• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.11
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• pp.2866-2874
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• 2000

This paper is concerned with the optimum design for external reinforcement of a nuclear reactor pressure vessel(RPV) in a severe accident. During the severe reactor accident of molten core, the temperature and the pressure in the nuclear reactor rise to a certain level depending on the initial and subsequent condition of a severs accident. The reis of the temperature and the internal pressure cause deterioration of the load carrying capacity and could cause failure of the RPV lower head. The deterioration of failure can be mitigated by the external cooling or the reinforcement of the lower head with additional structures. While the external cooling forces the temperature of an RPV to drop to the desired level, the reinforcement of the lower head can attain both the increase of the load carrying capacity and the temperature drop. The reinforcement of the lower head can be optimized to have the maximum effect on the collapse pressure and the temperature at the inner wall. Optimization results are compared to both the result without the reinforcement and the result with the designated reinforcement.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.782-787
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• 2001

This paper is concerned with the global rupture of a nuclear reactor pressure vessel(RPV) in a severe accident. During the severe reactor accident of molten core, the temperature and the pressure in the nuclear reactor rise to a certain level depending on the initial and subsequent condition of a severe accident. While the rise of the temperature cause the thermal softening of RPV material, the rise of the internal pressure could cause failure of the RPV lower head. The global rupture of an RPV is simulated by finite element limit analysis for the collapse pressure and mode and this analysis results have been compared with a variation of the internal pressure of RPV. The finite element limit method is a systematic tool to secure the safety criteria of a nuclear reactor and to evaluate the in-vessel corium retention.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.713-718
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• 1997

In the TMI-2 accident, approximately twenty(20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However, one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100$^{\circ}C$ for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant(KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options.

• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.299-308
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• 2000

A safety assessment of reactor vessel lower head integrity under in-vessel vapor explosion loads has been performed. The core melt relocation parameters were chosen within the ranges of physically realizable bounds. The premixing and explosion calculations were performed using TRACER-II code. Using the calculated explosion pressures imposed on the lower head inner wall, strain calculations were peformed using ANSYS code. Then, the calculated strain results and the established failure criteria were used in determining the failure probability of the lower head, In the explosion analyses, it is shown that the explosion impulses are not altered significantly by the uncertain parameters of triggering location and time, fuel and vapor volume fractions in uniform premixture bounding calculations. Strain analyses show that the vapor explosion-induced lower head failure is not possible under the present framework of assessment. The result of static analysis using the conservative explosion-end pressure of 50 MPa also supports the conclusion. It is recommended, however, that an assessment of fracture mechanics for preexisting cracks be also considered to obtain a more concrete conclusion.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.707-712
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• 1997

This paper presents the current state of knowledge on molten material relocation into the lower plenum. Consequences of movement of material to the lower head are considered with regardt to the potential for reactor pressure vessel failure from both thermal hydraulic and mechanical standpoints. The models are applied to evaluating various in-vessel retention strategies for the Korean Standard power plant (KSNPP) reactor The results are summarized in terms of thermal response of the reactor vessel from the very relevant severe accident management perspective.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.473-478
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• 1997

A severe accident management strategy, in-vessel retention corium through external reactor vessel cooling(ERVC) is being studied worldwide as a means to prevent reactor vessel failure following a core melt accident. An evaluation of feasibility of this ERVC for a large Advanced Light Water Reactor (ALWR) is presented. To account for the coolability of corium and metal in the reactor vessel, a thermal analysis is performed using an existing method. Results show that the peak heat flux along the inner surface of the reactor vessel lower head has a relatively smaller margin than a small capacity reactor such as AP600 in regards with the critical heat flux attainable at the outer surface of the reactor vessel lower head.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4522-4533
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• 2022

For severe accident assessment of reactor pressure vessel (RPV), it is important to develop an accurate model that can predict transient thermo-mechanical behavior of the RPV lower head under the given condition. The present study revisits the lower head failure with two- and three-dimensional finite element models. In particular, we aim to give clear insight regarding the effect of the three-dimensionality present in the distribution of the thickness and thermal load of the lower head. For a rigorous validation of the result, both the OLHF-1 and the OLHF-2 tests are considered in this study. The result suggests that the three-dimensional effect is not negligible as far as the failure location is concerned. The non-uniformity of the thickness distribution is found to affect the failure location and time. The thermal load, which may not be axisymmetric in general, has the most significant effect on the failure assessment. We also observe that the creep property can affect the global deformation of the lower head, depending on the applied mechanical load.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.237-248
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• 2012

Corium behavior in the lower plenum of the reactor vessel during a severe accident is very important, as this affects a failure mechanism of the lower head vessel and a thermal load to the outer reactor vessel under the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) condition. This paper discusses the state of the art and technical issues on corium behavior in the lower plenum, such as initial corium pool formation characteristics and its transient behavior, natural convection heat transfer in various geometries, natural convection heat transfer with a phase change of melting and solidification, and corium interaction with a lower head vessel including penetrations of the ICI (In-Core Instrumentation) nozzle are discussed. It is recommended that more detailed analysis and experiments are necessary to solve the uncertainties of corium behavior in the lower plenum of the reactor vessel.

• 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 Nuclear Society Conference
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• 1999.05a
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• pp.209-209
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• 1999