• 에너지공학
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• 제27권4호
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• pp.27-35
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• 2018

피동원자로건물냉각계통(PCCS)은 사고 발생 시 원자로건물로 방출된 열을 제거하여 원전의 건전성을 보장하기 위해 설계되었다. PCCS의 열제거 성능은 증기-공기 혼합물의 응축열전달에 의해 결정된다. 본 연구에서는 응축열전달계수의 예측 정확도를 향상시키기 위해 새로운 상관식을 이식한 MARS-KS 코드를 사용하여 PCCS의 열제거 성능을 평가하였다. MARS-KS 코드에 사용된 새로운 상관식은 압력, 벽면과냉도, 비응축성 기체 질량분율 및 응축튜브의 종횡비와 같은 열전달계수에 영향을 미치는 변수들을 이용하여 개발하였고, 이는 MARS-KS코드의 기본 응축 모델인 Colburn-Hougen 모델을 대체하여 적용되었다. 대형파단 냉각재상실사고 발생 시 PCCS의 운전에 따른 다양한 열수력학적 변수들을 분석하였고, 열제거 성능 평가를 위해 새로운 상관식이 적용된 MARS-KS 코드의 원자로건물 압력거동 계산결과와 기존의 응축모델을 이용한 해석결과를 비교하였다.

• 에너지공학
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• 제12권4호
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• pp.259-266
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• 2003

안전감압계통 작동시 수조에서 발생하는 공기 기포군의 진동 하중을 줄이기 위해 ABB-Atom sparger에 는 하중저감 링이 설치되어 있다. 하중저감 링이 압력장에 미치는 영향을 알아보기 위해, 본 연구에서는 하중저감 링이 없는 ABB-Atom sparger를 통해 수조 내로 방출되는 공기 기포군의 진동에 대한 수치해석을 상용 열수력 해석 코드인 FLUENT 4.5를 사용하여 수행하였다. 코드에 내재된 다상유동 모델 중 VOF(Volume Of Fluid)모델을 사용하여 물, 공기 및 증기 유동을 모의하였다. 해석결과를 기존의 해석결과와 비교하여 하중저감 링은 벽면 압력 하중을 줄이는 효과가 있음을 확인하였다. 아울러 배관내의 공기량과 배관 입구 조건이 벽면 압력 진동에 미치는 영향도 분석하였다.

• Nuclear Engineering and Technology
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• 제18권2호
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• pp.97-106
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• 1986

1981년 6일 9일 원자력 1호기에서 발생한 77.5% 출력상태에서의 외부전원상실사고를 열, 수력학적최적계산용 코드인 RELAP5/MODl/NSC를 사용하여 모의하였으며 해석결과는 발전소 실측자료와 잘 일치하였다. 원자로 냉각재펌프의 트립에 따른 flow coastdown후에 hot-cold leg온도차에 의하여 자연순환 유동이 형성됨이 확인되었으며 실측자료와 잘 일치하여 이와 관련된 전산코드의 열수력학 적모델의 타당성을 입증할 수 있었다. 또한 위의 사고전개가 정상운전상태인 전출력(100%)에서 재발하였을 경우를 가정하여 해석하였다. 이러한 해석을 통하여 보조급수의 공급과 더불어 증기발생기 PORV의 적절한 작동으로 원자력 1호기 노심잔열을 제거하여 안전성에 문제점을 야기하지 않음을 입증하였다. 최적 계산방법에 의한 사고해석에서는 turbine stop valve 작동시간, 증기 발생기 PORV 설정치 등 non-safety 관련요소들의 특성에 대한 정화한 모의가 필수적이다.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3236-3246
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• 2021

Lessons learned from the Fukushima Daiichi nuclear power plant accident directed that multiple failures should be considered more seriously rather than single failure in the licensing bases and safety cases because attempts to take accident management measures could be unsuccessful under the high radiation environment aggravated by multiple failures, such as complete loss of electric power, uncontrollable loss of coolant inventory, failure of essential safety function recovery. In the case of the complete loss of electric power called station blackout (SBO), if there is no mitigation action for recovering safety functions, the reactor core would be overheated, and severe fuel damage could be anticipated due to the failure of the active heat sink. In such a transient condition at CANDU-6 plants, the seal failure of the primary heat transport (PHT) pumps can facilitate a consequent increase in the fuel sheath temperature and eventually lead to degradation of the fuel integrity. Therefore, it is necessary to specify the regulatory guidelines for multiple failures on a licensing basis so that licensees should prepare the accident management measures to prevent or mitigate accident conditions. In order to explore the efficiency of implementing accident management strategies for CANDU-6 plants, this study proposed a realistic accident analysis approach on the SBO transient with multiple-failure sequences such as seal failure of PHT pumps without operator's recovery actions. In this regard, a comparative study for two PHT pump seal failure modes with and without coolant seal leakage was conducted using a best-estimate code to precisely investigate the behaviors of thermal-hydraulic parameters during transient conditions. Moreover, a sensitivity analysis for different PHT pump seal leakage rates was also carried out to examine the effect of leakage rate on the system responses. This study is expected to provide the technical bases to the accident management strategy for unmitigated transient conditions with multiple failures.

