• Nuclear Engineering and Technology
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• 제52권1호
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• pp.185-191
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• 2020

Plant-specific analyses of an advanced reactor have been performed to assure the structural integrity of the reactor pressure vessel during transient conditions, which are expected to initiate pressurized thermal shock (PTS) events. The vessel failure probabilities from the probabilistic fracture mechanics analyses are combined with the transient frequencies to generate the through-wall cracking frequencies, which are compared to the acceptance criterion. Several sensitivity analyses are performed, focusing on the orientations and sizes of cracks, the copper content, and a flaw distribution model. The results show that the integrity of the reactor vessel is expected to be maintained for long-term operation beyond the design lifetime from the PTS perspective using the design data of the advanced reactor. Moreover, a fluence level exceeding 9×10¹⁹ n/㎠ is found to be acceptable, generating a sufficient margin beyond the design lifetime.

• Journal of Mechanical Science and Technology
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• 제18권6호
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.309-314
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• 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.

• Nuclear Engineering and Technology
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• 제38권2호
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• pp.139-146
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• 2006

An experimental study was performed to evaluate the effects of surface coating and an enhanced insulation structure on the downward facing boiling process and the critical heat flux on the outer surface of a hemispherical vessel. Steady-state boiling tests were conducted in the Subscale Boundary Layer Boiling (SBLB) facility using an enhanced vessel/insulation design for the cases with and without vessel coatings. Based on the boiling data, CHF correlations were obtained for both plain and coated vessels. It was found that the nucleate boiling rates and the local CHF limits for the case with micro-porous layer coating were consistently higher than those values for a plain vessel at the same angular location. The enhancement in the local CHF limits and nucleate boiling rates was mainly due to the micro-porous layer coating that increased the local liquid supply rate toward the vaporization sites on the vessel surface. For the case with thermal insulation, the local CHF limit tended to increase from the bottom center at first, then decrease toward the minimum gap location, and finally increase toward the equator. This non-monotonic behavior, which differed significantly from the case without thermal insulation, was evidently due to the local variation of the two-phase motions in the annular channel between the test vessel and the insulation structure.

• Nuclear Engineering and Technology
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• 제54권1호
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.194-194
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• 1998

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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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.

• Nuclear Engineering and Technology
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• 제30권2호
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• pp.164-172
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• 1998

This paper presents the detailed modelling of reactor vessel ledge region for the dynamic analysis of the coupled internals and core model. The dynamic responses due to earthquake and pipe break are calculated using the input motions of reactor vessel taken from Ulchin nuclear power plant units 3 and 4. Two different representations for detailed and simplified models of the RV ledge region are made. The dynamic responses of the reactor internals components are compared between them. Response characteristics are reported and simplified model is suggested for earthquake and pipe break analysis for the future design of the reactor internals.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.351-361
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• 2015

External reactor vessel cooling (ERVC) for in-vessel retention (IVR) has been considered one of the most useful strategies to mitigate severe accidents. However, reliability of this common idea is weakened because many studies were focused on critical heat flux whereas there were diverse uncertainties in structural behaviors as well as thermal-hydraulic phenomena. In the present study, several key factors related to molten corium behaviors and thermal characteristics were examined under multi-layered corium formation conditions. Thereafter, systematic finite element analyses and subsequent damage evaluation with varying parameters were performed on a representative reactor pressure vessel (RPV) to figure out the possibility of high temperature induced failures. From the sensitivity analyses, it was proven that the reactor cavity should be flooded up to the top of the metal layer at least for successful accomplishment of the IVR-ERVC strategy. The thermal flux due to corium formation and the relocation time were also identified as crucial parameters. Moreover, three-layered corium formation conditions led to higher maximum von Mises stress values and consequently shorter creep rupture times as well as higher damage factors of the RPV than those obtained from two-layered conditions.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1871-1880
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• 2020

Round robin analyses for vessel failure probabilities due to PTS events are proposed for plant-specific analyses of all types of reactors developed in Korea. Four organizations, that are responsible for regulation, operation, research and design of the nuclear power plant in Korea, participated in the round robin analysis. The vessel failure probabilities from the probabilistic fracture mechanics analyses are calculated to assure the structural integrity of the reactor pressure vessel during transients that are expected to initiate PTS events. The failure probabilities due to various parameters are compared with each other. All results are obtained based on several assumptions about material properties, flaw distribution data, and transient data such as pressure, temperature, and heat transfer coefficient. The realistic input data can be used to obtain more realistic failure probabilities. The various results presented in this study will be helpful not only for benchmark calculations, result comparisons, and verification of PFM codes developed but also as a contribution to knowledge management for the future generation.