• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2005년도 추계 학술대회논문집
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• pp.39-44
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• 2005

In HANARO, a multi-purpose research reactor of 30 MWth, the emergency water supply system consists essentially of an emergency water storage tank located in the level of about thirteen meter (13 m) above the reactor core, a three inch ('3\%') diameter water injection pipe line including injection valves from the tank to the reactor cooling inlet pipe and a test loop to do periodic system performance test. When the water level of the reactor pool comes down to the extremely low due to a loss of reactor pool water accident the emergency water stored in the tank should be fed to the core by the gravity force and at that time the design flow rate is eleven point four kilogram per second (11.4 kg/s). But it is impossible periodically to measure the injection flow rate under the emergency condition because the normal water level should be maintained during the reactor operation. This paper describes a flow network analysis to simulate the flow rate under the emergency condition. As results, it was confirmed through the analysis results that the calculated flow rate agrees with the design requirement under the emergency condition.

• 전자공학회논문지
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• 제51권9호
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• pp.215-220
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• 2014

본 논문에서는 연구용 원자로의 출력제어기법 설계와 이를 실제 원자로에 적용하여 성능을 검증한 사례를 소개한다. 연구용 원자로의 출력제어를 위해 제안된 설계 원리는 오버슈트(overshoot)의 억제, 출력 증가율의 억제, 그리고 안전해석에 기반한 최대 출력치의 제한이라는 세 가지이며, 이를 만족키 위해 한국원자력연구원 내의 연구용 원자로인 하나로의 설계개념에 기반을 두어 제어 로직의 개념설계, 상세설계, 구현, 시운전을 통해 해외의 원자로에 적용하여 실제 제어 성능을 검증하였다.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.776-792
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• 2021

Conventional integral effect test facilities were constructed to enable the precise observation of thermal-hydraulic phenomena and reactor behaviors under postulated accident conditions to prove reactor safety. Although these facilities improved the understanding of thermal-hydraulic phenomena and reactor safety, applications of new technologies and their performance tests have been limited owing to the cost and large scale of the facilities. Various nuclear technologies converging 4th industrial revolution technologies such as artificial intelligence, drone, and 3D printing, are being developed to improve plant management strategies. Additionally, new conceptual passive safety systems are being developed to enhance reactor safety. A new integral effect test facility having a noticeable scaling ratio, i.e., the (UNIST reactor innovation loop (URI-LO), is designed and constructed to improve the technical quality of these technologies by performance and feasibility tests. In particular, the URI-LO, which is constructed using a transparent material, enables better visualization and provides physical insights on multidimensional phenomena inside the reactor system. The facility design based on three-level approach is qualitatively validated with preliminary analyses, and its functionality as a test facility is confirmed through a series of experiments. The design feature, design validation, functionality test, and future utilization of the URI-LO are introduced.

• 대한기계학회논문집A
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• 제36권10호
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• pp.1283-1288
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• 2012

수소를 대량으로 생산하기 위한 원자력수소생산시스템에서 공정열교환기는 초고온가스로로부터 생성된 초고온 열을 화학반응공정으로 전달하는 핵심기기이다. 한국원자력연구원에 구축되어 있는 소형 가스루프에서 $Hastelloy^{(R)}$-X 로 제작된 중형 공정열교환기 시제품에 대한 성능시험이 계획되어 있다. 본 연구에서는 중형 공정열교환기의 고온구조건전성을 파악하기 위한 선행 연구로서 소형가스루프 시험조건하에서 중형 공정열교환기 시제품의 고온구조해석을 이전 연구에서 확립된 경계조건을 적용하여 수행하였다. 해석결과는 소형가스루프에서의 중형 공정열교환기 시제품에 대한 성능시험 결과와 비교할 예정이다.

