• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.19-31
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• 1994

The objective of this study is to conduct a thermal-hydraulic analysis on the spent fuel pool and to evaluate a parametric effect for the thermal-hydraulic analysis of spent fuel pool. The selected parameters are the Reynolds Number and the gap flow through the oater gap between fuel cell and fuel bundle. The simplified flow network for a path of fuel cells is used to analyze the natural circulation phenomenon. In the flow network analysis, the pressure drop for each assembly from the entrance of the fuel rack to the exit of the fuel assembly is balanced by the driving head due to the density difference between the pool fluid and the average fluid in each spent fuel assembly. The governing equations ore developed using this relation. But, since the parameters(flow rate, pressure loss coefficient, decay heat, density)are coupled each other, iteration method is used to obtain the solution. For the analysis of the YGN 3&4 spent fuel rack, 12 channels are considered and the inputs such as decay heat and pressure loss coefficient are determined conservatively. The results show the thermal-hydraulic characteristics(void fraction, density, boiling height)of the YGN 3&4 spent fuel rack. There occurs small amount of boiling in the cells. Fuel cladding temperature is lower than 343.3$^{\circ}C$. The evaluation of parametric effect indicates that flow resistances by geometric effect are very sensitive to Reynolds number in the transition region and the gap flow is negligible because of the larger flow resistance in the gap flow path than in the fuel bundle.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.605-612
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• 2013

Removal of decay heat from an operating reactor during a prolonged station blackout condition is a big concern for reactor designers, especially after the recent Fukushima accident. In the case of a prolonged station blackout condition, heat removal is possible only by passive means since no pumps or active systems are available. Keeping this in mind, the AHWR has been designed with many passive safety features. One of them is a passive means of removing decay heat with the help of Isolation Condensers (ICs) which are submerged in a big water pool called the Gravity Driven Water Pool (GDWP). The ICs have many tubes in which the steam, generated by the reactor core due to the decay heat, flows and condenses by rejecting the heat into the water pool. After condensation, the condensate falls back into the steam drum of the reactor. The GDWP tank holds a large amount of water, about 8000 $m^3$, which is located at a higher elevation than the steam drum of the reactor in order to promote natural circulation. Due to the recent Fukushima type accidents, it has been a concern to understand and evaluate the capability of the ICs to remove decay heat for a prolonged period without escalating fuel sheath temperature. In view of this, an analysis has been performed for decay heat removal characteristics over several days of an AHWR by ICs. The computer code RELAP5/MOD3.2 was used for this purpose. Results indicate that the ICs can remove the decay heat for more than 10 days without causing any bulk boiling in the GDWP. After that, decay heat can be removed for more than 40 days by boiling off the pool inventory. The pressure inside the containment does not exceed the design pressure even after 10 days by condensation of steam generated from the GDWP on the walls of containment and on the Passive Containment Cooling System (PCCS) tubes. If venting is carried out after this period, the decay heat can be removed for more than 50 days without exceeding the design limits.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.22 no.7
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• pp.980-991
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• 1998

In this study, nucleate pool boiling heat transfer coefficients of alternative refrigerants on a plain, low fin, and two enhanced tubes were measured and compared against each other. To obtain data at conditions similar to the actual evaporator, a fluid heating method was employed instead of an electric heating method in the experiments. R123, R134a, R22 and R410a were used as working fluids and data were taken at 7 deg.C ar heat fluxes of 20 ~ 100 kW/m$\^$2/. Comparison of the plain tube data against some correlations showed that the simplest correlation of Cooper based on reduced pressure predicted the data for all fluids tested with a 10% deviation. For all refrigerants, enhanced tubes composed of subsurface and subtunnels, especially Thermoexcel-E tube, showed the highest heat transfer coefficients among the tubes tested with one exception that the low fin tube's performance was better than those of enhanced tubes for high vapor pressure fluid such as R410a at high heat flux. Finally, a low fin and enhanced tubes showed higher heat transfer enhancement for low vapor pressure of R123 than for high vapor pressure fluisd. For R123, the enhancement factors for Turbo-B and Thermoexcel-E tubes were 2.8 ~ 4.8 and 4.6 ~ 8.1 respectively.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.627-638
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• 2018

This article describes an integrated Level 1-Level 2 probabilistic risk assessment (PRA) methodology to evaluate the radiological risk during postulated accident scenarios initiated during the decommissioning phase of a typical Mark I containment boiling water reactor. The fuel damage scenarios include those initiated while the reactor is permanently shut down, defueled, and the spent fuel is located into the spent fuel storage pool. This article focuses on the integrated Level 1-Level 2 PRA aspects of the analysis, from the beginning of the accident to the radiological release into the environment. The integrated Level 1-Level 2 decommissioning PRA uses event trees and fault trees that assess the accident progression until and after fuel damage. Detailed deterministic severe accident analyses are performed to support the fault tree/event tree development and to provide source term information for the various pieces of the Level 1-Level 2 model. Source terms information is collected from accidents occurring in both the reactor pressure vessel and the spent fuel pool, including simultaneous accidents. The Level 1-Level 2 PRA model evaluates the temporal and physical changes in plant conditions including consideration of major uncertainties. The goal of this article is to provide a methodology framework to perform a decommissioning Probabilistic Risk Assessment (PRA), and an application to a real case study is provided to show the use of the methodology. Results will be derived from the integrated Level 1-Level 2 decommissioning PSA event tree in terms of fuel damage frequency, large release frequency, and large early release frequency, including uncertainties.

