• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3164-3182
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• 2023

The CRUD on the fuel cladding under the pressurized water reactor (PWR) operating condition causes several issues. The CRUD can act as thermal resistance and increases the local cladding temperature which accelerate the corrosion process. The hideout of boron inside the CRUD results in axial offset anomaly and reduces the plant's shutdown margin. Recently, there are efforts to revise the acceptance criteria of emergency core cooling systems (ECCS), and additionally require the modeling of the thermal resistance effect of the CRUD during the performance analysis. There is an urgent need for the evaluation of the effect of the CRUD deposition on the cladding heat transfer under PWR operating conditions, but the experimental database is very limited. The experimental facility called DISNY was designed and constructed to analyze the CRUD-related multi-physical phenomena, and the performance analysis of the constructed DISNY facility was conducted. The thermal-hydraulic and water chemistry conditions to simulate the CRUD growth under PWR operating conditions were established. The design characteristics and feasibility of the DISNY facility were validated by the MARS-KS code analysis and separate performance tests. In the current study, detailed design features, design validation results, and future utilization plans of the proposed DISNY facility are presented.

• 에너지공학
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• 제21권3호
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• pp.221-227
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• 2012

The CANDU 9 shield cooling system was designed and layout with the objective of promoting natural circulation on loss of forced flow. In the present study, the shield cooling natural circulation was analyzed using verified the thermal-hydraulic code when the coolant pump or the heat exchanger was lost. This study showed that thermosyphoning cooled the end shields and prevented the end shields and the reserve water tank from boiling for at least 8 hours on loss of the shield cooling pumps but the heat exchangers still operational. With the loss of both pumps and heat exchangers, the end shields remain subcooled for up to 4 hours. To enhance thermosyphoning, the bypass connection to the line from the reserve water tank should be relocated to a point as low as possible.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3918-3929
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• 2021

In this study, we experimentally investigate of a two-phase natural circulation loop that functions as a passive containment cooling system (PCCS). The experimental apparatus comprises two loops: a hot loop, for simulating containment under severe accidents, and a natural circulation loop, for simulating the PCCS. The experiment is conducted by controlling the pressure and inlet temperature of the hot loop in the range of 0.59-0.69 MPa (abs) and 119.6-158.8 ℃, respectively. The heat balance of the hot loop is established and compared with a natural circulation loop to assess the thermal reliability of the experimental apparatus, and an additional system is installed to measure the vapor mass flow rate. Furthermore, the thermal-hydraulic characteristics are considered in terms of a temperature, mass flow rate, heat transfer coefficient (HTC), etc. The flow rate of the natural circulation loop is induced primarily by flashing, and a distortion is observed in the local HTC because of the fully develop as well as subcooled boiling. As a result, we present the amount of heat capacity that the PCCS can passively remove according to the experimental conditions and compared the heat transfer performance using Chen's and Dittus-Boelter correlation.

• 에너지공학
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• 제25권1호
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• pp.69-75
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• 2016

기포크기는 다차원 이상유동에서 정확한 기포거동의 예측을 위해 중요한 인자이다. 현재 CFD 코드인 STAR CCM+에서는 유동채널에서 기포크기예측을 위해 역학적인 기포크기모델인 $S{\gamma}$ 모델을 제공하고 있다. 기포크기 예측을 위한 또 다른 모델로써 고압조건의 과냉 비등 실험인 DEBORA 실험을 바탕으로 개발된 Yun 모델이 있다. 본 연구에서는 상용 CFD 코드인 STAR CCM+ ver. 10.02를 이용하여 물-공기 이상유동에 대한 수치해석을 통해 $S{\gamma}$ 모델과 Yun 모델의 성능을 확인하고 평가하였다. 이를 위해 두 모델은 수직관에서의 물-공기 실험인 DEDALE 실험과 Hibiki 등의 실험에 대하여 평가되었다. 해석 결과 $S{\gamma}$ 모델은 이상유동 인자들을 합리적으로 예측하였으며, Yun 모델은 저압조건의 물-공기 유동에는 적합하지 않음을 확인하였다.

• 에너지공학
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• 제12권2호
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• pp.84-92
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• 2003

구리밀봉 증기발생기의 열적크기 계산을 위한 1차원 열수력코드를 개발하였다. 고온 및 저온측 전열관사이의 구리의 열전도는 1차원으로 가정하고, 전열관 피치의 함수로 열저항을 구하였다 물/증기측 유동영역은 아냉, 포화핵비등, 포화막비등, 과열영역의 4 구간으로 구분하였다. 매개변수 연구를 위한 전열관 갯수는 250에서 3500이며, P/D비율은 각각1.4, 1.6, 1.8로 하였다. 계산결과, 전열관 갯수가 2500일 때 전열관 길이는 약 12 m, 직경은 약 3 m이다. P/D를 증가시키면 구리에 의한 열저항 성분이 증가하지만 전열관 길이는 큰 차이가 없음을 알 수 있었다.

• Nuclear Engineering and Technology
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• 제55권2호
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• pp.460-474
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• 2023

A simple but accurate mathematical model is crucial for dynamic simulations and controller design of helical coil once-through steam generator (OTSG). This paper presents a three-region movable boundary dynamic model of the helical coil OTSG. Based on the secondary side fluid conditions, the OTSG is divided into subcooled region (two control volumes), two-phase region (two control volumes) and superheated region (three control volumes) with movable boiling boundaries between each region. The nonlinear dynamic model is derived based on mass, energy and momentum conservation equations. And the linear model is obtained by using the transfer function and state space transformation, which is a 37-order model of five input and three output. Validations are made under full-power steady-state condition and four transient conditions. Results show good agreements among the nonlinear model, linear model and the RELAP5 model, with acceptable errors. This model can be applied to dynamic simulations and controller design of helical coil OTSG with constant primary-side flow rate.