• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2775-2787
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• 2021

At present, most of the widely used system codes for nuclear safety analysis are one-dimensional, which cannot effectively simulate the flow field of the reactor core or other structures. This is true even for the system codes containing three-dimensional modules with limited three-dimensional simulation function such as RELAP-3D. In contrast, the computational fluid dynamics (CFD) codes excel at providing a detailed three-dimensional flow field of the reactor core or other components; however, the computational domain is relatively small and results in the very high computing resource consuming. Therefore, the development of coupling codes, which can make comprehensive use of the advantages of system and CFD codes, has become a research focus. In this paper, a review focus on the researches of coupled CFD and thermal-hydraulic system codes was carried out, which summarized the method of coupling, the data transfer processing between CFD and system codes, and the verification and validation (V&V) of coupled codes. Furthermore, a series of problems associated with the coupling procedure have been identified, which provide the general direction for the development and V&V efforts of coupled codes.

• Nuclear Engineering and Technology
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• 제26권1호
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• pp.9-18
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• 1994

월성 원자력발전소의 안전계통은 비상사태시에만 작동하는 3분의 2논리로 구성되어 있다. 그들의 작동성을 보증하기 위해 이 안전계통은 주기적으로 점검되어진다. 본연구에서 사람의 실수가 고려되어진 3분의 2논리 구성 시스템에서의 불이용도가 계산되어졌다. 그리고 우리는 시험기간중에 사람의 실수또는 기계의 고장으로 인해 발전정지를 일으킬 확률을 구했다. 우리는 이 불이용도와 발전정지를 일으킬 확률을 둘다 고려하여 적정한 최적점검주기를 계산하였다. 이렇게 얻어진 점검주기와 현재 사용되는 점검주기를 비교하면 사람의 실수를 최소(8.24 $\times$ $10^{-6}$ )로 보았을때 최적점검주기는 현재 사용되는 점검주기 보다 조금 짧았고 사람의 실수를 최대 (4.44 $\times$ $10^{-4}$ )로 보았을 때 최적점검주기는 현재 사용하는 점검 주기보다 다소 긴 것으로 계산되어졌다.

• 해양환경안전학회지
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• 제12권4호
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• pp.253-259
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• 2006

본 논문은 가장 극한 조건(공정모델-2)에서 화학 제염한 경우 원자로 냉각재 펌프용 스테인리스강의 내식성 평가에 관하여 연구하였다. 사이클 경과에 따른 304 스테인리스강의 전기화학적 특성은 다른 스테인리스 강보다 우수한 특성을 나타냈다. 또한 공정모델-1과 공정모델-2의 304 스테인리스강은 가장 낮은 무게 감량을 나타냈다. 공정모델 용액에서 304 스테인리스강, 415 스테인리스강, 431 스테인리스강에 대한 실험 결과 공정모델-1에 대한 공정모델-2의 무게감량비는 각각 2.908, 2.372,그리고 2.370배를 나타냈다. 그 이유는 공정모델-2의 경우가 공정모델-1에 비하여 화학약품 농도나 온도가 높은 가혹한 조건에 기인한 것으로 사료된다.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.605-612
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• 2013

Removal of decay heat from an operating reactor during a prolonged station blackout condition is a big concern for reactor designers, especially after the recent Fukushima accident. In the case of a prolonged station blackout condition, heat removal is possible only by passive means since no pumps or active systems are available. Keeping this in mind, the AHWR has been designed with many passive safety features. One of them is a passive means of removing decay heat with the help of Isolation Condensers (ICs) which are submerged in a big water pool called the Gravity Driven Water Pool (GDWP). The ICs have many tubes in which the steam, generated by the reactor core due to the decay heat, flows and condenses by rejecting the heat into the water pool. After condensation, the condensate falls back into the steam drum of the reactor. The GDWP tank holds a large amount of water, about 8000 $m^3$, which is located at a higher elevation than the steam drum of the reactor in order to promote natural circulation. Due to the recent Fukushima type accidents, it has been a concern to understand and evaluate the capability of the ICs to remove decay heat for a prolonged period without escalating fuel sheath temperature. In view of this, an analysis has been performed for decay heat removal characteristics over several days of an AHWR by ICs. The computer code RELAP5/MOD3.2 was used for this purpose. Results indicate that the ICs can remove the decay heat for more than 10 days without causing any bulk boiling in the GDWP. After that, decay heat can be removed for more than 40 days by boiling off the pool inventory. The pressure inside the containment does not exceed the design pressure even after 10 days by condensation of steam generated from the GDWP on the walls of containment and on the Passive Containment Cooling System (PCCS) tubes. If venting is carried out after this period, the decay heat can be removed for more than 50 days without exceeding the design limits.

