Core Melt Stabilization Concepts for Existing and Future LWRs and Associated Research and Development Needs

Fischer, M.; Bechta, Sevostian; Bezlepkin, V. V.; Hamazaki, Ryoichi; Miassoedov, A.

doi:10.13182/nt16-19

Cited by 12 publications

(5 citation statements)

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“…In the EVR strategy, the surface-to-volume ratio of corium must be maximized to increase coolability. In the EPR core catcher concept (Fischer, et al, 2016), and as a backfit for some Gen 2 reactors (Bonnet, et al, 2016), spreading is a means to reduce the melt thickness and thus the average heat fluxes to concrete, as shown Figure 15. Spreading is, indeed, controlled by the melt properties, mainly kinematic viscosity (Journeau, et al, 2003).…”

Section: Discussion About the Influence On Ex-vessel Retention Strate...mentioning

confidence: 99%

“…During a severe accident in a nuclear reactor, the molten core-or corium-may be relocated into the reactor vessel's lower plenum in case of core support plate failure. Different Severe Accident Management (SAM) strategies have been proposed for the current fleet and new-built reactors in which the molten core retention is to be achieved either in-vessel or ex-vessel (Fischer, et al, 2016;Amidu, et al, 2022). The severe accident management strategy for IVR consists in stabilizing the corium within the Reactor Pressure Vessel (RPV) by external cooling of the vessel's lower head (Ma, et al, 2016).…”

Section: Introductionmentioning

confidence: 99%

“…In the case of EVR through the Concrete Cavity [EVRCC, (Amidu, et al, 2022)], there are two options depending whether the cavity is filled or not by water at the time of the vessel melt-through. In the case of a wet cavity (Fischer, et al, 2016), the corium jet fragments, a process largely controlled by surface tension (Thakre, 2015), and forms a debris bed that will be cooled by the surrounding water. In addition, surface tension will also play an important role, for example, in oxide-metal ex-vessel configurations (stratification and emulsion).…”

Section: Introductionmentioning

confidence: 99%

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High-Temperature Characterization of Melted Nuclear Core Materials: Investigating Corium Properties Through the Case Studies of In-Vessel and Ex-Vessel Retention

2022

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During a severe accident in a nuclear reactor, the molten core—or corium—may be relocated into the reactor vessel’s lower plenum in case of core support plate failure. The severe accident management strategy for In-Vessel Retention—or IVR—consists in stabilizing the corium within the reactor pressure vessel by external cooling of the vessel’s lower head. If now, the vessel fails due to excessive thermal loading on its walls, the Ex-Vessel Retention—or EVR—strategy is adopted. In this case, the core melt stabilization can be achieved by effective corium spreading, either in the reactor vessel cavity or in a dedicated “core-catcher”, and cooling by water. The success of both strategies highly depends on the corium behavior at high temperatures, conditioning vessel’s integrity for IVR, and promotion for the spreading of the EVR. This involves a variety of fundamental mechanisms closely related to heat and mass transfer regimes prevailing at the system scale, which requires further analytical and experimental insight to determine the primary mechanisms and feed the modeling tools, allowing the numerical simulations of severe accident scenarios.Within the framework of corium characterization at high temperatures, the present study aims at filling the lack of such fundamental data as density, surface tension, liquidus and solidus temperatures, and viscosity. In order to accurately measure these properties at high temperatures, the VITI facility is designed with various configurations. Concerning IVR, the influence of density and surface tension is particularly highlighted through VITI-SD and VITI-MBP configurations, and practical applications of experimental results are finally discussed, in link with the focusing effect issue at the thin upper metallic layer of the corium pool. Concerning EVR, the properties of interest are solidus/liquidus temperature and dynamic viscosity, and typical experimental results obtained through VITI-VPA and VITI-GFL configurations are discussed in view of characterizing corium spreading.

