• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2743-2759
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• 2020

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the T_hot and T_cold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.

• 암석학회지
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• 제23권3호
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• pp.249-259
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• 2014

해수의 순환이 지구표면의 열 분배 및 궤도함수주기와 천년주기의 기후변화와 밀접한 관계가 있음이 알려지면서 과거 해수의 흐름을 지시할 수 있는 지시자의 개발이 관심을 끌고 있다. 이 논평에서는 해수 자생성 네오디뮴 동위원소 비를 해수순환의 지시자로 활용하는 원리와 분석방법 그리고 두 가지 적용사례를 소개한다. 먼저 지난 빙하기-간빙기와 아빙기-아간빙기에 걸쳐 북대서양심층수 세기의 변화를 명확히 볼 수 있는 예를 소개한다. 다음으로는 북극해에서 담수의 유입과 해수순환을 재구성한 예를 보인다.

• 한국지구과학회지
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• 제31권3호
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• pp.246-258
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• 2010

대류 상자 실험을 통해 대류 순환 모형을 유추하지 못하는 학생들의 원인을 분석하여 대류 실험을 재설계하였다. 초기 형식적 조작기에 있는 학생들이지만 직접 보지 못한 자연 현상에 대해서는 유추하기가 어려우므로 자연 현상을 축소한 대류 상자를 개발하여 학생들의 과학적 이해도를 높일 수 있는 방법을 연구하였다. 기존 대류 상자 실험의 문제점을 개선하여 제작한 대류 상자를 통해 수업에 적용한 결과 학생들의 이해도가 크게 증가하였다. 바람의 이동 현상만을 관찰하는 것이 아니라 향을 여러 위치에서 넣어봄으로써 과학적 사고를 자극하고 직접 관찰할 수 있었다. 또한 자연 현상의 축소판에 가깝게 설계하여 학생들이 과학적 오개념을 갖지 않는다. 모둠원이 양방향에서 관찰함으로써 수업의 참여도가 높아졌다. 새로 개발된 대류상자는 기존 대류상자에 비하여 저렴한 비용으로 제작할 수 있으며, 선명한 대류 현상을 볼 수 있다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.2194-2199
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• 2007

An experimental study is performed to investigate the thermal and flow characteristics of subcooled liquid nitrogen in a natural circulation loop. Experimental apparatus is designed and constructed such that a closed loop is cooled at the top by a cryocooler and heated nearly at the bottom by cartridge heaters. Steady state is obtained by controlling the heating power to the cartridge heaters and a thin-film heater to reduce the cooling power of the cryocooler. Temperature is measured at several locations of the loop and the mass flow rate through the loop is estimated from the energy balance in terms of the measured temperatures. Experiment is repeated for various values of the vertical height between the cooling and heating parts. The results show that the heat transfer capability of the loop has a maximum at a certain value of height. The optimal height to maximize the heat transfer is in a good agreement with analytical prediction to take into account the buoyancy and viscous forces in the loop.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.514-519
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• 1997

The MRX is an integral type ship reactor with 100 MWt power, which is designed by Japan Atomic Energy Research Institute. It is characterized by integral type PWR, in-vessel type control roe drive mechanism, water-filled containment vessel and passive decay heat removal system. Marine reactor should have high passive safety. Therefore, in this study, we simulated the loss of flow accident to verify the passive decay heat removal by natural circulation using RETRAN-03 code. auxiliary feed water systems are used for decay heat removal mechanism and results are compared with the loss of flow accident analysis using emergency decay heat removal system by JAERI. Results are very similar to case of EDRS 1 loop operation in JAERI analysis and decay heat is successfully removed by natural circulation.

• 설비공학논문집
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• 제23권9호
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• pp.595-603
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• 2011

In this study, an effort has been made to analyze the subcooled boiling heat transfer in a natural circulation vertical evaporator. To verify the analysis, a small-scale model was made and tested. The friction correlation by Ueda, void fraction and quality correlation by Saha and Zuber along with the superposition heat transfer model by Rohsenow yielded a satisfactory agreement with the model test data. The analysis was extended to simulate a 1 ton/day concentration system. Comparison with the test results of 1 ton/day prototype revealed that the data were overpredicted by 13%. The capacity of the prototype was 1.2 ton/day with COP of 5.77.

• 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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• 제30권5호
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• pp.424-434
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• 1998

RELAP5/Mod3.1 code was assessed with the semiscale experiment S-NC-3, and S-NC-4, which simulated the two-phase natural circulation and reflux condensation for the SBLOCA of PWR, respectively . Test S-NC-3 and S-NC-4 calculation results showed that RELAP5/Mod3.1 quite well describes the influence of steam generator secondary side heat transfer degradation on both two-phase natural circulation and reflux condensation. A comparison between the calculated and measured two-phase mass flow rate in test S-NC-3 shows good agreement for primary mass inventory more than 92%. And RELAP5/Mod3.1 have a good mass flow rate prediction capability for the transient such as S-NC-4 except some flow oscillations. The reflux flow rate for S-NC-4 test is under predicted, and the overall results verify that the correct prediction of the reduced liquid level appears to be required for the correct calculation of the overall phenomena.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3597-3611
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• 2021

The present paper studies the thermal-hydraulic behaviour of the rectangular supercritical natural circulation loop (SCNCL) using numerical model of one dimensional. Then the results of this model is confirmed with experimental and benchmark results. Variations with several geometric parameters like loop diameter, riser length, and heater length and operating conditions like heater inlet enthalpy, pressure, friction factor, and inlet and exit loss coefficient on steady-state performance are investigated for various orientations like HHHC, HHVC, VHVC and VHHC of the heater and cooler. The chances of existing static instability (Ledinegg excursion) has been investigated, which reveals that it can arise only in a low inlet enthalpy condition, far from the suggested various orientations of heater and cooler.

• 한국압력기기공학회 논문집
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• 제19권1호
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• pp.36-43
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• 2023

Lots of researches have been conducted on in-vessel retention (IVR) to prevent or mitigate severe accident in nuclear power plants. Various methodologies were proposed and the external reactor vessel cooling was selected as a part of promising IVR strategy. In this study, the strategy is strengthened by enhancing the natural circulation performance through the adoption of insulation in the reactor cavity. A thermal analysis was carried out based on an assumed accident scenario and its results were used as boundary conditions for subsequent seven flow analysis cases. By comparing the natural circulation performance, effects of annular gaps and insulation shapes on the mass flow rate and flow velocity were quantified. The improvement in cooling performance can be reflected in actual design via detailed assessment.