• Nuclear Engineering and Technology
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• 제34권5호
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• pp.421-432
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• 2002

As one of the advanced design features of the APR1400, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood phase of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is under progress. In this paper, test results of direct ECC bypass performed in the steam-water test facility tailed MIDAS (Multi-dimensional Investigation in Downcomer Annulus Simulation) are presented. The test condition is determined, based on the preliminary analysis of TRAC code, by applying the ‘modified linear scaling method’with the l/4.93 length scale . From the tests, ECC direct bypass fraction, steam condensation rate and information on the flow distribution in the upper annulus downcomer region are obtained.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.328-333
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• 1996

Boric acid concentrations of the refueling water storage tank (RWST) and the accumulators for Ulchin 1&2 (UCN 1&2) are increased to meet the post loss of coolant accident (post-LOCA) shutdown requirement for the extended fuel cycles from 12 months to 18 months. To maintain long term cooling (LTC) capability following a LOCA, the switchover tine is examined using BORON code to prevent the boron precipitation in the reactor core with the increased boron concentrations. The analysis results show that, at 8 hours after the initiation of LOCA. the emergency core noting system (ECCS) should be manually realigned to the simultaneous recirculation mode from the cold leg recirculation mode.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 2000년도 추계 학술발표회 논문집
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• pp.127-137
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• 2000

Accelerator Driven System (ADS) named HYPER(Hybrid Power Extraction Reactor) is being developed for the transmutation of nuclear waste in Korea Atomic Energy Research Institute(KAERI). The concept of the HYPER is using 1GeV proton to drive a subcritical core. HYPER system is believed to have much more stable dynamics than the critical system in terms of neutronics. However, the HYPER system is supposed to have some drawbacks for the cooling system accidents. Loss of Flow(LOF) and Loss of Heat Sink (LOHS) cause a strong damage. As results, those accidents would stop the power production in the critical system. On the other hand, the negative reactivity feedback could not stop the HYPER system because the HYPER is driven by an accelerator rather than reactivity.(omitted)

• 한국초전도ㆍ저온공학회논문지
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• 제5권3호
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• pp.34-37
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• 2003

A 1 MVA single-phase high temperature superconducting (HTS) transformer with BSCCO-2223 wire was designed in this paper. The rated voltages of each sides of the transformer are 22.9 kV and 6.6 kV respectively. Double pancake HTS windings arranged reciprocally will be used for the transformer windings, because of the advantages of insulation and distribution of surge voltage in case of a large power and high voltage transformer. Single HTS wire was used for the primary windings and four parallel wires were used for the secondary windings of the transformer with transposition. A core of the transformer was designed as a shell type core separated with the windings by a cryostat made of GFRP with a room temperature bore. The operating temperature of the HTS windings will be about 65K with sub-cooled liquid nitrogen. A cryogenic cooling system using a GM-cryocooler for this HTS transformer by natural convection of liquid nitrogen was designed. This type of cooling system can be a good option for compactness, efficiency, and reliability of the HTS transformer.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.19-26
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• 2020

Small nuclear reactor features higher power capacity, longer operation life than conventional power sources. It could be an ideal alternative of existing power source applied for special equipment for terrestrial or underwater missions. In this paper, a 25kWe heat pipe cooled reactor power source applied for multiple use is preliminary designed. Based on the design, a thermal-hydraulic analysis code for heat pipe cooled reactor is developed to analyze steady and transient performance of the designed nuclear reactor. For reactor design, UN fuel with 65% enrichment and potassium heat pipes are adopted in the reactor core. Tungsten and LiH are adopted as radiation shield on both sides of the reactor core. The reactor is controlled by 6 control drums with B₄C neutron absorbers. Thermoelectric generator (TEG) converts fission heat into electricity. Cooling water removes waste heat out of the reactor. The thermal-hydraulic characteristics of heat pipes are simulated using thermal resistance network method. Thermal parameters of steady and transient conditions, such as the temperature distribution of every key components are obtained. Then the postulated reactor accidents for heat pipe cooled reactor, including power variation, single heat pipe failure and cooling channel blockage, are analyzed and evaluated. Results show that all the designed parameters satisfy the safety requirements. This work could provide reference to the design and application of the heat pipe cooled nuclear power source.

