Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors

Fichot, F.; Carénini, L.; Sangiorgi, Marco; Hermsmeyer, S.; Miassoedov, A.; Bechta, Sevostian; Zdarek, J.; Guenadou, D.

doi:10.1016/j.anucene.2018.03.040

Cited by 38 publications

(10 citation statements)

References 11 publications

Supporting

Mentioning

Contrasting

Unclassified

Order By: Relevance

“…A thermal flux of 2 megawatts per square meter can be reached or exceeded where the metal layer is only 10 cm thick. This so-called focussing effect [96][97][98], particularly during transient phases, can lead to critical and dangerous situations, capable of rapidly melting the vessel wall locally. Some actual reactor designs include a system able to mitigate this risk by injecting water around the vessel to cool the outer wall (Figure 5) and prevent it from breaking through.…”

Section: High Temperature Regime (Above 2200 • C)mentioning

confidence: 99%

“…This strategy-known as IVR for in-vessel retention-which is currently attracting a lot of attention was introduced about 20 years ago [99,100] and first applied to two reactor designs, AP600 [101] (later replaced by the AP1000 design but keeping the IVR option) and VVER-440, this last case leading to practical application at the Lovisa Plant (Finland) [102]. Some studies are still ongoing on the extension of this strategy to high power reactors [97,98].…”

Section: High Temperature Regime (Above 2200 • C)mentioning

confidence: 99%

See 1 more Smart Citation

Corium Experimental Thermodynamics: A Review and Some Perspectives

Barrachin

2021

Thermo

View full text Add to dashboard Cite

More than 30 years ago a specialist meeting was held at Joint Research Center Ispra (Italy) from 15 to 17 January 1990 to review the current understanding of chemistry during severe accidents in light water reactors (LWR). Let us consider that, at the end of the 1980s, thermodynamics introduced in the severe accident codes was really poor. Only some equilibrium constants for a few simple reactions between stoichiometric compounds were used as well as some simple correlations giving estimates of solidus and liquidus temperatures. In the same time, the CALPHAD method was developed and was full of promise to approximate the thermodynamic properties of a complex thermochemical system by the way of a critical assessment of experimental data, a definition of a simple physical model and an optimisation procedure to define the values of the model parameters. It was evident that a nuclear thermodynamic database had to be developed with that new technique to obtain quite rapidly prominent progress in the knowledge of thermochemistry in the severe accident research area. Discussions focused on the important chemical phenomena that could occur across the wide range of conditions of a damaged nuclear plant. The most pressing need for improved chemical models is identified with condensed phase mixtures to model the corium progression. This paper reviews more than 30 years of experimental data production in the field of corium thermodynamics. This work has been conducted through multiple international programs (EURATOM, ISTC, OECD) as well as through more specific studies conducted at the national scale. This research has been capitalised in specific databases such as NUCLEA and TAF-ID, databases developed at IRSN and at CEA, respectively, and are now used in degradation models of the severe accident simulation codes. This research is presented in this paper. In the conclusion, we outline the research perspectives that need to be considered in order to address today’s and tomorrow’s issues.

show abstract

Section: High Temperature Regime (Above 2200 • C)mentioning

confidence: 99%

Section: High Temperature Regime (Above 2200 • C)mentioning

confidence: 99%

Corium Experimental Thermodynamics: A Review and Some Perspectives

Barrachin

2021

Thermo

View full text Add to dashboard Cite

show abstract

“…Although this IVR strategy is not new, its potential is still not fully investigated. In order to consolidate the results of these investigations, perform additional studies for the missing parts, and explore the possibilities of IVR in more detail, the HORIZON 2020 project In-Vessel Melt Retention Severe Accident Management Strategy for Existing and Future NPPs (IVMR) [9] was initiated.…”

Section: Mass Of Zircaloymentioning

confidence: 99%

Modelling of Severe Accident and In-Vessel Melt Retention Possibilities in BWR Type Reactor

