Modelling issues related to molten pool behaviour in case of In-Vessel Retention strategy

Carénini, L.; Fichot, F.; Seignour, N.

doi:10.1016/j.anucene.2018.04.032

Cited by 37 publications

(15 citation statements)

References 20 publications

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“…Nevertheless, it is worth noting that ASTEC's prediction of a very high top metal layer temperature can be tied to the earlier vessel failure. The reasons for ASTEC to predict such a high temperature can be related to the following modelling details (these issues are discussed in more detail in [50]): (1) ASTEC assumes that the molten steel relocated to the top of the corium pool has same temperature as the oxidic pool because it originates from melting of the internal structures surrounding by the oxide pool or from the ablation of the vessel wall in contact with the oxide. This leads to an overheated metal layer compared to MAAP, which accelerates ablation of the vessel wall adjacent to the metal layer.…”

Section: Discussion On the Significant Discrepancies Between The Codesmentioning

confidence: 99%

ASTEC–MAAP Comparison of a 2 Inch Cold Leg LOCA until RPV Failure

Blul

Brumm

Mascari

et al. 2018

Science and Technology of Nuclear Installations

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A 2 inch, cold-leg loss-of-coolant accident (LOCA) in a 900 MWe generic Western PWR was simulated using ASTEC 2.1.1 and MAAP 5.02. The progression of the accident predicted by the two codes up to the time of vessel failure is compared. It includes the primary system depressurization, accumulator discharge, core heat-up, hydrogen generation, core relocation to lower plenum, and lower head breach. The purpose of the code comparison exercise is to identify modelling differences between the two codes and the user choices affecting the results. The two codes predict similar primary system depressurization behaviour until the accumulation injection, confirming similar break flow and primary system thermal-hydraulic response calculations between the two codes. The choice of the accumulator gas expansion model, either isentropic or isothermal, affects the rate and total amount of coolant injected and thereby determines whether the core is quenched or overheated and attains a noncoolable geometry during reflooding. A sensitivity case was additionally simulated by each code to allow comparisons to be made with either accumulator gas expansion models. The two codes predict similar amount of in-vessel hydrogen generated and core quench status for a given accumulator gas expansion model. ASTEC predicts much larger initial core relocation to lower plenum leading to an earlier vessel failure time. MAAP predicts more gradual core relocation to lower plenum, prolonging the lower plenum debris bed heat-up and time to vessel failure. Beside the effect of the code user in conducting severe accident simulations, some discrepancies are found in the modelling approaches in each code. The biggest differences are found in the in-vessel melt progression and relocation into the lower plenum, which deserve further research to reduce the uncertainties.

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Section: Discussion On the Significant Discrepancies Between The Codesmentioning

confidence: 99%

ASTEC–MAAP Comparison of a 2 Inch Cold Leg LOCA until RPV Failure

Blul

Brumm

Mascari

et al. 2018

Science and Technology of Nuclear Installations

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“…The IVMR project [9,10], coordinated by IRSN between 2015 and 2019, aimed at providing new experimental data and a harmonized methodology for the in-vessel melt retention (IVR). The IVR strategy for LWR intends to stabilize and isolate the corium and the fission products inside the reactor pressure vessel and in the primary circuit.…”

Section: Ivmr Projectmentioning

confidence: 99%

Safety assessments and severe accidents, impact of external events on nuclear power plants and on mitigation strategies

Dorsselaere

Boutarfaïa

Albiol

et al. 2020

EPJ Nuclear Sci. Technol.

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The Fukushima-Daiichi accidents in 2011 underlined the importance of severe accident management (SAM), including external events, in nuclear power plants (NPP) and the need of implementing efficient mitigation strategies. To this end, the Euratom work programmes for 2012 and 2013 was focused on nuclear safety, in particular on the management of a possible severe accident at the European level. Relying upon the outcomes of the successful Euratom SARNET and SARNET2 projects, new projects were launched addressing the highest priority issues, aimed at reducing the uncertainties still affecting the main phenomena. Among them, PASSAM and IVMR project led by IRSN, ALISA and SAFEST projects led by KIT, CESAM led by GRS and sCO2-HeRO lead by the University of Duisburg-Essen. The aim of the present paper is to give an overview on the main outcomes of these projects.

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“…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

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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.

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Modelling issues related to molten pool behaviour in case of In-Vessel Retention strategy

Cited by 37 publications

References 20 publications

ASTEC–MAAP Comparison of a 2 Inch Cold Leg LOCA until RPV Failure

ASTEC–MAAP Comparison of a 2 Inch Cold Leg LOCA until RPV Failure

Safety assessments and severe accidents, impact of external events on nuclear power plants and on mitigation strategies

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

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