• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1924-1930
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• 2021

By using miniature SENB specimens, the fracture properties of the materials in the region of welded metal, 321 stainless steel heat affected zone, 690 alloy heat affected zone of 321/690 dissimilar metal girth welded joints were tested. Both the J-resistance curves and critical fracture toughness of the three different materials are affected by the crack size because of the effect of crack tip constraint. Groups of constraint corrected J-resistance curves of the three materials are obtained according to J-Q-M approach. The welded metals exhibit the best fracture resistance but the worst fracture resistance is observed in the material of 690 alloy heat affected zone.

• Nuclear Engineering and Technology
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• 제26권2호
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• pp.179-189
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• 1994

이상유동시 원자로 냉각재 펌프의 성능을 펌프의 기하학적 형상 및 단상 유동시의 펌프 성능을 이용하여 예측하였다. 단상 유동시의 원자로 냉각재 펌프의 벽면 마찰손실은 Truckenbrodt의 경계층 이론을 이용하여 예측하였으며, 계산된 벽면 마찰 손실 및 분리 손실을 사용하여 이상유동시의 수두손실을 예측하였다. 해석결과는 Combustion Engineering 사의 펌프 실험 데이터와 비교하였다. 또한 냉각재 상실사고시 이상유동배수가 첨두 피복재 온도에 미치는 영향을 RELAP5를 사용하여 평가하였으며, 분석결과는 이상유동배수의 정확성이 중요한 영향을 미치는 것으로 나타났다.

• 한국안전학회지
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• 제32권1호
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• pp.140-144
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• 2017

Fire protection for nuclear power plants should be designed according to the concept of "Defense in Depth" to achieve the reactor safety shutdown. This concept focuses on fire prevention, fire suppression and safe shutdown. Fire prevention is the first line of "Defense in Depth" and the licensee should establish administrative measures to minimize the potential for fire to occur. Administrative measures should include procedures to control handling and use of combustibles. Electrical cables is the major contributor of fire loads in nuclear power plants, therefore electrical cables should be fire retardant. Electrical cables installed in nuclear power plants should pass the flame test in IEEE-383 standard in accordance with NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants". To assure the fire retardant of electrical cables during design life, both aged and unaged cable specimens should be tested in accordance with IEEE-383. It can be generally thought that the flammability of electrical cables has been increased by wearing as time passed, however the results from fire retardant tests performed in U.S.A and Korea indicate the inconsistent tendency of aging and consequential decrease in flammability. In this study, it is expected that the effective methodology for validation of fire retardant performance would be identified through the review of the results from fire retardant tests.

• 대한기계학회논문집B
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• 제36권3호
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• pp.363-369
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• 2012

CANDU형 원전의 칼란드리아 내부 감속재 열유동은 입구 노즐에서 나온 제트 유동에 의해 발생하는 관성력과 감속재로 전달되는 열부하에 의해 발생하는 부력의 상호작용으로 인해 복잡한 난류 특성을 나타낸다. 본 연구에서는 이러한 복잡한 감속재 열유동을 정확히 예측하기 위한 난류모델의 성능을 평가하기 위해 상용 유동해석 프로그램인 FLUENT에 탑재된 난류모델들을 사용해서 계산한 결과를 Sheridan Park Engineering Laboratory (SPEL)의 실험값과 비교하였다. 결론적으로 CANDU형 원전의 칼란드리아 내부 감속재 열유동을 신뢰할 수 있게 예측하기 위해서는 부력이 난류 유동에 미치는 영향을 고려해주는 생성항을 포함한 난류 모델이 사용되어야 한다.

• 한국유체기계학회 논문집
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• 제17권6호
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• pp.104-108
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• 2014

The recent Fukushima nuclear power plant accidents shows that the core make up at high RCS pressure condition is very important to prevent core melting. The core make up flow at high pressure condition should be driven by gravity force or passive forces because the AC-powered safety features are not available during a Station Black Out (SBO) accident. The reactor Coolant System (RCS) mass inventory is continuously decreased by releasing steam through the pressurizer safety valves after reactor trip during a SBO accident. The core will be melted down within 2~3 hours without core make up action by active or passive mode. In the new design concept of a Hybrid Safety Injection Tank (Hybrid SIT) both for low and high RCS pressure conditions, the low pressure nitrogen gas serves as a charging pressure for a LBLOCA injection mode, while the PZR high pressure steam provides an equalizing pressure for a high pressure injection mode such as a SBO accident. After the pressure equalizing process by battery driven initiation valve at a high pressure SBO condition, the Hybrid SIT injection water will be passively injected into the reactor downcomer by gravity head. The SBO simulation by MARS code show that the core makeup injection flow through the Hybrid SIT continued up to the SIT empty condition, and the core heatup is delayed as much.