Valinčius

Kaliatka

Кaliatka

et al. 2018

Science and Technology of Nuclear Installations

View full text Add to dashboard Cite

One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection of cooling water. A computer code RELAP/SCDAPSIM MOD 3.4 was used for the numerical simulation of the accident. Using a model of the entire reactor, a full accident sequence from the large break to core uncover and heat-up as well as corium relocation to the lower head is presented. The ex-vessel cooling was modelled in order to evaluate the applicability of RELAP/SCDAPSIM code for predicting the heat fluxes and reactor pressure vessel wall temperatures. The results of different ex-vessel heat transfer modes were compared and it was concluded that the implemented heat transfer correlations of COUPLE module in RELAP/SCDAPSIM should be applied for IVR analysis. To investigate the influence of debris separation into oxidic and metallic layers in the molten pool on the heat transfer through the wall of the lower head the analytical study was conducted. The results of this study showed that the focusing effect is significant and under some extreme conditions local heat flux from reactor vessel could exceed the critical heat flux. It was recommended that the existing RELAP/SCDAPSIM models of the processes in the debris should be updated in order to consider more complex phenomena and at least oxide and metal phase separation, allowing evaluating local distribution of the heat fluxes.

show abstract

“…В настоящее время на основе анализа произошедших тяжелых аварий и опыта борьбы с их последствиями принят принцип глубокоэшелонированной защиты, опирающийся на множественность и дублирование уровней защиты и включающий последовательность барьеров на пути выхода радиоактивных материалов в окружающую среду [6]. При этом для обоснования работоспособности как существующих, так и разрабатываемых систем защиты проводились и проводятся экспериментальные исследования и теоретический анализ процессов, в том числе, а в последнее время преимущественно, физикохимических процессов, протекающих на всех стадиях тяжелой аварии [7][8][9][10][11][12][13].…”

Section: Introductionunclassified

“…IVR" расплава кориума), при этом основным условием удержания расплава является отсутствие кризиса теплообмена на наружной водоохлаждаемой поверхности корпуса [1,2,7,8,10,14]. После формирования на днище корпуса реактора двухжидкостной оксидно-металлической ванны расплава с верхним расположением металлической жидкости (расплав представляет собой систему двух несмешивающихся жидкостей -одна на основе оксидов, другая на основе металлов) наиболее опасным местом с точки зрения возникновения кризиса теплообмена является зона контакта корпуса с металлической жидкостью, в которой фокусируется тепловой поток, передаваемый от оксидной жидкости к металлической [7,8,10,15,16]. Величина подводимого от металлического слоя расплава к корпусу реактора теплового потока при прочих равных условиях зависит от сценария тяжелой аварии, уровня остаточного энерговыделения вследствие распада присутствующих в расплаве радионуклидов, физико-химических процессов, протекающих в расплаве, а непосредственно -от температуры, теплофизических свойств и толщины металлического слоя расплава.…”

Section: Introductionunclassified

Моделирование Окисления Расплава Активной Зоны Ядерного Реактора При Наличии Оксидной Корки На Поверхности Расплава

Khabenskii¹,

Альмяшев²,

Granovskii³

et al. 2021

Журнал Технической Физики

View full text Add to dashboard Cite

At a severe accident of nuclear power plants with light-water reactors, the most effective way to localize the forming melt (corium) is to keep it in the cooled reactor vessel, the integrity of which depends on the value of heat flux from the melt to the reactor vessel. In this case, one of the critical processes is the melt oxidation by a water steam or a steam-air mixture. It process can lead to a significant increase in the thermal load on the reactor vessel due to a heat of exothermic reactions of oxidation of reducing agents, which presents in the melt, a thickness decreasing of the metallic part of the molten pool, and a hydrogen release. All of these factors strictly depends on the rate of oxidation. When considering the conditions of melt oxidation, it taken into account that for the accepted scenarios of a severe accident, the most realistic situation is the presence of a solid-phase oxide layer (oxidic crust) on the melt surface. Under these conditions, a dependence for calculating the rate of core melt oxidation based on the diffusion model proposed and its validation by using the obtained experimental data performed.

show abstract

Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors

Cited by 38 publications

References 11 publications

Corium Experimental Thermodynamics: A Review and Some Perspectives

Corium Experimental Thermodynamics: A Review and Some Perspectives

Modelling of Severe Accident and In-Vessel Melt Retention Possibilities in BWR Type Reactor

Моделирование Окисления Расплава Активной Зоны Ядерного Реактора При Наличии Оксидной Корки На Поверхности Расплава

Contact Info

Product

Resources

About