• Journal of Energy Engineering
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• v.2 no.2
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• pp.179-186
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• 1993

Heat pipes. applied to a flat plate solar collector, have a long and slender configuration with relatively low heat flux in the evaporator section. Such a heat pipe has a tendency to build-up a liquid pool at the lower part of the evaporator section. and at this pool occurs such complicated phenomena of evaporation and fluid dynamics as superheat, sudden generation of bubble, its likely explosive growth process and flooding, etc. In the present study. we tried to solve these problems by means of adjusting two principal design parameters, the liquid inventory and the installation region of the wick, using 4 heat pipes and 3 thermospheres. The corresponding results can be summarized as follows$\^$1)/. The effective thermal conductances of the heat pipe was greatly improved by eliminating the wick in the adiabatic and condenser sections$\^$2)/. The liquid inventory should be increased by about 40% larger than what is saturated the wick$\^$3)/. In the evaporator section the wick has a favorable effect to reduce both unstable operation by intermittent occurrence of nucleate boiling and response time at the initial start-up process.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.12 no.7
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• pp.642-648
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• 2000

Falling film heat transfer analyses with aqueous lithium bromide solution were peformed to investigate the transfer characteristics of the copper tubes. Finned(knurled) tube and a smooth tube were selected as test specimens. Averaged generation fluxes of water and the heat transfer performances(heat flux, heat transfer coefficient) were obtained. The results of this work were compared with the data reported previously. As the film flow rate of the solution increased, the generation fluxes of water decreased for both tubes. The reason is estimated by the fact that the heat transfer resistance with the film thickness increased as the film flow rate increased. The effect of the enlarged surface area at the knurled tube was supposed to be dominant at a small flow rate. The generation fluxes of water increased with the increasing degree of tube wall superheat. Nucleate boiling is supposed to occur at a wall superheat of 20 K for a smooth tube, and at 10 K for a knurled tube. The heat transfer performance of the falling film was superior to pool boiling at a low wall superheat below 10 K for both tubes tested. The knurled tube geometry showed good performance than the smooth tube, and the increased performance was mainly came from the effect of the increased heating surface area.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1845-1850
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• 2022

In-vessel corium retention (IVR) during external reactor vessel cooling (ERVC) is a key severe accident management strategy adopted in advanced nuclear power plants. The injection of nanofluids has been regarded as a means of enhancing CHF when using the IVR-ERVC strategy to safeguard high-power nuclear reactors. However, a critical practical concern is that various types of debris flowing from the contaminant sump during operation of an ERVC system might degrade CHF enhancement by nanofluids. Our objective here was to experimentally assess the viability of nanofluid use to enhance CHF in practical ERVC contexts (e.g., when fluids contain various types of debris). The types and characteristics of debris expected during IVR-ERVC were examined. We performed pool boiling CHF experiments using nanofluids containing these types of debris. Notably, we found that debris did not cause any degradation of the CHF enhancement characteristics of nanofluids. The nanoparticles are approximately 1000-fold smaller than the debris particles; the number of nanoparticles in the same volume fraction is 1 billion-fold greater. Nanofluids increase CHF via porous deposition of nanosized particles on the boiling surface; this is not hindered by extremely large debris particles.

• Journal of Energy Engineering
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• v.12 no.2
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• pp.139-144
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• 2003

Augmentation of CHF by ultrasonic vibration in water pool is experimentally investigated under pool boiling condition. The experiments are carried out using copper coated plates and distilled water. Measurements of CHF on flat plate heated surface were made with and without ultrasonic wave and with variations in inclined angle of the surface and water subcooling. Experimental apparatus consists of a bath, power supply, test section, ultrasonic generator, and data acquisition system. The measurements show that ultrasonic wave enhances CHF and its extent is dependent upon inclination angle as well as water subcooling. The rate of increase in CHF increases with an increase in water subcooling while it decreases with an increase in inclination angle. Visual observation shows that the cause of CHF augmentation is closely related with the dynamic behavior of bubble generation and departure in acoustic field.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.3175-3180
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• 2008

A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.27-36
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• 2020

Natural convection and thermal stratification phenomena are found in large water pools that are being used as heat sinks for decay heat removal from the reactor core using passive heat removal systems. In this study, the two-phase (water and air) natural convection and thermal stratification phenomena with conjugate heat transfer in the rectangular enclosure were investigated numerically using ANSYS Fluent 17.2 code. The transient numerical simulations of these phenomena in the full-scale computational domain of the experimental facility were performed. Generation of water vapour bubbles around the heater rod and evaporation phenomena were included in this numerical investigation. The results of numerical simulations are in good agreement with experimental measurements. This shows that the natural convection is formed in region above the heater rod and the water is thermally stratified in the region below the heater rod. The heat from higher region and from the heater rod is transferred to the lower region via conduction. The thermal stratification disappears and the water becomes well mixed, only after the water temperature reaches the saturation temperature and boiling starts. The developed modelling approach and obtained results provide guidelines for numerical investigations of thermal-hydraulic processes in the water pools for passive residual heat removal systems or spent nuclear fuel pools considering the concreate walls of the pool and main room above the pool.