• Nuclear Engineering and Technology
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• 제39권2호
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• pp.103-110
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• 2007

The High Temperature Gas-cooled Reactor (HTGR) possesses inherent safety features and is recognized as a representative advanced nuclear system for the future. Based on the success of the HTR-10, the long-time operation test and safety demonstration tests were carried out. The long-time operation test verifies that the operation procedure and control method are appropriate for the HTR-10 and the safety demonstration test shows that the HTR-10 possesses inherent safety features with a great margin. Meanwhile, two new projects have been recently launched to further develop HTGR technology. One is a prototype modular plant, denoted as HTR-PM, to demonstrate the commercial capability of the HTGR power plant. The HTR-PM is designed as $2{\times}250$ MWt, pebble bed core with a steam turbine generator that serves as an energy conversion system. The other is a gas turbine generator system coupled with the HTR-10, denoted as HTR-10GT, built to demonstrate the feasibility of the HTGR gas turbine technology. The gas turbine generator system is designed in a single shaft configuration supported by active magnetic bearings (AMB). The HTR-10GT project is now in the stage of engineering design and component fabrication. R&D on the helium turbocompressor, a key component, and the key technology of AMB are in progress.

• International Journal of Safety
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• 제1권1호
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• pp.58-61
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• 2002

A preparation method of cobalt supported alumina catalyst for a emergency escape mask cartridge has been studied. Catalysts were prepared by incipient wetness impregnation method using pre-shaped $\gamma$=alumina powders of 70-100 mesh. The catalyst was tested in a continuous-flow reactor system and characterized by elemental analysis, BET and TGA-DTA techniques. Cobalt shows higher activity than platinum or nickel for carbon monoxide oxidation at room temperature. Optimum loading amount of cobalt was 10 wt.% for CO oxidation and the reaction activity increases gradually with the increase of calcination temperature up to $450^{\circ}C.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.581-588
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• 2013

This paper deals with the Safety Analysis for $CANDU^{(R)}$ 6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermal-hydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR) analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermal-hydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY) aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermal-hydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

• 전기학회논문지
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• 제62권7호
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• pp.994-1002
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• 2013

This study suggests a generalized availability and safety evaluation model to evaluate the influences to the system's fault tolerant capabilities depending on automatic fault detection function such as the automatic periodic testings. The conventional evaluation model of automatic fault detection function deals only with the self diagnostics, and supposes that the fault detection coverage of self diagnostics is always constant. But all of the fault detection methods could be degraded. For example, the periodic surveillance test has the potential human errors or test equipment errors, the self diagnostics has the potential degradation of built-in logics, and the automatic periodic testing has the potential degradation of automatic test facilities. The suggested evaluation models have incorporated the loss or erroneous behaviors of the automatic fault detection methods. The availability and the safety of each module of the safety grade platform have been evaluated as they were applied the automatic periodic test methodology and the fault tolerant evaluation models. The availability and safety of the safety grade platform were improved when applied the automatic periodic testing. Especially the fault tolerant capability of the processor module with a weak self-diagnostics and the process parameter input modules were dramatically improved compared to the conventional cases. In addition, as a result of the safety evaluation of the digital reactor protection system, the system safety of the digital parts was improved about 4 times compared to the conventional cases.

• 한국안전학회지
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• 제35권4호
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• pp.84-91
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• 2020

The Korean nuclear industry developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code and this code adpots two-phase flows, two-fluid, three-field models which are comprised of gas, continuous liquid and droplet fields and has a capability to simulate three-dimensional model. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for accident management plan of nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification work for separate and integral effect experiments is required. In this reason, the goal of this work is to verify calculation capability of SPACE code for multiple failure accident. For this purpose, it was selected the experiment which was conducted to simulate a Multiple Steam Generator Tube Rupture(MSGTR) accident with Passive Auxiliary Feedwater System(PAFS) operation by Korea Atomic Energy Research Institute (KAERI) and focused that the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The MSGR accident has a unique feature of the penetration of the barrier between the Reactor Coolant System (RCS) and the secondary system resulting from multiple failure of steam generator U-tubes. The PAFS is one of the advanced safety features with passive cooling system to replace a conventional active auxiliary feedwater system. This system is passively capable of condensing steam generated in steam generator and feeding the condensed water to the steam generator by gravity. As the results of overall system transient response using SPACE code showed similar trends with the experimental results such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it could be concluded that the SPACE code has sufficient capability to simulate a MSGTR accident.

• 대한기계학회논문집
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• 제16권11호
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• pp.2126-2135
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• 1992

본 연구에서는 국내 원자력 발전소중 영광 3,4호기의 설계자료를 토대로 1/6 크기로 축소한 모델실험을 통해서 공기흡입이 발생하는 임계수위를 결정하는 상관식을 개발하였으며 또한 공기흡입구를 reducer type으로 개선함으로써 공기흡입을 방지할 수 있음을 밝혔다.