• 대한기계학회논문집A
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• 제35권10호
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• pp.1249-1255
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• 2011

공정열교환기는 초고온가스로로부터 생성된 $950^{\circ}C$ 정도의 초고온 열을 대량의 수소를 생산하기 위한 화학반응공정으로 전달하는데 필요한 핵심기기이다. Hastelloy-X 로 제작된 소형 공정열교환기 시제품이 한국원자력연구원에 있는 소형가스루프에서 성능시험이 계획되어 있다. 본 연구에서는 소형가스루프 시험조건하에서 소형 공정열교환기 시제품의 고온 구조건전성을 사전에 평가하기 위한 작업의 일환으로 소형 공정열교환기 시제품에 대한 고온 구조해석 모델링, 거시적 열 해석 및 탄 소성구조 해석을 수행하고 그 결과들을 정리한 것이다. 해석 결과는 공정열교환기 수정 시제품 성능시험장치 설계에 반영할 것이다.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1083-1095
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• 2016

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodiumshowed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.103-112
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• 2017

Application of the supercritical condition in reactor core cooling needs to be properly justified based on the extreme level of parameters involved. Therefore, a numerical study is presented to compare the thermalhydraulic performance of supercritical and single-phase natural circulation loops under low-to-intermediate power levels. Carbon dioxide and water are selected as respective working fluids, operating under an identical set of conditions. Accordingly, a three-dimensional computational model was developed, and solved with an appropriate turbulence model and equations of state. Large asymmetry in velocity and temperature profiles was observed in a single cross section due to local buoyancy effect, which is more prominent for supercritical fluids. Mass flow rate in a supercritical loop increases with power until a maximum is reached, which subsequently corresponds to a rapid deterioration in heat transfer coefficient. That can be identified as the limit of operation for such loops to avoid a high temperature, and therefore, the use of a supercritical loop is suggested only until the appearance of such maxima. Flow-induced heat transfer deterioration can be delayed by increasing system pressure or lowering sink temperature. Bulk temperature level throughout the loop with water as working fluid is higher than supercritical carbon dioxide. This is until the heat transfer deterioration, and hence the use of a single-phase loop is prescribed beyond that limit.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 춘계학술대회논문집
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• pp.308-311
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• 2008

As a severe accident mitigation strategy in a nuclear power plant, ERVC(External Reactor Vessel Cooling) has been proposed. Under ERVC conditions, where a molten corium is relocated in a reactor vessel lower head, a natural circulation two-phase flow is driven in the annular gap between the reactor vessel wall and its insulation. This flow should be sufficient to remove the decay heat of the molten corium and maintain the integrity of the reactor vessel. Preliminary experimental study was performed to estimate the natural circulation two-phase flow. The experimental facility which is one dimensional, the half height, and the 1/238 channel area of APR1400, was prepared and the experiments were carried out to estimate the natural circulation two-phase flow with varying the parameters of the coolant inlet area, the heat rate, and the coolant inlet subcooling. In results, the periodic circulation flow was observed and the characteristics were varied from the experimental parameters. The frequency of the natural circulation flow rate increased as the wall heat flux increased.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2014년도 추계학술대회 논문집
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• pp.110-115
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• 2014

The nuclear power plant was launched by Kori unit 1 in 1978 years. Currently, 23 nuclear power plants have been operating in Korea since 1978 years. The localization was completed for most of the reactor facility from Hanbit(Youngkwang) unit 3&4. However, RCP(Reactor Coolant Pump) and MMIS(Man Machine Interface System) is an important technology that has been excluded from the scope of the technical transfer has been dependent on a specific overseas vendor. Recent success in RCP development through co-operation with government and industries. Developed RCP will be applied to Shin-Hanul unit 1&2 nuclear power plants. The RCP operates in high speed and high pressure condition and only rotating component in the NSSS(Nuclear Steam Supply System). Therefore, the problem of vibration has arisen caused by the hydraulic forces of the working fluid. These forces can influence on the stability characteristics for entire RCS(Reactor Coolant System) loop, and can act as significant destabilizing forces. In this study, vibration evaluation of the pump shaft of development RCP estimated under normal operation and over speed conditions. In order to predict the vibration characteristics and dynamic behavior, modal analysis, critical speed analysis and unbalance response spectrum analysis were performed.

• 한국압력기기공학회 논문집
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• 제15권1호
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• pp.33-39
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• 2019

An 1,500MW advanced power reactor required the standard design approval by a Korean regulatory body in 2014. The reactor has been designed to have a 4-train independent safety concept and a passive auxiliary feedwater system (PAFS). The full power risk or core damage frequency (CDF) of 1,500MW advanced power reactor has been reduced more than that of APR1400. However, the risk during the low power and shutdown (LPSD) operation should be reduced because CDF of LPSD is about 4.7 times higher than that of internal full power. The purpose of paper is to analysis design alternatives to reduce risk during the LPSD. This paper suggests design alternatives to reduce risk and presents sensitivity analysis results.