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

• Nuclear Engineering and Technology
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• v.28 no.6
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• pp.576-582
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• 1996

The integrity of reactor vessel support system of the Korean Standard Nuclear Power Plant (KSNPP) is investigated for a postulated reactor vessel closure head drop event. The closure head is disassembled from the reactor vessel during refueling process or general inspection of reactor vessel and internal structures, and carried to proposed location by the head lift rig. A postulated closure head drop event could be anticipated during closure head handling process. The drop event may cause an impact load on the reactor vessel and supporting system. The integrity of the supporting system is directly relevant to that of reactor vessel and reactor internals including fuels. Results derived by elastic impact analysis, linear and non-linear buckling analysis and elasto-plastic stress analysis of the supporting system implied that the integrity of the reactor vessel supporting system is intact for a postulated reactor vessel closure head drop event.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.66-73
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• 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.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.3017-3029
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• 2023

In the late in-vessel phase of a nuclear reactor severe accident, the internal heat transfer and crust evolution during the debris bed melting process have important effects on the thermal load distribution along the vessel wall, and further affect the reactor pressure vessel (RPV) failure mode and the state of melt during leakage. This study coupled the phase change model and large eddy simulation to investigate the variations of the temperature, melt liquid fraction, crust and heat flux distributions during the debris bed melting process in the hypothetical severe accident of HPR1000. The results indicated that the heat flow towards the vessel wall and upper surface were similar at the beginning stage of debris melting, but the upward heat flow increased significantly as the development of the molten pool. The maximum heat flux towards the vessel wall reached 0.4 MW/m². The thickness of lower crust decreased as the debris melting. It was much thicker at the bottom region with the azimuthal angle below 20° and decreased rapidly at the azimuthal angle around 20-50°. The maximum and minimum thicknesses were 2 and 90 mm, respectively. By contrast, the distribution of upper crust was uniform and reached stable state much earlier than the lower crust, with the thickness of about 10 mm. Moreover, the sensitivity analysis of initial condition indicated that as the decrease of time interval from reactor scram to debris bed dried-out, the maximum debris temperature and melt fraction became larger, the lower crust thickness became thinner, but the upper crust had no significant change. The sensitivity analysis of in-vessel retention (IVR) strategies indicated that the passive and active external reactor vessel cooling (ERVC) had little effect on the internal heat transfer and crust evolution. In the case not considering the internal reactor vessel cooling (IRVC), the upper crust was not obvious.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.27 no.3
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• pp.191-198
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• 2014

In this paper, structural integrity evaluation of reactor pressure vessel bottom head without penetration nozzles in core melting accident has been performed. Considering the analysis results of thermal load, weight of molten core debris and internal pressure, thermal load is the most significant factor in reactor vessel bottom head. The failure probability was evaluated according to the established failure criteria and the evaluation showed that the equivalent plastic strain results are lower than critical strain failure criteria. Thermal-structural coupled analyses show that the existence of elastic zone with a lower stress than yield strength is in the middle of bottom head thickness. As a result of analysis, the elastic zone became narrow and moved to the internal wall as the internal pressure increases, and it is evaluated that the structural integrity of reactor vessel is maintained under core melting accident.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2006.10a
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• pp.33-34
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• 2006

• Journal of the Korean Society of Systems Engineering
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• v.13 no.1
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• pp.15-23
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• 2017

This paper describes verification of APR1400 reactor vessel by applying the system engineering approach, in which the design re-engineering method is used to check the design parameters of APR1400 RV (reactor vessel). The RV is classified as safety class 1 and therefore must adhere strictly to the rules of ASME BPVC section III, subsection NB and seismic category I. This study explores designing the RV by following the ASME guidelines and making a comparative study with the current design. To meet this objective we apply system engineering methodologies to structure the process and allow for verification and validation of the major RV design parameters such as thickness of RV. The structural thicknesses of various part of RV are determined as well as reinforcements on the RV major nozzles. A 3D virtual reality model was created based on the design parameters using CATIA V5 and animation using Dassault Composer V2016. A comparison of re-engineered ARP1400 RV and standard APR1400 RV was done to show which design parameters were taken more conservative approach.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2004.10a
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• pp.1327-1330
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• 2004

An underwater robotic system has been developed and applied to visual inspection of reactor vessel internals. The Korea Electric Power Robot for Visual Test (KeproVt) consists of an underwater robot, a vision processor-based measuring unit, a master control station and a servo control station. The robot guided by the control station with the measuring unit can be controlled to have any motion at any position in the reactor vessel with $\pm$1 cm positioning and $\pm$2 degrees heading accuracies with enough precision to inspect reactor internals. A simple and fast installation process is emphasized in the developed system. The developed robotic system was successfully deployed at the Younggwang Nuclear Unit 1 for the visual inspection of reactor internals.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.28-36
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• 1993

Much effort is being done to standardize the PWR-type nuclear power plant in Korea. This paper presents the development of seismic design criteria for the reactor internals as a part of the standardization program for nuclear power plant. The seismic design loads of the reactor internals are calculated using the reference input motions of reactor vessel taken from Yong-gwang Nuclear Power Plant Units 3 and 4. An overview of analysis related to the basic parameters and methodologies is presented. Also, the response of internal components for the reactor vessel motions is carefully investigated.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.996-1007
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• 2019

We investigated the influence of the aspect ratio (H/R) of the oxide layer on the reactor vessel heating in three-layer configuration. Based on the analogy between heat and mass transfers, we performed mass transfer experiments to achieve high Rayleigh numbers ranging from $6.70{\times}10^{10}$ to $7.84{\times}10^{12}$. Two-dimensional (2-D) semi-circular apparatuses having the internal heat source were used whose surfaces of top, bottom and side simulate the interfaces of the oxide layer with the light metal layer, the heavy metal layer, and the reactor vessel, respectively. Multi-cell flow pattern was identified when the H/R was reduced to 0.47 or less, which promoted the downward heat transfer from the oxide layer and possibly mitigated the focusing effect at the upper metallic layer. The top boundary condition greatly affected the natural convection of the oxide layer due to the presence of secondary flows underneath the cold light metal layer.