• Journal of Mechanical Science and Technology
• /
• v.18 no.6
• /
• pp.990-1001
• /
• 2004

The retention of a molten pool vessel cooled by internal vessel reflooding and/or external vessel reactor cavity flooding has been considered as one of severe accident management strategies. The present numerical study investigates the effect of both internal and external vessel mixed cooling on an internally heated molten pool. The molten pool is confined in a hemispherical vessel with reference to the thermal behavior of the vessel wall. In this study, our numerical model used a scaled-down reactor vessel of a KSNP (Korea Standard Nuclear Power) reactor design of 1000 MWe (a Pressurized Water Reactor with a large and dry containment). Well-known temperature-dependent boiling heat transfer curves are applied to the internal and external vessel cooling boundaries. Radiative heat transfer has been considered in the case of dry internal vessel boundary condition. Computational results show that the external cooling vessel boundary conditions have better effectiveness than internal vessel cooling in the retention of the melt pool vessel failure.

• Proceedings of the KSME Conference
• /
• 2001.06d
• /
• pp.127-132
• /
• 2001

An experimental study on penetration integrity of the reactor vessel has been performed under external vessel cooling during a core melt accident. In this study a series of experiments are performed for the verification of the effects of coolant in the annulus between the ICI(In-Core Instrumentation) nozzle and the thimble tube and also the effects of external vessel cooling on the integrity of the penetration using the test section including only one penetration and $Al_{2}O_{3}$ melt as a corium simulant. The experimental results have shown that penetration is more damaged in the case of no external vessel cooling compared with the case of external vessel cooling. It is preliminarily concluded that the external vessel cooling is very effective measure for the improvement of the penetration integrity. Also it is confirmed from the experimental results that the coolant in the annulus reduces the melt penetration distance through the annulus and enhance the integrity of the reactor vessel penetration in the end.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.11a
• /
• pp.309-314
• /
• 1996

Feasibility study has been accomplished to evaluate the effectiveness of the in-vessel core debris cooling through lower cavity flooding using two dimensional finite difference scheme. The volume of cerium pool and decay power rate generated in corium pool were evaluated as important parameters to the temperature distribution on the reactor vessel lower head through previous works. In this study, the corium volume based on the System 80＋ core structure and time dependent decay power rate are considered for feasibility evaluation. In addition, preliminary plans for the in-vessel core debris cooling through lower cavity flooding as severe accident management strategy, i.e. flooding timing, method and capacity, are suggested based on the result of the numerical study, international tendency related to in-vessel core debris cooling through lower cavity flooding.

• Journal of Advanced Marine Engineering and Technology
• /
• v.32 no.8
• /
• pp.1157-1163
• /
• 2008

Fuel represents a significant portion of the operating costs of voyages of the vessel. Under this situation, dramatic increase in fuel price caused a problem that earnings of ship becomes decreased seriously. Pumps in a cooling system of the vessel are often operated inefficiently. The reasons will vary from process to process and application to application. Inefficiency of pumps is one of the reason for making the cost of operating the vessel increased. This paper suggests energy-effective cooling system that controls sea water pumps with inverters to reduce energy consumption in vessel by conducting simulation with LabVIEW. Results shows that electric energy consumption of pumps are significantly deceased. If this energy-effective cooling system is applied to other systems which need cooling, it could be useful in reducing electric energy wastage.

• Proceedings of the KSME Conference
• /
• 2004.04a
• /
• pp.2141-2145
• /
• 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
• /
• v.47 no.1
• /
• pp.66-73
• /
• 2015

Background: A numerical simulation was carried out to investigate the difference between internal and external heat-flux distributions at the reactor vessel wall under in-vessel retention through external reactor vessel cooling (IVR-ERVC). Methods: Total loss of feed water, station blackout, and large break loss of coolant accidents were selected as the severe accident scenarios, and a transient analysis using the element-birth-and-death technique was conducted to reflect the vessel erosion (vessel wall thickness change) effect. Results: It was found that the maximum heat flux at the focusing region was decreased at least 10% when considering the two-dimensional heat conduction at the reactor vessel wall. Conclusion: The results show that a higher thermal margin for the IVR-ERVC strategy can be achieved in the focusing region. In addition, sensitivity studies revealed that the heat flux and reactor vessel thickness are dominantly affected by the molten corium pool formation according to the accident scenario.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.17 no.2
• /
• pp.183-190
• /
• 2005

In the process of ice-slurry making, ice adhesion on cooling wall or in narrow flow Path disturbs continuous ice formation. In this study, the effect on the ice adhesion to cooling surface by some freezing experiments was investigated, quantitatively. Three types of solutions were frozen in various coating vessels with stirring. In the experiment, the ice adhesion between cooling wall and Ice-slurry was evaluated by measuring stirring power. From the result, the stirring power of slurry mixture in PTFE-coating vessel was smaller than those in PE-coating, PFA-coating and bare SUS vessel. Especially, in EG H PG 1.S/ HD 1.5 mass$\%$ solution, the stirring power in the PE-coating vessel was smaller than that in the PFA-coating or SUS vessel.

• Journal of Energy Engineering
• /
• v.28 no.4
• /
• pp.103-110
• /
• 2019

The domestic innovative power reactor named iPOWER will employ the passive molten corium cooling system(PMCCS) to cool down and stabilize the core melt in the severe accident. The final design concept of the PMCCS is yet to be determined, but the in-vessel retention through external reactor vessel cooling has been also considered as a viable strategy to cope with the severe accident. In this study, the two-phase natural circulation flow established between the reactor vessel and the insulation was simulated using a thermal-hydraulic system code, MARS-KS. The flow path of cooling water was modeled with one-dimensional nodes, and the boundary condition of the heat load from the molten core was defined to estimate the naturally-driven flow rate. The evolution of major thermal-hydraulic parameters were also evaluated, including the temperature and the level of cooling water, the void fraction around the lower head of the reactor vessel, and the heat transfer mode on its external surface.

• Nuclear Engineering and Technology
• /
• v.32 no.4
• /
• pp.379-394
• /
• 2000

As an in-vessel retention (IVR) design concept in coping with a severe accident in the nuclear power plant during which time a considerable amount of core material may melt, external cooling of the reactor vessel has been suggested to protect the lower head from overheating due to relocated material from the core. The efficiency of the ex-vessel management may be estimated by the thermal margin defined as the ratio of the critical heat flux (CHF）to the actual heat flux from the reactor vessel. Principal factors affecting the thermal margin calculation are the amount of heat to be transferred downward from the molten pool, variation of heat flux with the angular position, and the amount of removable heat by external cooling In this paper a thorough literature survey is made and relevant models and correlations are critically reviewed and applied in terms of their capabilities and uncertainties in estimating the thermal margin to potential failure of the vessel on account of the CHF Results of the thermal margin calculation are statistically treated and the associated uncertainties are quantitatively evaluated to shed light on the issues requiring further attention and study in the near term. Our results indicated a higher thermal margin at the bottom than at the top of the vessel accounting for the natural convection within the hemispherical molten debris pool in the lower plenum. The information obtained from this study will serve as the backbone in identifying the maximum heat removal capability and limitations of the IVR technology called the Cerium Attack Syndrome Immunization Structures (COASISO) being developed for next generation reactors.

• Nuclear Engineering and Technology
• /
• v.54 no.1
• /
• pp.72-83
• /
• 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.