• Nuclear Engineering and Technology
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• 제41권1호
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• pp.53-62
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• 2009

A set of experiments has been conducted on the performance sensitivity of the passive residual heat removal system (PRHRS) for an advanced integral type reactor, SMART, by using a high temperature and high pressure thermal-hydraulic test facility, the VISTA facility. In this paper the effects of the opening delay of the PRHRS bypass valves and the closing delay of the secondary system isolation valves, and the initial water level and the initial pressure of the compensating tank (CT) are investigated. During the reference test a stable flow occurs in a natural circulation loop that is composed of a steam generator secondary side, a secondary system, and a PRHRS; this is ascertained by a repetition test. When the PRHRS bypass valves are operated 10 seconds later than the secondary system isolation valves, the primary system is not properly cooled. When the secondary system isolation valves are operated 10 or 30 seconds later than the PRHRS bypass valves, the primary system is effectively cooled but the inventory of the PRHRS CT is drained earlier. As the initial water level of the CT is lowered to 16% of the full water level, the water is quickly drained and then nitrogen gas is introduced into the PRHRS, resulting in the deterioration of the PRHRS performance. When the initial pressure of the PRHRS is at 0.1MPa, the natural circulation is not performed properly. When the initial pressures of the PRHRS are 2.5 or 3.5 MPa, they show better performance than did the reference test.

• Journal of Radiation Protection and Research
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• 제41권4호
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• pp.368-372
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• 2016

Background: Methodologies for a series of radiological consequence assessments show a distinctive difference according to the design principles of the original nuclear suppliers and their technical standards to be imposed. This is due to the uncertainties of the accidental source term, radionuclide behavior in the environment, and subsequent radiological dose. Both types of PWR and PHWR are operated in Korea. However, technical standards for evaluating atmospheric dispersion have been enacted based on the U.S. NRC's positions regardless of the reactor types. For this reason, it might cause a controversy between the licensor and licensee of a nuclear power plant. Materials and Methods: It was modelled under the framework of the NRC Regulatory Guide 1.145 for light-water reactors, reflecting the features of heavy-water reactors as specified in the Canadian National Standard and the modelling features in MACCS2, such as atmospheric diffusion coefficient, ground deposition, surface roughness, radioactive plume depletion, and exposure from ground deposition. Results and Discussion: An integrated accident consequence assessment code, ACCESS (Accident Consequence Assessment Code for Evaluating Site Suitability), was developed by taking into account the unique regulatory positions for reactor types under the framework of the current Korean technical standards. Field tracer experiments and hand calculations have been carried out for validation and verification of the models. Conclusion: The modelling approaches of ACCESS and its features are introduced, and its applicative results for a hypothetical accidental scenario are comprehensively discussed. In an applicative study, the predicted results by the light-water reactor assessment model were higher than those by other models in terms of total doses.

• Nuclear Engineering and Technology
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• 제42권4호
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• pp.339-364
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• 2010

Following a joint OECD/NEA-IAEA-sponsored meeting to define the current role and future perspectives of the application of Computational Fluid Dynamics (CFD) to nuclear reactor safety problems, three Writing Groups were created, under the auspices of the NEA working group WGAMA, to produce state-of-the-art reports on different aspects of the subject. The work of the second group, WG2, was to document the existing assessment databases for CFD simulation in the context of Nuclear Reactor Safety (NRS) analysis, to gain a measure of the degree of quality and trust in CFD as a numerical analysis tool, and to take initiatives to extend the existing databases. The group worked over the period of 2003-2007 and produced a final state-of-the-art report. The present paper summarises the material gathered during the study, illustrating the points with a few highlights. A total of 22 safety issues were identified for which the application of CFD was considered to potentially bring real benefits in terms of better understanding and increased safety. A list of the existing databases was drawn up and synthesised, both from the nuclear area and from other parallel, non-nuclear, industrial activities. The gaps in the technology base were also identified and discussed. In order to initiate new ways of bringing experimentalists and numerical analysts together, an international workshop -- CFD4NRS (the first in a series) -- was organised, a new blind benchmark activity was set up based on turbulent mixing in T-junctions, and a Wiki-type web portal was created to offer online access to the material put together by the group giving the reader the opportunity to update and extend the contents to keep the information source topical and dynamic.

• 대한산업공학회지
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• 제28권2호
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• pp.178-186
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• 2002

The objective of this study is to evaluate impact of computer-based man-machine interfaces of Korea Next Generation Reactor (KNGR) on the operator mental workload. Empirical experiments were conducted to measure the operator mental workloads of KNGR and Yong-Gwang Unit 3 and 4, respectively. A comparison analysis based on a NASA TLX revealed that Yong-Gwang Unit 3 and 4 were superior to KNGR in terms of the mental workload. Post-hoc analyses showed that the mental workload of senior reactor operators was significantly higher than those of reactor and turbine operators, regardless of plant types. The implications of the findings were discussed in detail.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2001년도 ICCAS
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• pp.133.6-133
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• 2001

Reactor vessel is one of the most important equipment of Nuclear Power Plant (NPP) with regard to the nuclear safety. Thus reactor vessel must be examined periodically by certified experts. Currently, ultra-sonic(UT) non-destructive inspection is executed on reactor vessel. Two different techniques are used in this inspection. One is using the movable manipulator fixed with the support-guide placed on the vessel, and the other is using mobile robot moving in the vessel. Movable manipulator machine is very heavy, hard to handle, and very expensive. Mobile robot equipment is small and convenient but has a weak point on positional precision. To solve these problems we developed a reactor inspection system based on laser-driven mobile robot. This paper describes the main concept and structure of integrated inspection units and the feature of implemented units.