• 대한인간공학회지
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• 제30권1호
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• pp.55-64
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• 2011

Objective: The aim of this paper is to introduce the activities and research trends of human reliability analysis including brief summary about contents and methods of the analysis. Background: Various approaches and methods have been suggested and used to assess human reliability in field of risk assessment of nuclear power plants. However, it has noticed that there is high uncertainty in human reliability analysis which results in a major bottleneck for risk-informed activities of nuclear power plants. Method: First and second generation methods of human reliability analysis are reviewed and a few representative methods are discussed from the risk assessment perspective. The strength and weakness of each method is also examined from the viewpoint of reliability analyst as a user. In addition, new research trends in this field are briefly summarized. Results: Human reliability analysis has become an important tool to support not only risk assessment but also system design of a centralized complex system. Conclusion: Human reliability analysis should be improved by active cooperation with researchers in field of human factors. Application: The trends of human reliability analysis explained in this paper will help researchers to find interest topics to which they could contribute.

• 한국소음진동공학회논문집
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• 제27권2호
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• pp.148-154
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• 2017

This paper discussed design for resonance avoidance of sensor plates of loose-parts monitoring systems (LPMS) in nuclear power plants (NPP). An LPMS monitors impact of loose parts in primary loop of NPP by using accelerometers, which is mounted on sensor plates. Resonance of the plates may cause false alarms at frequencies over 10 kHz, which can be misunderstood as impact signals of loose parts with small mass and cause unnecessary response of NPP operators. Modal analysis was carried out for the existing sensor plate and design parameters affecting natural frequencies were chosen. Frequency response functions of plates were analyzed by changing the parameters and the optimized plate design for avoiding resonance was determined. Experiments was carried out for the plate specimen with improved design and verified the proposed approach and design.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1519-1523
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• 2021

Radioactive iodine (¹³¹I) released from nuclear power plants has been a critical environmental concern for workers. The effective trapping of radioactive iodine isotopes from the off-gas stream generated from nuclear facilities is an important issue in radioactive waste treatment systems evaluation. Numerous studies on retaining methyl iodide (CH₃I¹³¹) by impregnated activated carbons under the high content of moisture have been extensively studied so far. But there have been no good results on how to remove methyl iodide at high humid conditions up to now. A new challenge is to introduce other promising impregnating chemical agents that are able to uptake enough radioactive methyl iodide under high humid conditions. In order to develop a good removal efficiency to control radioiodine gas generated from a high humid process, activated carbons (ACs) impregnated with triethylene diamine (TEDA) and qinuclidine (QUID) were prepared. In addition, the removal efficiencies of the activated carbons (ACs) under humid conditions up to 95% RH were evaluated by applying the standard method specified in ASTM-D3808. Quinuclidine impregnated activated carbon showed a much higher decontamination factor above 1,000, which is enough to meet the regulation index for the iodine filters in nuclear power plants (NPPs).

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3571-3580
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• 2023

The application of the induction bending process to pipe systems in various industrial fields is increasing. Recently, efforts have also been made to apply this bending process to nuclear power plants because it can innovatively reduce welded parts of the curved pipes, such as elbows. However, there have been no cases of the application of induction bending to the piping of nuclear power plants. In this study, the applicability of the P91 induction bending piping for the sodium-cooled fast reactor PGSFR was validated through high temperature low cycle fatigue tests and creep tests using P91 induction bending pipe specimens. The tests confirmed that the materials sufficiently satisfied the fatigue life and the creep rupture life requirements for P91 steel at 550 ℃ in the ASME B&PV Code, Sec. III, Div. 5. The results show that the effects of heating and bending by the induction bending process on the material properties were not significant and the induction bending process could be applicable to piping system of PGSFR well.

• Journal of Radiation Protection and Research
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• 제20권4호
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• pp.285-296
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• 1995

