• 시스템엔지니어링학술지
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• 제12권2호
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• pp.91-99
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• 2016

The safety of nuclear power plants has received much attention; this safety largely depends on the continuous availability of electrical energy source during all modes of nuclear power plant operation. A station blackout (SBO) describes the loss of the off-site electric power, the failure of the emergency diesel generators, and the unavailability of the alternate AC (AAC) power. Consequently, all systems that are AC powered such as the safety injection, shutdown cooling, component cooling water, and essential service water systems are unavailable. The aim of this study is to investigate the deficiencies of the existing alternatives for coping with an extended SBO for APR1400 design. The method is analyzing the existing deficiencies and proposing an optimal solution for the NPP design during the extended SBO. This study, established a new passive system, called passive decay heat removal system (PDHRS), using systems engineering approach.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.764-775
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• 2020

Periodic safety reviews (PSRs) are conducted on operating nuclear power plants (NPPs) and have been mandated also for research reactors in Korea, in response to the Fukushima accident. One safety review tool, the probabilistic safety assessment (PSA), aims to identify weaknesses in the design and operation of the research reactor, and to evaluate and compare possible safety improvements. However, the PSA for research reactors is difficult due to scarce data availability. An important element in the analysis of research reactors is the reactor protection system (RPS), with its functionality and importance. In this view, we consider that of the AGN-201K, a zero-power reactor without forced decay heat removal systems, to demonstrate a risk-informed safety improvement study. By incorporating risk- and safety-significance importance measures, and sensitivity and uncertainty analyses, the proposed method identifies critical components in the RPS reliability model, systematically proposes potential safety improvements and ranks them to assist in the decision-making process.

• 한국초전도ㆍ저온공학회논문지
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• 제3권2호
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• pp.15-20
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• 2001

This paper describes a model flux pump experiment recently performed at the MIT Francis Bitter Magnet Laboratory. The results of the model flux pump will be used in the development of a prototype flux pump that will be couple to a high-temperature superconductor (HTS) insert coil of a high-field NMR (Nuclear Magnetic Resonance) magnet, Such an HTS insert is unlikely to operate in persistent model because of the conductors low index(n) The flux pump can compensate fro field decay in the HTS insert coil and make the insert operate effectively in persistent mode . The flux pump, comprised essentially of a transformer an two switches. all made of superconductor, transfers into the insert coil a fraction of a magnetic energy that is first introduced in the secondary circuit of the transformer by a current supplied to the primary circuit. A model flux pump has been designed. fabricated, and operated to demonstrate that a flux pump can indeed supply a small metered current into a load superconducting magnet. A current increment in the range of microamperes has been measured in the magnet after each pumping action. The superconducting model flux pump is made of Nb$_3$ Sn tape, The pump is placed in a gaseous environment above the liquid helium level to keep its heat dissipation from directly discharged in the liquid: the effluent helium vapor maintains the thermal stability of the flux pump.

• 대한기계학회논문집B
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• 제31권5호
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• pp.482-490
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• 2007

When a molten corium is relocated in a lower head of a reactor vessel, the ERVC (External Reactor Vessel Cooling) system is actuated as coolant is supplied into a reactor cavity to remove a decay heat from the molten corium during a severe accident. To achieve this severe accident mitigation strategy, the two-phase natural circulation flow in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. For this reason, one-dimensional natural circulation flow tests and the simple analysis were conducted to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The calculated circulation flow rate was similar to experimental ones within about ${\pm}$15% error bounds and depended on the form loss due to the inlet/outlet area.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.358-362
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• 2003

산화물 형태의 사용후핵연료를 용융염에서 금속 형태로 전환하여, 발열량, 부피 및 방사능을 1/4로 감소시킬 수 있는 전기화학적 금속전환 공정을 개발하고, 5kg $U_3O_8$/Batch 규모의 mock-up 실험을 수행하였다. 본 연구에서는 전해 셀의 운전변수를 해석하였으며, 아울러 hot test를 위한 장치개발 연구도 병행하였다. 전기화학적 금속전환 공정을 이용하여 $U_3O_8$ 형태의 천연우라늄 분말을 99% 이상 금속전환할 수 있었으며, 또한 20kg $U_3O_8$/batch 규모 장치의 설계자료를 산출할 수 있었다.

