• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.1927-1932
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• 2004

A 1/21.6 scaled non-heating experimental facility was prepared utilizing the results of a scaling analysis to simulate the APR1400 reactor and insulation system. The behaviors of the air bubble-induced two-phase natural circulation flow in the insulation gap were observed, and the liquid mass flow rates driven by natural circulation loop were measured by varying the injected air flow rate and distribution. As the injected air flow rates increased, the natural circulation flow rates also increased. Both the longitudinal and the latitudinal distributions of the injected air affected the natural circulation flow rates, especially, the longitudinal effect is more larger.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.3086-3091
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• 2007

When a molten corium is relocated in a lower head of a reactor vessel, the ERVC (External Reactor Vessel Cooling) system is actuated as coolant is supplied into a reactor cavity to remove a decay heat from the molten corium during a severe accident. To achieve this severe accident mitigation strategy, the two-phase natural circulation flow in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. For this reason, one-dimensional natural circulation flow tests were conducted to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled-down as the half height and 1/238 rectangular channel area of the APR1400 reactor vessel. As the water inlet area increased, the natural circulation mass flow rate asymptotically increased, that is, it converged at a specific value. And the circulation mass flow rate also increased as the outlet area, injected air flow rate, and outlet height increased. But the circulation mass flow rate was not changed along with the external water level variation if the water level was higher than the outlet height.

• 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.

• 대한기계학회논문집B
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• 제22권12호
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• pp.1763-1773
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• 1998

In the present experimental study, the flow behavior in a semi-closed two-phase natural circulation loop was examined. Water was used as the working fluid. Heat flux, heater-inlet subcooling, and flow restrictions at the heater-inlet and at the expansion-tank-line were taken as the controlling parameters Six circulation modes were identified by changing heat flux and inlet subcooling conditions ; single-phase continuous circulation, periodic circulation (A), two-phase continuous circulation, and periodic circulations (B), (C), and (D). Among these, the single-phase and two-phase continuous-circulation modes exhibit no significant oscillations and are considered to be stable. Periodic circulation (A) is characterized by the large amplitude two-phase f10w oscillations with the temporal single-phase circulation between them, while periodic circulation (B) featured by the flow oscillations with continuous boiling inside the heater section. Periodic circulation (C) appears to be the manometric oscillation with continuous boiling. Periodic circulation (D) has the longer period than periodic circulation (B) and a substantial amount of liquid flow back and forth through the expansion-tank-line periodically ; this mode is considered the pressure drop oscillation. Parametric study shows that the increases of the inlet- and expansion-tank-line- restrictions and the decrease of inlet subcooling broaden the range of the stable two-phase(continuous circulation) mode.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2789-2802
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• 2020

Small Modular Reactors (SMRs) are attracting wide attention due to their outstanding performance, extensive studies have been carried out for lead-based fast reactors (LFRs) that cooled with Lead or Lead-bismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRs and LFRs development. One of the challenges for the design small modular natural circulation LFR is to master the natural circulation thermal-hydraulic performance in the reactor primary circuit, while the natural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, the primary loop layout and the operating state of secondary cooling system etc. Thus, accurate predicting the natural circulation LFRs thermal-hydraulic features are highly required for conducting reactor operating condition evaluate and Thermal hydraulic design optimization. In this study, a thermal-hydraulic analysis code is developed for small modular natural circulation LFRs, which is based on several mathematical models for natural circulation originally. A small modular natural circulation LBE cooled fast reactor named URANUS developed by Korea is chosen to assess the code's capability. Comparisons are performed to demonstrate the accuracy of the code by the calculation results of MARS, and the key thermal-hydraulic parameters agree fairly well with the MARS ones. As a typical application case, steady-state analyses were conducted to have an assessment of thermal-hydraulic behavior under nominal condition, and several parameters affecting natural circulation were evaluated. What's more, two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacity were established. The analyses show that the core thermal power, thermal center difference and flow resistance is the main factors affecting the reactor natural circulation. Improving the core thermal power, increasing the thermal center difference and decreasing the flow resistance can significantly increase the reactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulation capacity of natural circulation LFR under normal operating conditions.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.826-833
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• 2022