각 원자력 발전소에서 발생되는 방사성폐기물을 한 곳에 모아 집중관리하기 위한 방사성폐기물 처분장의 부지선정이 국가적 과제로 부각되어 있으며, 충분한 검토를 거친 후 임해부지로 선정될 것이다. 이로 인하여 현재 각 원전부지내에 임시로 보관되어 있는 방사성폐기물에 대하여 전용선박에 의한 해상수송을 하여야 하면, 한국원자력연구소의 원자력환경관리센터(NEMAC)에서는 원전부지로부터 처 분장까지 안전하고 효율적으로 방사성폐기물을 수송할 수 있는 종합해상 수송체계를 개발중에 있다. 이 글은 해상수송체계가 갖추어야 할 수송선박의 기술기준을 설정하기 위한 것으로, 원자력 선진국의 진보된 방사성폐기물 해상수송기술에 관한 현황을 조사, 분석하고 국내의 제반여건을 고려하여 우리나라에서 사용될 수송선박의 설계 및 건조추진방향을 제시하였다. 따라서, 만일의 사고에도 방사성물질이 선박의 외부로 누출되지 않는 개념의 선박을 설계, 건조하여 방사성폐기물을 안전하게 해상수송하게 될 것이다.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.637-652
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• 2013

Instrumentation and control systems in nuclear power plants have been digitalized for the purpose of maintenance and precise operation. This digitalization, however, brings out issues related to cyber security. In the most recent past, international standard organizations, regulatory institutes, and research institutes have performed a number of studies addressing these systems cyber security.. In order to provide information helpful to the system designers in their application of cyber security for the systems, this paper presents methods and considerations to define attack vectors in a target system, to review and select the requirements in the Regulatory Guide 5.71, and to integrate the results to identify applicable technical security control requirements. In this study, attack vectors are analyzed through the vulnerability analyses and penetration tests with a simplified safety system, and the elements of critical digital assets acting as attack vectors are identified. Among the security control requirements listed in Appendices B and C to Regulatory Guide 5.71, those that should be implemented into the systems are selected and classified in groups of technical security control requirements using the results of the attack vector analysis. For the attack vector elements of critical digital assets, all the technical security control requirements are evaluated to determine whether they are applicable and effective, and considerations in this evaluation are also discussed. The technical security control requirements in three important categories of access control, monitoring and logging, and encryption are derived and grouped according to the elements of attack vectors as results for the sample safety system.

• Nuclear Engineering and Technology
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• 제7권2호
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• pp.159-165
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• 1975

The Nuclear power industry today faces a serious and rapidly emerging problem in reactor maintenance and occupational radiation exposure control. The basic problem is the need for much maintenance on nuclear power plants. The problem is seriously compounded by radiation exposure control requirements. Many studies are underway seeking solutions tut the industry is developing rapidly and new plants will not await the results of such studies. It is essential that attention be given to maintenance and exposure control in all phases of plant design, construction and operation.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 심포지엄 논문집 정보 및 제어부문
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• pp.75-77
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• 2005

An integrated I&C architecture for nuclear power plants is designed by the systems and devices being developed in a project. Its design reference is the APR1400 that was design certified in Korea. Digital equipment and several kinds of data communication networks (DCN) are used. To confirm the validity of DCN based architecture design, the traffic loads fur each network were calculated assuming the anticipated maximum traffic condition. The analysis showed that the utilizations of all networks satisfied the design requirements.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.244-250
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• 1996

Uncertainties in core neutronic calculations of CASMO-3/MASTER, which is a KAERI developed core nuclear design code system, were evaluated via comparisons with measured data. Comparisons were performed with plant measurement data from one Westinghouse type and one ABB-CE type plant and two Korean standard type plants. The CASMO-3/MASTER capability and levels of accuracy are concluded to be sufficient for the neutronics design including safety related parameters related with reactivity, power distributions, temperature and power coefficients, inverse boron worth and control bank worth.

• Nuclear Engineering and Technology
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• 제37권2호
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• pp.191-200
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• 2005

Shaking table tests of the seismic behavior of a steel frame structure model were performed. The purpose of these tests was to estimate the effects of a near-fault ground motion and a scenario earthquake based on a probabilistic seismic hazard analysis for nuclear power plant structures. Three representative kinds of earthquake ground motions were used for the input motions: the design earthquake ground motion for the Korean nuclear power plants, the scenario earthquakes for Korean nuclear power plant sites, and the near-fault earthquake record from the Chi-Chi earthquake. The probability-based scenario earthquakes were developed for the Korean nuclear power plant sites using the PSHA data. A 4-story steel frame structure was fabricated to perform the tests. Test results showed that the high frequency ground motions of the scenario earthquake did not damage the structure at the nuclear power plant site; however, the ground motions had a serious effect on the equipment installed on the high floors of the building. This shows that the design earthquake is not conservative enough to demonstrate the actual danger to safety related nuclear power plant equipment.