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Section: Discussion About the Influence On Ex-vessel Retention Strate...mentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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High-Temperature Characterization of Melted Nuclear Core Materials: Investigating Corium Properties Through the Case Studies of In-Vessel and Ex-Vessel Retention

2022

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“…The combination of engineering judgment and probabilistic methods is used to determine the prevention and mitigation measures. Some examples of mitigation strategies for Gen-III concepts of reactors may be found for in-vessel [24,25] or ex-vessel retention [26]. For Gen-IV concepts, the in-vessel retention with internal core catcher is considered as mitigation strategy, together with devices aiming at insuring fuel discharge from the core zone.…”

Section: Severe Accident Management Strategiesmentioning

confidence: 99%

A Comparative Study on Severe Accident Phenomena Related to Melt Progression in Sodium Fast Reactors and Pressurized Water Reactors

Bachrata

Bertrand

Marie

et al. 2020

Journal of Nuclear Engineering and Radiation Science

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The nuclear safety approach has to cover accident sequences involving core degradation in order to develop reliable mitigation strategies for both existing and future reactors. In particular, the long-term stabilization of the degraded core materials and their coolability has to be ensured after a severe accident. This paper focuses on severe accident phenomena in Pressurized Water Reactors (PWR) compared to those potentially occurring in future GenIV-type Sodium Fast Reactors (SFR). Firstly, the two considered reactor concepts are introduced by focusing on safety aspects. The severe accident scenarios leading to core melting are presented and the initiating events are highlighted. The paper focuses on in-vessel severe accident phenomena, including the chronology of core damage, major changes in the core configuration and molten core progression. Regarding the mitigation means, the in-vessel retention phenomena and the core catcher characteristics are reviewed for these different nuclear generation concepts (II, III and IV). A comparison between the PWR and SFR severe accident evolution is provided as well as the relation between governing physical parameters and the adopted mitigation provisions for each reactor concept. Finally, it is highlighted how the robustness of the safety demonstration is established by means of a combined probabilistic and deterministic approach.

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“…To mitigate these consequences, an ex-vessel core catcher is configured in the design, which can retain and cool the molten corium for a prolonged period of time and eliminate the MCCI. 2,3 Fishcher et al 4 explained the modes of interaction of molten corium and its coolability by bottom flooding for European Union PWR (EU-PWR). To validate the different design aspects of the core catcher, analytical and experimental studies were conducted by them.…”

Section: Introductionmentioning

confidence: 99%

Experimental investigations on melt coolability with simulated decay heat—The influence of delay time in flooding

2021

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The primary objective of core catchers is to minimize the potential risk of core meltdown by stabilization and cooling of molten corium within the reactor containment for a prolonged period by strategically cooling it. Quenching of the high-temperature melt by top flooding strongly depends on the timing of injection of the cooling water, that is, immediate flooding or delayed flooding. To understand this aspect, we conducted experiments in a simulated core catcher of an Indian advanced nuclear reactor. In the experiments, sodium borosilicate glass, as a mixture of corium and sacrificial material simulant, was melted at about 1200°C and cooled by water from the side of a simulated core catcher and then flooding at the top of the melt pool. Two experiments were conducted; the first one was done with immediate flooding at the top of the melt pool; the second was performed with delayed top flooding of 45 min gap. The results show that immediate flooding quenches the melt pool rapidly compared to delayed flooding. A stable crust formation at the top of the melt pool and wall adhesion is observed in the delayed flooding. It was also found that no ingression of the water in the molten pool occurred with delayed flooding. On the other hand, with immediate flooding, water was found to ingress into the melt pool due to gap formations between melt and core catcher sidewall, causing the melt to break up through eruption by ingressed water. This phenomenon resulted in the formation of sand-type debris and in faster melt cooling.

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Core Melt Stabilization Concepts for Existing and Future LWRs and Associated Research and Development Needs

Cited by 12 publications

References 0 publications

High-Temperature Characterization of Melted Nuclear Core Materials: Investigating Corium Properties Through the Case Studies of In-Vessel and Ex-Vessel Retention

High-Temperature Characterization of Melted Nuclear Core Materials: Investigating Corium Properties Through the Case Studies of In-Vessel and Ex-Vessel Retention

A Comparative Study on Severe Accident Phenomena Related to Melt Progression in Sodium Fast Reactors and Pressurized Water Reactors

Experimental investigations on melt coolability with simulated decay heat—The influence of delay time in flooding

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