A two-phase natural circulation test using SMART integral test loop (SMART-ITL) was conducted to explore thermo-hydraulic phenomena of two-phase natural circulation in the SMART reactor. Specifically, the test examined the natural circulation in the primary loop under a stepwise coolant inventory loss while keeping the core power constant at 5% of the scaled full power. Based on the test results, three flow regimes were observed: single-phase natural circulation (SPNC), two-phase natural circulation (TPNC), and boiler-condenser natural circulation (BCNC). The flow rate remained steady in the SPNC, slightly increased in the TPNC, and dropped abruptly and maintained in the BCNC. Using a natural circulation flow map, the natural circulation characteristic in the SMART-ITL was compared with those in pressurized water reactor simulators. In the SMART-ITL, a BCNC regime appeared instead of siphon condensation and reflux condensation regimes because of the use of once-through steam generators.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1851-1859
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• 2022

Experimental research on flow characteristics of open natural circulation system was performed, to figure out the mechanism of the open natural circulation behaviors. The influence factors, such as the heating power, the inlet subcooled and the level of cooling tank on the flow characteristics of the system were examined. It was shown that within the scope of the experimental conditions, there are five flow types: single-phase stable flow, flash and geyser coexisting unstable flow, flash stable flow, flash unstable flow, and flash and boiling coexisting unstable flow. The geyser flow in flash and geyser coexisting unstable flow is different from classic geysers flow. The flow oscillation period and amplitude of the former are more regular, is a newly discovered flow pattern. By drawing the flow instability boundary diagram and sorting out the flow types, it is found that the two-phase unstable flow is mainly characterized by boiling and flash, which determine the behavior of open natural circulation respectively or jointly. Moreover, compared with full liquid level system, non-full liquid level system is more prone to boiling phenomenon, and the range of heat flux density and undercooling degree corresponding to unstable flow is larger.

• 대한기계학회논문집
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• 제17권5호
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• pp.1313-1320
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• 1993

본 연구에서는 가시화가 가능한 상승부를 가진 개방된2상 자연순환 회로를 설치하여 각 운전 조건에 따른 순환 유동 특성을 살펴보았다. 즉, 가열량 증가에 따 른 상승부에서의 2상 유동 양식의 변화를 관찰하였고 동시에 가열기 입구 과냉 액체의 순환 유속 및 상승부의 기공률(void fraction)을 측정하였다. 또한 가열기 입구 및 출구에 설치된 밸브의 마찰저항, 가열기 입구 액체의 과냉 정도, 그리고 충전수위등이 전반적인 유동특성에 미치는 영향을 연구하였다.

• 한국추진공학회지
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• 제23권5호
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• pp.43-52
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• 2019

극저온 유체를 사용하는 발사체는 극저온 유체의 자연순환회로를 이용하여 발사체의 엔진 입구를 냉각한다. 자연순환회로의 질량유량은 순환시스템을 구성하는 배관의 길이 및 직경과 시스템으로 들어오는 열유입에 의하여 결정된다. 극저온 유체의 자연순환회로의 순환 검증 및 질량유량 측정을 위하여 실험을 진행하였으며, 이론적 계산 결과와 비교하였다. 비교 결과 12%의 오차가 있음을 확인하였다. 이 결과를 바탕으로 발사체 상단에서 저중력 구간 및 가속 구간에서의 자연순환 질량유량을 예측한 내용을 포함한다. 가속구간에서는 산화제탱크가 100 kPa 내외로 유지하는 것이 자연순환유량 증가에 이로웠으며, 저중력구간에서는 중력가속도의 크기에 따른 최적 압력으로 조절해야 자연순환유량의 최고값을 유지할 수 있었다.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.72-83
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• 2022

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.