• Nuclear Engineering and Technology
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• 제44권6호
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• pp.597-610
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• 2012

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL ($\underline{P}$AF$\underline{S}$ $\underline{C}$ondensing Heat Removal $\underline{A}$ssessment $\underline{L}$oop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.

• 에너지공학
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• 제24권1호
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• pp.78-87
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• 2015

중대사고시 금속용융물층의 열속 집중 현상(Focusig effect)에 대해 상부와 측면벽의 냉각 조건과 높이를 변화시키면서 실험과 수치해석을 수행하였다. 상사성(Analogy) 원리를 이용해 열전달 실험 대신 물질전달 실험을 수행하였으며 황산-황산구리 수용액의 전기도금계를 물질전달계로 채택하였다. $Ra_H$ 수 $8.49{\times}10^7{\sim}5.43{\times}10^9$ 범위에서 상부와 측면벽의 냉각 조건을 세 가지로, 높이를 네 가지로 변화시키면서 열전달을 측정하였다. 상부만 냉각인 경우의 실험결과를 동일한 조건인 Rayleigh-Benard 자연대류 상관식과 비교한 바 Dropkin과 Somerscales, Globe와 Dropkin의 상관식과 매우 일치하였다. 측면벽만 냉각인 경우, 상부와 측면벽 모두 냉각인 경우, 상부만 냉각인 경우 순으로 열전달이 감소하였고, 냉각 조건을 고정한 상태에서 높이를 감소시킬수록 측면 열전달이 향상되었다.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2001년도 춘계학술대회 논문집
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• pp.594-597
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• 2001

As many industrial systems become more complex, they become extremely difficult to diagnose the cause of failures. The subject of this paper is ECCS(Emergency Core Cooling System) part of PWR(Pressurized Water Reactor). This paper presents modeling and diagnoser construction of ECCS based on discrete event system theory. Also, this paper presents that the ECCS system is diagnosible in our approach.

• 한국초전도ㆍ저온공학회논문지
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• 제5권1호
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• pp.48-51
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• 2003

A 1 MVA transformer with BSCCO-2223 high Tc superconducting (HTS) tapes was designed. The rated voltages of each sides of the transformer are 22.0 kV and 6.6 kV respectively. Double pancake HTS windings, which have advantages of insulations and distribution of high voltage, were adopted. Four HTS tapes were wound in parallel fer the windings of low voltage side. Each winding was composed of several double pancake windings made of four parallel conductors were transposed in order to distribute the currents equally in each conductor. A core of the transformer was designed as a shell type core made of laminated silicon steel plate and the core is separated with the windings by a cryostat with a room temperature bore. The operating temperature of HTS windings will be 65K with liquid nitrogen, and a cooling system using a cryocooler was proposed and designed conceptually. This HTS transformer is going to be manufactured in near future based on the design parameters presented in this paper.

• Journal of Mechanical Science and Technology
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• 제19권5호
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• pp.1182-1193
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• 2005

The capability of TRAC-M code to predict downcomer boiling effect during reflood phase in postulated PWR LOCA is evaluated using the results of downcomer effective water head and Cylindrical Core Test Facility (CCTF) experiments, which were performed at JAERI. With a full height downcomer simulator, effective water head experiment was carried out to investigate the applicability of the TRAC-M best estimate LOCA code to evaluate the effective water head with superheated wall temperature in downcomer. In order to clarify the effect of the initial superheat of the downcomer wall on the system and the core cooling behaviors during the reflood phase, two sets of analysis were also performed with a CCTF. Results show that TRAC­M code tends to under-predict downcomer effective water head and core differential pressure. However, the code results show a good agreement with the experimental results in downcomer temperature, heat flux and pressure. Finally, both experiment and calculation showed that the downcomer water head with the superheated downcomer wall is lower than that of the saturated wall temperature.