• Nuclear Engineering and Technology
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• 제39권3호
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• pp.193-206
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• 2007

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.803-811
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• 2024

Nuclear accidents such as Fukushima Daiichi have highlighted the potential of passive safety systems to replace or complement active safety systems as part of the overall prevention and/or mitigation strategies. In addition, passive systems are key features of Small Modular Reactors (SMRs), for which they are becoming almost unavoidable and are part of the basic design of many reactors available in today's nuclear market. Nevertheless, their potential to significantly increase the safety of nuclear power plants still needs to be strengthened, in particular the ability of computer codes to determine their performance and reliability in industrial applications and support the safety demonstration. The PASTELS project (September 2020-February 2024), funded by the European Commission "Euratom H2020" programme, is devoted to the study of passive systems relying on natural circulation. The project focuses on two types, namely the SAfety COndenser (SACO) for the evacuation of the core residual power and the Containment Wall Condenser (CWC) for the reduction of heat and pressure in the containment vessel in case of accident. A specific design for each of these systems is being investigated in the project. Firstly, a straight vertical pool type of SACO has been implemented on the Framatome's PKL loop at Erlangen. It represents a tube bundle type heat exchanger that transfers heat from the secondary circuit to the water pool in which it is immersed by condensing the vapour generated in the steam generator. Secondly, the project relies on the CWC installed on the PASI test loop at LUT University in Finland. This facility reproduces the thermal-hydraulic behaviour of a Passive Containment Cooling System (PCCS) mainly composed of a CWC, a heat exchanger in the containment vessel connected to a water tank at atmospheric pressure outside the vessel which represents the ultimate heat sink. Several activities are carried out within the framework of the project. Different tests are conducted on these integral test facilities to produce new and relevant experimental data allowing to better characterize the physical behaviours and the performances of these systems for various thermo-hydraulic conditions. These test programmes are simulated by different codes acting at different scales, mainly system and CFD codes. New "system/CFD" coupling approaches are also considered to evaluate their potential to benefit both from the accuracy of CFD in regions where local 3D effects are dominant and system codes whose computational speed, robustness and general level of physical validation are particularly appreciated in industrial studies. In parallel, the project includes the study of single and two-phase natural circulation loops through a bibliographical study and the simulations of the PERSEO and HERO-2 experimental facilities. After a synthetic presentation of the project and its objectives, this article provides the reader with findings related to the physical analysis of the test results obtained on the PKL and PASI installations as well an overall evaluation of the capability of the different numerical tools to simulate passive systems.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.721-733
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• 2020

In nuclear engineering, the occurrence of critical heat flux (CHF) is complicated for rod bundle, and it is much more difficult to predict the CHF when it is in natural circulation under motion condition. In this paper, the dryout-type CHF is investigated for the rod bundle in a natural circulation loop under rolling motion condition based on the coupled analysis of subchannel method, a one-dimensional system analysis method and a CHF mechanism model, namely the three-fluid model for annular flow. In order to consider the rolling effect of the natural circulation loop, the subchannel model is connected to the one-dimensional system code at the inlet and outlet of the rod bundle. The subchannel analysis provides the local thermal hydraulic parameters as input for the CHF mechanism model to calculate the occurrence of CHF. The rolling motion is modeled by additional motion forces in the momentum equation. First, the calculation methods of the natural circulation and CHF are validated by a published natural circulation experiment data and a CHF empirical correlation, respectively. Then, the CHF of the rod bundle in a natural circulation loop under both the stationary and rolling motion condition is predicted and analyzed. According to the calculation results, CHF under stationary condition is smaller than that under rolling motion condition. Besides, the CHF decreases with the increase of the rolling period and angular acceleration amplitude within the range of inlet subcooling and mass flux adopted in the current research. This paper can provide useful information for the prediction of CHF in natural circulation under motion condition, which is important for the nuclear reactor design improvement and safety analysis.

