• 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.

• 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.

• 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.

• 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.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2002년도 유체기계 연구개발 발표회 논문집
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• pp.419-424
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• 2002

The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial criticality In February, 1995. Many experiments should be safely performed to activate the utilization of the HANARO. A flow simulation facility is being developed for the endurance test of reactivity control units for extended life times and the verification of structural integrity of those experimental facilities prior to loading in the HANARO. This test facility is composed of three major parts; a half-core structure assembly, flow circulation system and support system. The flow circulation system is composed of a circulation pump, a core flow pipe, a core bypass flow pipe and instruments. The system is to be filled with de-mineralized water and the flow should be met the design flow to simulate similar flow characteristics in the core channel of the half-core test facility to the HANARO. This paper, therefore, describes an analytical analysis to study the flow behavior of the system. The computational flow analysis has been performed for the verification of system pressure variation through the three-dimensional analysis program with standard k-$\epsilon$ turbulence model and for the verification of the structural piping integrity through the finite element method. The results of the analysis are satisfied the design requirements and structural piping integrity of flow circulation system.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.2626-2631
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• 2007

Air-water two phase natural circulation flow in the T-HERMES (Thermo-Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow)-1D experiment has been evaluated to verify and evaluate the experimental results by using the RELAP5/MOD3 computer code. The RELAP5 results have shown that an increase in the coolant inlet area leads to an increase in the water circulation mass flow rate. However, the water outlet area does not effective on the water circulation mass flow rate. As the coolant outlet moves to a lower position, the water circulation mass flow rate decreases. The water level is not effective on the water circulation mass flow rate. As the height increases in the air injection part, the void fraction increases. However, the void fraction in the upper part of the air injector maintains a constant value. An increase in the air injection mass flow rate leads to an increase in the local void fraction, but it is not effective on the local pressure.

• 한국유체기계학회 논문집
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• 제7권1호
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• pp.30-35
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• 2004

The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial criticality in February, 1995. Many experiments should be safely performed to activate the utilization of the HANARO. HANARO flow simulation facility is being developed for the endurance test of reactivity control units for extended life time and the verification of structural integrity of those experimental equipments prior to loading in the HANARO. This facility is composed of three major parts; a half-core structure assembly, a flow circulation system and a support system. The flow circulation system is composed of a circulation pump, a core flow piping, a core bypass flow piping and instruments. The system is to be filled with de-mineralized water and the flow should be met the design requirements to simulate a similar flow characteristics in the core channel of the half-core structure assembly to the HANARO. This paper, therefore, presents an analytical analysis to study the flow behavior of the system. Computational flow analysis has been performed for the verification of system pressure variation through the three-dimensional analysis program with the standard $k-{\epsilon}$ turbulence model and for the verification of the structural piping integrity through the finite element method. According to the analysis results, it could be said that the design requirements and the structural piping integrity of the flow circulation system are satisfied.

• 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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.2141-2145
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• 2004

As part of study on thermal hydraulic behavior in the reactor cavity under external vessel cooling in the APR (Advanced Power Reactor) 1400, one dimensional two phase flow of steady state in the reactor cavity have been analyzed to investigate a coolant circulation mass flow rate in the annulus region between the reactor vessel and the insulation material using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that a two phase natural circulation flow of 300 - 600 kg/s is generated in the annulus region between the reactor vessel and the insulation material when the external vessel cooling has been applied in the APR 1400. An increase in the heat flux of the inner vessel leads to an increase of the coolant mass flow rate. An increase in the coolant outlet area leads to an increase in the coolant circulation mass flow rate, but the coolant inlet area does not effective on the coolant circulation mass flow rate. The change of the lower coolant outlet to a lower position affects the coolant circulation mass flow rate, but the variation trend is not consistent.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1955-1962
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• 2020

In a sudden shutdown of primary pump or coolant loss accident in a marine nuclear power plant, the primary flow decreases rapidly in a transition process from forced circulation (FC) to natural circulation (NC), and the lower flow enters the steam generator (SG) causing reverse flow in the U-tube. This can significantly compromise the safety of nuclear power plants. Based on the marine natural circulation steam generator (NCSG), an experimental loop is constructed to study the characteristics of reverse flow under middle-temperature and middle-pressure conditions. The transition from FC to NC is simulated experimentally, and the characteristics of SG reverse flow are studied. On this basis, the experimental loop is numerically modeled using RELAP5/MOD3.3 code for system analysis, and the accuracy of the model is verified according to the experimental data. The influence of the flow variation rate on the reverse flow phenomenon and flow distribution is investigated. The experimental and numerical results show that in comparison with the case of adjusting the mass flow discontinuously, the number of reverse flow tubes increases significantly during the transition from FC to NC, and the reverse flow has a more severe impact on the operating characteristics of the SG. With the increase of flow variation rate, the reverse flow is less likely to occur. The mass flow in the reverse flow U-tubes increases at first and then decreases. When the system is approximately stable, the reverse flow is slightly lower than obverse flow in the same U-tube, while the flow in the obverse flow U-tube increases.