• Nuclear Engineering and Technology
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• 제33권2호
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• pp.209-224
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• 2001

The SSC-K system analysis code is under development at the Korea Atomic Energy Research Institute (KAERI) as a part of the KALIMER project. The SSC-K code is being used as the principal tool for analyzing a variety of off-normal conditions or accidents of the preliminary KALIMER design. The SSC-K code features a multiple-channel core representation coupled with a point kinetics model with reactivity feedback. It provides a detailed, one-dimensional thermal-hydraulic simulation of the primary and secondary sodium coolant circuits, as well as the balance-of-plant steam/water circuit. Recently a two-dimensional hot pool model was incorporated into SSC-K for analysis of thermal stratification phenomena in the hot pool. In addition, SSC-K contains detailed models for the passive decay heat removal system and a generalized plant control system. The SSC-K code has also been applied to the computational engine for an interactive simulation of the KALIMER plant. This paper presents an overview of the recent activities concerned with SSC-K code model development This paper focuses on both descriptions of the newly adopted thermal hydraulic and neutronic models, and applications to KALIMER analyses for typical anticipated transients without scram.

• Nuclear Engineering and Technology
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• 제24권3호
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• pp.311-318
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• 1992

원자력 발전소에서 Mid-loop 운전시 배관내에서 발생하는 자유수면 와동으로 인해 잔열 제거계통 배관내 공기가 흡입될 가능성이 있으며 이로 인한 계통상실 방지를 위하여 수위와 흡입유량과의 관계를 실험을 통해서 H/d, 프라우드수, 레이놀즈 수 등과 같은 무차원 수로 구하였다. 실험결과 레이놀즈수는 크게 영향을 미치지 않았으며 주로 프라우드수가 자유수면 와동을 지배하는 것으로 판명되었다. 한편 운전시 펌프나 밸브의 개폐로 인한 수면의 섭동이 와동에 많은 영향을 미치는 것이 밝혀졌다. 원자력 발전소의 안전과 관련하여 배관내에서 와동으로 인한 공기흡입 방지책으로 Reducer형의 흡입구 개선방안을 제시하였다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.3175-3180
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• 2008

A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.179-190
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• 2013

A station blackout experiment called SBO-01 was performed at the ATLAS facility. From the SBO-01 test, the station blackout scenario can be characterized into two typical phases: A first phase characterized by decay heat removal through secondary safety valves until the SG dryouts, and a second phase characterized by an energy release through a blowdown of the primary system after the SG dryouts. During the second phase, some physical phenomena of the change over a pressurizer function, i.e., the pressurizer being full before the POSRV $1^{st}$ opening and then its function being taken by the RV, and the termination of normal natural circulation flow were identified. Finally, a core heatup occurred at a low core water level, although under a significant amount of PZR inventory, whose drainage seemed to be hindered owing to the pressurizer function by the RV. The transient of SBO-01 is well reproduced in the calculation using the MARS code.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2274-2284
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• 2020

The Nuclear Fuel Cycle Code "ANICCA" has been developed by SCK•CEN to answer particular questions about the Belgian nuclear fleet. However, the wide range of capabilities of the code make it also useful for international or regional studies that include advanced technologies and strategies of cycle. This paper shows the main features of the code and the facilities that can be simulated. Additionally, a comparison between several codes and ANICCA has also been made to verify the performance of the code by means of a simulation proposed in the last NEA (OECD) Benchmark Study. Finally, a case study of the Belgian nuclear fuel cycle phase out has been carried out to show the possible impact of the transmutation of the minor actinides on the nuclear waste by the use of an Accelerator Driven System also known as ADS. Results show that ANICCA accomplishes its main purpose of simulating the scenarios giving similar outcomes to other codes. Regarding the case study, results show a reduction of more than 60% of minor actinides in the Belgian nuclear cycle when using an ADS, reducing significantly the radiotoxicity and decay heat of the high-level waste and facilitating its management.