• Journal of the Korean Society of Propulsion Engineers
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• v.23 no.5
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• pp.43-52
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• 2019

Cryogenic propellant rockets utilize a natural circulation loop of cryogenic fluid to cool the engine inlet temperature before launch. The geometric information about the circulation system, such as length and diameter of the pipes and the heat input to the system, defines the mass flow rate of the natural circulation loop. We performed experiments to verify the natural circulation mass flow rate and compared the results with the analytical results. The comparison of the mass flow rate between experiments and numerical simulations showed a 12% offset. We also included a prediction of the natural circulation flow rate in the low-gravity section and in the acceleration section in the upper stage of the launch vehicle. The oxygen tank should have 100 kPa(a) of pressure in the acceleration section to maintain a high flow rate for the natural circulation loop. In the low-gravity section, there should be an optimal tank pressure that leads to the maximum natural circulation flow rate.

• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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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.

• Progress in Superconductivity and Cryogenics
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• v.15 no.3
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• pp.49-52
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• 2013

An experiment to investigate the natural circulation of a cryogen has been performed. The study is motivated mainly by our recent development of cryogenic cooling system for prototype superconducting cyclotron without any circulating pump. In the natural circulation loop system, a cooling channel is attached on the outer surface of the aluminium block and the liquid nitrogen passes through inside of the channel to cool the block indirectly. A cryocooler as a heat sink is located at the top to re-condense cryogenic vapor coming from the aluminium block in which electrical heater is installed as a heat source. The main dimensions are determined using the relevant analysis and the natural circulation loop is successfully fabricated. The temperature distributions in the loop are measured during initial cool-down process and in steady state, from which the modified Grashof numbers are calculated and compared with the existing correlation estimated with one-dimensional analysis for steady state flow.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.308-320
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• 1989

The objective of this work is to investigate the effects of diminished primary coolant inventory and the presence of noncondensible gas during single- and two-phase natural circulation in a PWR loop model. The test model was composed of two loops with a U-tube heat exchanger in each loop. Through a series of tests, it has been confirmed that the two-phase natural circulation flow rates were greatly dependent on primary coolant inventory as previous investigators observed. The primary coolant inventory limit to maintain two-phase natural circulation was found to be the amount of the coolant necessary to keep the waterline of the coolant nozzle hole center in this model. The presence of noncondensible gas impede the single-phase natural circulation, but it did not affect the two-phase natural circulation significantly.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.772-777
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• 1997

Previous MELCOR and SCDAP/RELAP5 nodalizations for simulating the counter-current, natural circulation behavior of vapor flow within the RCS hot legs and SG U-tubes when core damage progress can not be applied to the steady state and water-filled conditions during the initial period of accident progression because of the artificially high loss coefficients in the hot legs and SG U-tubes which were chosen from results of COMMIX calculation and the Westinghouse natural circulation experiments in a 1/7-scale facility for simulating steam natural circulation behavior in the vessel and in the hot leg and SG during the TMLB' scenrio. The objective of this study is to develop a natural circulation modeling which can be used both for the liquid flow condition at steady state and for the vapor flow condition at the later period of in-vessel core damage. For this, the drag forces resulting from the momentum exchange effects between the two vapor streams in the hot leg was modeled as a pressure drop by pump model. This hot leg natural circulation modeling of MELCOR was able to reproduce similar mass flow rates with those predicted by previous models.

• Nuclear Engineering and Technology
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• v.54 no.1
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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.

• Journal of computational fluids engineering
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• v.20 no.4
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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.

• Transactions of the Korean Society of Mechanical Engineers
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• v.17 no.5
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• pp.1313-1320
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• 1993

Flow patterns inside the riser section and the effects of the heater inlet-and exit-restrictions, liquid charging level and the heater inlet subcooling on the flow characteristics inside an open two-phase natural circulation loop were studied experimentally. Three basic circulation modes were observed ; periodic circulation (A)(flow oscillations with incubation(no boiling) period), continuous circulations(stable operation mode with no flow oscillations), and periodic circulation (B) (flow oscillations with continuous boiling). The circulation rate increases and then decreases with the increase of the heating rate and the maximum circulation rate appears with the continuous circulation mode. The decrease of the inlet-restriction or the increase of the exitrestriction destabilizes the system. When the liquid charging level or the inlet subcooling decreases, the continuous circulation mode starts at the lower heating rate and the system is stabilized.

• Nuclear Engineering and Technology
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• v.53 no.12
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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.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.721-733
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• 2020

In nuclear engineering, the occurrence of critical heat flux (CHF) is complicated for rod bundle, and it is much more difficult to predict the CHF when it is in natural circulation under motion condition. In this paper, the dryout-type CHF is investigated for the rod bundle in a natural circulation loop under rolling motion condition based on the coupled analysis of subchannel method, a one-dimensional system analysis method and a CHF mechanism model, namely the three-fluid model for annular flow. In order to consider the rolling effect of the natural circulation loop, the subchannel model is connected to the one-dimensional system code at the inlet and outlet of the rod bundle. The subchannel analysis provides the local thermal hydraulic parameters as input for the CHF mechanism model to calculate the occurrence of CHF. The rolling motion is modeled by additional motion forces in the momentum equation. First, the calculation methods of the natural circulation and CHF are validated by a published natural circulation experiment data and a CHF empirical correlation, respectively. Then, the CHF of the rod bundle in a natural circulation loop under both the stationary and rolling motion condition is predicted and analyzed. According to the calculation results, CHF under stationary condition is smaller than that under rolling motion condition. Besides, the CHF decreases with the increase of the rolling period and angular acceleration amplitude within the range of inlet subcooling and mass flux adopted in the current research. This paper can provide useful information for the prediction of CHF in natural circulation under motion condition, which is important for the nuclear reactor design improvement and safety analysis.