• 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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• 제54권2호
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• pp.554-564
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• 2022

The natural circulation phenomena occurring in fully integrated nuclear reactors are associated with a unique formation mechanism. The phenomenon results from a structural feature of these reactors involving upward flow from the core, located in the central-bottom region of a single vessel, and downward flow to the steam generator in the annulus region. In this study, to understand the natural circulation in a single vessel involving a multi-layered flow path, single-phase and two-phase natural circulation tests were performed using the SMART-ITL facility, and validation analysis of the TASS/SMR-S code was performed by comparing the corresponding test results. Three single-phase natural circulation tests were sequentially conducted at 15%, 10%, and 5% of full-scaled core-power without RCP operation, following which a two-phase natural circulation test was successively conducted with an artificial discharge of coolant inventory. The simulation capability of the TASS/SMR-S code with respect to the natural circulation phenomena was validated against the test results, and somewhat conservative but reasonably comparative results in terms of overall thermalhydraulic behavior were shown.

• 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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• 제33권6호
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• pp.638-651
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• 2001

The present study is to assess the applicability of the best-estimate thermal-hydraulic code, MARS 1.4, for the analysis of thermal-hydraulic behavior in PWRs during natural circulation conditions. The code simulates a natural circulation test, BETHSY test 4. la, which was conducted on the integral test facility of BETHSY. The test represented the cooling states of the primary cooling system under single-phase natural circulation, two-phase natural circulation and the reflux condensation mode with conditions corresponding to the residual power, 2% of the rated core power value and 6.8 MPa at the secondary system. Based on MARS 1.4 calculations, the major thermal-hydraulic behaviors during natural circulation are evaluated and the differences between the experimental data and calculated results are identified. The calculated results show generally good behavior with regard to the experimental results; the region of single-phase natural circulation is 100-92% of the initial mass inventory, two-phase natural circulation is 84-63 %, and the reflux condensation mode occurred below 58 %. U-tubes empty and the core uncovery are obtained at 39 % and 34 % of the initial mass inventory, respectively.

• 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.

• Nuclear Engineering and Technology
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• 제21권4호
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• pp.308-320
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• 1989

이 연구는 PWR를 모의한 2루프장치에서 1차냉각재의 재고량 및 비응축성가스가 단상 및 이상 자연순환에 미치는 영향을 실험적으로 조사하고져 한 것이다. 실험장치는 U튜브를 가진 2개의 열교환기로 구성되었다. 일련의 실험을 통하여 다음 사실을 확인하였다. 이상 자연순환의 유량은 1차 냉각재 재고량의 크기에 크게 의존한다. 본 실험에서는 이상 자연순환을 유지하기 위해서는 1차 냉각재 재고량의 수위가 노즐 중심선을 유지해야 함을 알게 되었다. 비응축성 가스의 존재는 단상 자연순환을 정지시킬 수 있으며 그러나 이상 자연순환에는 큰 영향을 주지 않는다.

• 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.

• 한국전산유체공학회지
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• 제20권4호
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• pp.14-20
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• 2015

Without electirc power to cool down the hot reactor core, passive systems utilizing natural circulation are becoming a big specialty of recent neculear systems after the severe accident in Fukusima. When we consider the natural circulation in a pool, thermal mixing phenomena may start from single phase circulation and can continue to two phase condition. Since the CUPID code, which has been developed for two-phase flow analysis, can deal with the phase transition phenomena, the CUPID would be pertinent to natural convection problems in single- and two-phase conditions. Thus, the CUPID should be validated against single- and two-phase natural circulation phenomena. For the first step of the validation process, this study is focused on the validation of single-phase natural circulation. Moreover, the CUPID code solves the fluid properties by the relationship to pressure and temperature from the steam table considering non-condensable gas effects, so that the effects from variable properties are included. Simple square thermal cavity problems are tested for laminar and turbulent conditions against numerical and experimental data. Throughout the investigation, it is found that the variable properties can affect the flow field in laminar condition, but the effect becomes weak in turbulence condition, and the CUPID code implementing steam table is capable of analyzing single phase natural circualtion phenomena.

• 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 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.