• 대한전기학회논문지:전기기기및에너지변환시스템부문B
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• 제52권1호
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• pp.23-26
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• 2003

This paper presents conceptual design and 3-D electromagnetic analysis of IMVA transformer with BSCCO-2223 High Tc Superconducting (HTS) tapes. The rated voltages of each sides of the transformer are 22.9 kV and 6.6 kV, and double pancake windings were adopted. High voltage and Low voltage sides were composed of several double pancake windings. Four HTS tapes were wound in parallel for the windings of low voltage side and were transposed in order to distribute the currents equally in each conductor The transformer core was designed as a shell type core made of laminated silicon steel plates and the core is separated with the windings by a cryostat with Fiberglass Reinforced Plastics(FRP). A sub-cooling system using L$N_2$ were designed to maintain the coolant temperature 65K. Finally perpendicular components of magnetic field applied to tapes were calculated 0.247 in the rated operation using 3-D analysis. A real 1MVA HTS transformer will be manufactured in near future based on the design parameters presented in this paper.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2803-2815
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• 2021

This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core operation. The uncertainty propagation is conducted using a two-step analysis code system comprising the lattice code STREAM, nodal code RAST-K, and spent nuclear fuel module SNF through the random sampling of microscopic cross-sections, fuel rod sizes, number densities, reactor core total power, and temperature distributions. Overall, the statistical analysis of the calculated samples demonstrates that the decay heat uncertainty decreases with the cooling time. The nuclear data and assembly design parameters are proven to be the largest contributors to the decay heat uncertainty, whereas the reactor core power and inlet coolant temperature have a minor effect. The majority of the decay heat uncertainties are delivered by a small number of isotopes such as ²⁴¹Am, ¹³⁷Ba, ²⁴⁴Cm, ²³⁸Pu, and ⁹⁰Y.

• 전력전자학회:학술대회논문집
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• 전력전자학회 2001년도 Proceedings ICPE 01 2001 International Conference on Power Electronics
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• pp.93-97
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• 2001

A new compact, low-energy electron beam irradiator has been developed. The core-loss of silicon steels can be reduced by magnetic-domain refinement method. The irradiator was developed for the application of core-loss reduction using the method. The beam energy of the irradiator can be varied from 35 to 80 keV and the maximum current is 3mA. The irradiation area is designed to be $30\times30mm2$ now and will be upgraded to $30\times150mm2$ using a scanning magnet and scanning cone. The electron beam generated from 3 mm diameter LaB6 is extracted to the air for the irradiation of the silicon steels in the air. A special irradiation port was developed for this low-energy irradiator. A havar foil with $4.08{\mu}m$ thickness were used for the window and a cold air-cooling system keeps the foil structure by removing heat at the window. The irradiator system and its operation characteristics will be discussed.

• Nuclear Engineering and Technology
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• 제21권4호
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• pp.259-266
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• 1989

LOFT L2-5 대형 냉각재상실사고 모사실험에 대해 RELAP5/MOD2/Cycle 36.04로부터 개선된 코드를 이용, 코드평가를 수행하였다. 강수관(Downcomer)모델 및 노심유로모델에 따른 코드 민감도분석을 위해 보충계산을 수행하였다. 계산결과는 1차계통의 압력, 파단부를 통과하는 질량유량, 노심 고온부의 피복재 온도등에 대해서 실험결과와 비교 분석되었다. 분석결과 RELAP5/MOD2 계산에 의해 1차계통의 수력학적 거동은 잘 묘사될 수 있으며, 단일 노심유로 모델을 이용한 계산에서는 사고발생이후 감압기간동안 노심이 과대 냉각되는 현상이 발견되었다. 노심의 고온유로에서의 수력학적 거동을 잘 묘사할 수 있는 이중 노심유로 모델계산을 이용하여 이 현상을 극복하고 실험치에 근사하는 결과를 얻을 수 있음을 알 수 있었다.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.33-44
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• 2006

This paper presents a review of small-to-medium-sized, pressurized-water-cooled nuclear power reactors whose major primary coolant systems are integrated into a reactor pressure vessel, the concepts categorized as Integrated Primary System Nuclear Reactors (IPSRs). Typical examples of these proposals of interest in this review are CAREM, SMART, IRIS and IMR, all of which are being aimed at the near term deployment. Emphasis is placed on thermal hydraulic aspects. A brief characterization of the IPSR concepts is made and comparisons of plant key parameters are shown. Discussions will follow for the core cooling under rated power conditions and natural circulation heat removal on the basis of the design data available in the public domain.