• 한국컴퓨터정보학회논문지
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• 제19권4호
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• pp.1-8
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• 2014

멀티코어 프로세서는 여러 개의 코어가 하나의 칩에 배치됨에 따라 전력 밀도가 상승하여 높은 발열이 발생한다. 이러한 발열 문제를 해결하기 위해서 최근까지 다양한 연구가 진행되고 있다. 마이크로프로세서의 온도 감소를 위한 기법으로는 기계적 냉각 기법, 동적 온도 관리 기법 등이 있지만 이러한 기법들은 추가적인 냉각 비용이 발생하거나 성능의 저하가 발생한다. 플로어플랜기법은 추가적인 냉각비용이 발생하지 않으며, 성능저하가 거의 발생하지 않는다는 장점을 지닌다. 본 논문에서는 멀티코어 프로세서의 특정 구성요소의 발열 문제를 해결하기 위해 코어 내부 구성요소와 L2 캐쉬의 다양한 플로어플랜을 활용하고자 한다. 실험 결과, 코어의 뜨거운 구성요소를 L2 캐쉬와 인접하게 배치할 경우 칩의 온도 감소에 매우 효과적임을 알 수 있다. 코어를 캐쉬 상단-가운데 배치하는 기본 플로어플랜과 비교하여, 코어를 중앙에 배치하고 뜨거운 구성요소를 L2 캐쉬와 인접하게 배치하는 플로어플랜의 경우에는 $8.04^{\circ}C$, 코어를 외곽에 배치하고 뜨거운 구성요소를 L2 캐쉬와 인접하게 배치하는 플로어플랜의 경우에는 $8.05^{\circ}C$의 최고온도 감소 효과를 보임을 알 수 있다.

• 대한기계학회논문집
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• 제14권2호
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• pp.470-483
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• 1990

Turbulent jet in a crossflow, issuing from a row of holes on a convex surface of 90 .deg. bend duct, is predicted by a 3-dimensional numerical method. The Cartesian coordinate system in adopted in upstream and downstream tangents and the cylindrical polar coordinate system in curved region. The Reynolds stresses and heat fluxes are obtained from a standard k-e model in the core region and van Driest model in the vicinity of the wall. The governing equations are discretized by a finite volume method and solutions are obtained by a locally elliptic calculation procedure. Pressure and convective terms are treated by SIMPLE algorithm and hybrid scheme respectively. A vortex initially induced by the injected jet has been built up due to the interaction with the secondary flow caused by pressure gradient and centrifugal force. The vortex structure has a strong influence on the wall cooling effectiveness. Another vortex like horseshoe is formed in the vicinity of the injection hole and its strength is getting weak as it moves downward.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.709-718
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• 2012

The pressurized water reactor APR1400 adopts DVI (Direct Vessel Injection) for the emergency cooling water in the upper downcomer annulus. The International Standard Problem number 50 (ISP-50) was launched with the aim to investigate thermal hydraulic phenomena during a 50% DVI line break scenario with best estimate codes making use of the experimental data available from the ATLAS facility located at KAERI. The present work describes the calculation results obtained for the ISP-50 using the RELAP5/MOD3.3 system code. The work aims at validation and assessment of the code to reproduce the observed phenomena and investigate about its limitations to predict complicated mixing phenomena between the subcooled emergency cooling water and the two-phase flow in the downcomer. The obtained results show that the overall trends of the main test variables are well reproduced by the calculations. In particular, the pressure in the primary system show excellent agreement with the experiment. The loop seal clearance phenomenon was observed in the calculation and it was found to have an important influence on the transient progression. Moreover, the collapsed water levels in the core are accurately reproduced in the simulations. However, the drop in the downcomer level before the activation of the DVI from safety injection tanks was underestimated due to multi-dimensional phenomena in the downcomer that are not properly captured by one-dimensional simulations.

• Nuclear Engineering and Technology
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• 제41권9호
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• pp.1157-1170
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• 2009

A variety of Generation III/III+ reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world to solve the future energy supply shortfall. Nanofluid coolants showing an improved thermal performance are being considered as a new key technology to secure nuclear safety and economics. However, it should be noted that there is a lack of comprehensible design works to apply nanofluids to Generation III+ reactor designs. In this work, the review of accident scenarios that consider expected nanofluid mechanisms is carried out to seek detailed application spots. The Axiomatic Design (AD) theory is then applied to systemize the design of nanofluid-engineered nuclear safety systems such as Emergency Core Cooling System (ECCS) and External Reactor Vessel Cooling System (ERVCS). The various couplings between Gen-III/III+ nuclear safety features and nanofluids are investigated and they try to be reduced from the perspective of the AD in terms of prevention/mitigation of severe accidents. This study contributes to the establishment of a standard communication protocol in the design of nanofluid-engineered nuclear safety systems.

• 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.