 항목별로 비교분석을 수행하였다. 향후 본 조사결과는 국내의 가동 원자력발전소의 폐로 및 해체에 대비한 국내 제도 개선방향 수립에 참조자료로서 활용될 예정이다.

• Nuclear Engineering and Technology
• /
• 제56권7호
• /
• pp.2534-2544
• /
• 2024

This paper proposes a flooding safety system (FSS) and its operation strategy that can provide long-term safety and effective maintenance for modules of small modular reactor (SMR) and metal containment maintained at dried environment during normal operation. During hypothesized accidents, the FSS re-collects the evaporated steam into the common pool by the condenser installed above the common water pool and provides an emergency coolant for the cavities and auxiliary pools. This study suggested that the condensate re-collection strategy using the FSS can effectively delay the depletion of available water in response to the accidents. Without recollection, the achievable grace periods ranged from 44 to 1507 days for six-module and one-module accidents, respectively. However, with a full re-collection (ratio = 1.0), the time to total depletion of emergency coolant was estimated indefinite. Even with a partial re-collection ratio of 0.3, a grace period of 83.5 days could be ensured for a six-module transient. This study reported the effectiveness of condensate re-collection and the FSS as an innovative safety management strategy and system. Employing a condensate re-collection strategy with a high re-collection ratio can enhance the long-term safety and effective convenience of SMR operations and maintenance.

• 디지털산업정보학회논문지
• /
• 제16권1호
• /
• pp.99-109
• /
• 2020

In the past, conservative concepts have been applied in terms of the characteristic of nuclear power plants(NPPs), resulting in analog-based equipment and closed networks. However, as digital technology has recently been applied to the design, digital-based facilities and communication networks have been used in nuclear power plants, increasing the risk of cybersecurity than using analog-based facilities. Nuclear power plant facilities are divided into a safety system and a non-safety system. It is essential to identify the difference and cope with cybersecurity threats to the safety system according to its characteristics. In this paper, we examine the cybersecurity regulatory guidelines for safety systems in nuclear power plant facilities. Also, we analyze cybersecurity threats to a programmable logic controller of the safety system and suggest cybersecurity requirements be applied to it to respond to the threats. By implementing security functions suitable for the programmable logic controller according to the suggested cybersecurity requirements, regulatory guidelines can be satisfied, and security functions can be extended according to other system requirements. Also, it can effectively cope with cybersecurity attacks that may occur during the operation of nuclear power plants.

• 신재생에너지
• /
• 제16권4호
• /
• pp.49-64
• /
• 2020

Photovoltaic and Thermal collector (PV/T) systems are renewable energy devices that can produce electricity and heat energy simultaneously using solar panels and heat exchangers. Since PV/T systems are exposed to the outdoors, their reliability is affected by various environmental factors. This paper presents a reliability test for a PV/T system and evaluates the test results. The reliability assessment entails performance, environment, safety, and life tests. The factor that had the greatest influence on the life of the system was the hydraulic pressure applied to the heat exchanger. A test was conducted by repeatedly applying pressure to the PV/T system, and a reliability analysis was conducted based on the test results. As a result, the shape parameter (β) value of 5.6658 and the B10life 308,577 cycles at the lower 95% confidence interval were obtained.

• Nuclear Engineering and Technology
• /
• 제49권2호
• /
• pp.295-305
• /
• 2017

This paper reviews current status of the unified approach known as integrated safety assessment (ISA), as well as the associated SCAIS (simulation codes system for ISA) computer platform. These constitute a proposal, which is the result of collaborative action among the Nuclear Safety Council (CSN), University of Madrid (UPM), and NFQ Solutions S.L, aiming to allow independent regulatory verification of industry quantitative risk assessments. The content elaborates on discussions of the classical treatment of time in conventional probabilistic safety assessment (PSA) sequences and states important conclusions that can be used to avoid systematic and unacceptable underestimation of the failure exceedance frequencies. The unified ISA method meets this challenge by coupling deterministic and probabilistic mutual influences. The feasibility of the approach is illustrated with some examples of its application to a real size plant.

• International Journal of Safety
• /
• 제3권1호
• /
• pp.38-46
• /
• 2004

This paper introduces the software life-cycle V&V (verification and validation) tasks for the KNICS (Korea nuclear instrumentation and control system) project. The objectives of the V&V tasks are mainly to develop a programmable logic controller (PLC) for safety critical instrumentation and control (I&C) systems, and then to apply the PLC to developing the prototype of an engineered safety features-component control system (ESF-CCS) in nuclear power plants. As preparative works for the software V&V, various kinds of software plans and V&V task procedures have been developed according to the software life-cycle management. A number of software V&V tools have been adopted or developed to efficiently support the V&V tasks. The V&V techniques employed in this work include a checklist-based review and inspection, a requirement traceability analysis, formal verification, and life-cycle based software testing.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 추계학술발표회
• /
• pp.1047-1048
• /
• 2005

• Nuclear Engineering and Technology
• /
• 제55권6호
• /
• pp.2047-2052
• /
• 2023

Risk-informed approach has been widely applied in the safety design, regulation, and operation of nuclear reactors. It has been commonly accepted that risk-informed design optimization should be used in the innovative reactor designs to make nuclear system highly safe and reliable. In spite of the risk-informed approach has been used in some advanced nuclear reactors designs, such as Westinghouse IRIS, Gen-IV sodium fast reactors and lead-based fast reactors, the process of risk-informed design of nuclear reactors is hardly to carry out when passive system reliability should be integrated in the framework. A practical method for new passive safety reactors based on probabilistic safety assessment (PSA) and passive system reliability analyze linking is proposed in this paper. New three-dimension frequency-consequence curve based on risk concept with three variables is used in this method. The proposed method has been applied to the determination optimization of design options selection in a 10 MWth lead-based research reactor(LR) to obtain one optimized system design in conceptual design stage, using the integrated reliability and probabilistic safety assessment program RiskA, and the computation resources and time consumption in this process was demonstrated reasonable and acceptable.

• Nuclear Engineering and Technology
• /
• 제41권6호
• /
• pp.849-858
• /
• 2009

Risk caused by safety-critical instrumentation and control (I&C) systems considerably affects overall plant risk. As digitalization of safety-critical systems in nuclear power plants progresses, a risk model of a digitalized safety system is required and must be included in a plant safety model in order to assess this risk effect on the plant. Unique features of a digital system cause some challenges in risk modeling. This article aims at providing an overview of the issues related to the development of a static fault-tree-based risk model. We categorize the complicated issues of digital system probabilistic risk assessment (PRA) into four groups based on their characteristics: hardware module issues, software issues, system issues, and safety function issues. Quantification of the effect of these issues dominates the quality of a developed risk model. Recent research activities for addressing various issues, such as the modeling framework of a software-based system, the software failure probability and the fault coverage of a self monitoring mechanism, are discussed. Although these issues are interrelated and affect each other, the categorized and systematic approach suggested here will provide a proper insight for analyzing risk from a digital system.