• 에너지공학
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• 제11권3호
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• pp.237-246
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• 2002

한국형차세대원자로 APR-1400의 안전감압계통이 작동하면 물, 공기 및 증기가 sparger를 통해 격납건물 내 핵연료재장전 수조로 차례로 방출된다. 방출 과정 중 생기는 여러 현상 중에서 수조 내의 공기 기포군은 저주파, 고진폭의 진동 하중을 발생하며, 주파수가 침수 구조물의 고유 주파수와 거의 같은 경우에는 구조물에 심각한 영향을 줄 수 있다. 이러한 현상은 복잡하기 때문에 주파수와 하중에 대한 규명은 주로 실험에 의존해 왔으며 수치해석적 연구는 이루어지지 않았다. 본 연구에서는 sparger를 통해 수조 내로 방출되는 공기 기포군의 거동에 대한 수치해석을 상용 열수력 해석 코드인 FLUENT Version 4.5를 사용하여 수행하였다. 다상유동 해석모델중 VOF(Volume Of Fluid)모델을 사용하여 물, 공기 및 증기 등의 다상유동을 모의하였다. 해석결과를 sparger 개발을 위해 ABB-Atom이 수행하였던 실험결과와 비교하여 만족할만한 결과를 얻었다.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.613-624
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• 2013

Multi-dimensional two-phase phenomena occur in many industrial applications, particularly in a nuclear reactor during steady operation or a transient period. Appropriate modeling of complicated behavior induced by a multi-dimensional flow is important for the reactor safety analysis results. SPACE, a safety analysis code for thermal hydraulic systems which is currently being developed, was designed to have the capacity of multi-dimensional two-phase thermo-dynamic phenomena induced in the various phases of a nuclear system. To validate the performance of SPACE, a two-dimensional two-phase flow test was performed with slab geometry of the test section having a scale of $1.43m{\times}1.43m{\times}0.11m$. The test section has three inlet and three outlet nozzles on the bottom and top gap walls, respectively, and two outlet nozzles installed directly on the surface of the slab. Various kinds of two-dimensional air/water flows were simulated by selecting combinations of the inlet and outlet nozzles. In this study, two-dimensional two-phase void fraction profiles were quantified by measuring the local gap impedance at 225 points. The flow conditions cover various flow regimes by controlling the flow rate at the inlet boundary. For each selected inlet and outlet nozzle combination, the water flow rate ranged from 2 to 20 kg/s, and the air flow rate ranged from 2.0 to 20 g/s, which corresponds to 0.4 to 4 m/s and 0.2 to 2.3 m/s of the superficial liquid and gas velocities based on the inlet port area, respectively.

• Nuclear Engineering and Technology
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• 제18권3호
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• pp.200-208
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• 1986

원자로 냉각 계통의 배관 파열에 근거한 냉각재 상실 사고를 방출계수 0.4에 대하여 분석하였다. 분석은 원자로 냉각계통의 배관 파열에 의하여 발생된 감압부터 노심 복구까지의 전 과도 상태를 포함한다. 계통 열수력과 핵연료 성능 평가를 위하여 BLOWDOWN 단계에서는 RELAP4/MOD6-EM 코드와 RELAP4/MOD6-HOT CHANNEL 코드를 사용하였으며 REFLOOD 단계에서는 RELAP4/ MOD6-FLOOD 코드와 TOODEE2 코드를 각각 사용하였다. LOWER PLENUM 충전을 고려하기 위하여 DOWNCOMER에서 증기-물역방향 유동과 과열벽효과를 근사하여 간단한 해석적 모델이 개발되었다. EOB 발생시의 정보를 근거로 하여 재충전지속 시간과 초기 복구 온도가 계산되었으며 RELAP4/MOD6에 의한 분석결과와 비교하여 상당한 일치를 보였다. 또한, 조기 EOB 발생에 영향을 미치는 계통변수의 연구가 수행되어졌다. DOWNCOMER와 UPPER HEAD사이의 마찰손실이 조기 EOB 발생에 지대한 영향을 미쳤으며 적당한 마찰손실계수의 선택을 통하여 조기 EOB 발생을 방지할 수 있었다. 노심 nodalization이 여섯 개인 경우와 세 개인 경우의 분석 결과가 계통열수력학적 면에서 유사한 결과를 나타내지만, 좋은 결과를 얻기 위하여 전자의 경우가 요구된다.