Core Melt Research Articles

Effect of certain metrological parameters on the extension of urgent protective action zone in Tehran research reactor accidents

In a core melting accident of a reactor, the release of radioactive material into the environment is inevitable. The amount of effective dose received by people around these facilities and consequently the urgent protective action zone (UPZ) strongly depend on the climatic conditions at the time of the accident and afterwards. In this study, the effect of wind speed, air stability class, temperature, and rainfall on the total effective dose equivalent (TEDE) at different time/place intervals of a hypothetical partial core melt accident around a research reactor was investigated using simulation method. The results showed that the stability class E, warm temperature, and light rain could lead to the highest dose of individuals. In addition, very low wind speeds were recommended for the facilities located in cities with high density of high-rise buildings and/or tropical areas in this regard.

Mass transfer between a continuous oxide phase (U-O-Zr) and a steel droplet at liquid state: Potential impact on corium pool behaviour during in-vessel melt retention

In the frame of light water reactors’ severe accidents (SA), the success of a mitigation based on the core melt retention in the vessel lower head largely depends on the corium pool phases stratification. Such a melt is composed of molten partially oxidised materials relocated from the core and molten steel mainly coming from the ablation of lower head internal structures and the vessel wall during the transient. Because of the miscibility gap in the U-O-Zr-steel thermodynamic system at liquid state, such a pool can go through various stratified states during a transient with a heavy metal phase at the bottom or a light metal one at the top.Actually, there is a significant knowledge gap related to steel migration to the pool that influences the stratification transient and can lead in some cases to a large uncertainty on the heat flux that the pool imposes on the vessel lower head. In particular, when the ablated vessel wall is considered, the molten steel relocation, either as a continuous lighter phase above the pool or as dispersed phase in the oxide pool, depends on the behaviour of the refractory crust at the pool/vessel interface (that can undergo dissolution or mechanical damage). In the case of a dispersed metal phase in the oxide pool, determining the subsequent state of the lower head pool requires the evaluation of the thermochemical interaction between both materials and their possible hydrodynamic separation. To the best of our knowledge, no model described these coupled phenomena in integral SA codes.In this paper, an integral model is proposed to better assess the phenomena at stake in such a case and clarify the need (or not) to enhance stratified pool models in SA codes. To do so, this integral model was constructed on “first-order” hypotheses; in particular, thermal equilibrium is assumed and a simplified representation of the U-O-Zr-steel thermodynamic system is considered. The model has been implemented in the PROCOR software platform and used for a parametric analysis that provides insightful order of magnitudes and trends. The main parameters are the droplet size, the corium chemical composition and the model closures associated with interfacial mass transfer.The results reported here show that mass transfer, “competing” with droplet hydrodynamics, plays an important role for droplets of millimetric size that can, depending on the pool composition, relocate under the oxide under the form of a continous heavy metal layer. For larger droplets, it is a light metal layer above the pool that can be expected. The prolonged existence of a dispersed metal phase that would remain mixed in the oxide phase seems to be only possible for a very narrow range of pool composition near the stratification inversion threshold.

Современные ангионеврологические тренды медицинской помощи в случае мозгового инсульта среди населения, в том числе проживающего на временно загрязненных территориях Украины, и УЛПА на ЧАЭС: современный взгляд на проблему и собственные исследования (к 35-летию техногенной аварии на ЧАЭС)

В статье освещен инсульт как проблема мирового масштаба, его связь с тремя группами факторов риска (патогенетическими, психогенными и экологическими — загрязнение окружающей среды). В преддверии 35-й годовщины техногенной катастрофы в Чернобыле (26 апреля 1986 года) автор сравнил данные по инсульту в Украине и в Японии после аварии на АЭС «Фукусима-1» (11–12 марта 2011 года), которая привела к расплавлению активной зоны реакторов сразу на трех энергоблоках. Поднимаются вопросы оптимизации оказания специализированной медицинской помощи больным с мозговым инсультом, факторами риска инсульта. Необходимо минимизировать последствия, оптимизировать лечение, профилактику факторов риска, особенно в тех районах, где население подвергается опасности инкорпорированного радиационного поражения организма через пищевые продукты и воду. Также необходимо улучшение специализированной инсультной помощи, создание инсультных центров в областях Украины. Проведенные исследования свидетельствуют, что причин для возникновения мозговых инсультов много, и на сегодняшний день они недостаточно изучены. Особенностями клинического течения мозговых инсультов у участников ликвидации последствий аварии (УЛПА) на ЧАЭС являются развитие их в среднем в 40–49 лет и пожилом возрасте, длительный сосудистый анамнез, что характеризуется трансформацией в дебюте вегетососудистой дистонии в дисциркуляторную (пострадиационную) энцефалопатию, преобладание ишемического характера мозгового инсульта над геморрагическим, поражение преимущественно каротидного бассейна, относительная доброкачественность клинического течения, прогрессирующий характер психических расстройств, сопутствующая соматическая коморбидная патология, наличие устойчивого патоморфологического субстрата со стороны как сосудистой эндотелиальной дисфункции, так и мозговой ликворной системы. Полученная информация дополняет данные о патогенезе, клинике и диагностике мозгового инсульта у УЛПА и представляет практический интерес. Поднятые вопросы оптимизации оказания специализированной помощи в инсультных отделениях, центрах, достаточной их комплектации как кадрами, так и аппаратурой, а также вопросы последующей реабилитации становятся краеугольным камнем в преодолении эпидемии инсульта. Исходя из сказанного, сохраняется необходимость продолжения фундаментальных исследований в этой области и скорейшей разработки, внедрения научной концепции по профилактике, раннему выявлению основных модифицируемых факторов риска инсульта совместно с семейными терапевтами, специалистами узкого профиля. Надлежащее лечение инсульта, безусловно, улучшит качество жизни пациентов и их родственников.

A scoping study on debris bed formation from metallic melt coolant interactions

During a hypothetical severe accident in a light water reactor, the core and internal structures melt down and the molten materials (corium) may fall into a water pool, forming a debris bed from so-called fuel coolant interactions (FCI), whose characteristics are important to corium coolability. The present study is concerned with the characterization of an ex-vessel debris bed forming form the FCI of a metal rich corium jet falling into a water pool in the reactor cavity. Five scoping tests were carried out on the DEFOR-M test facility using metallic Tin melt at different superheats. The molten tin was employed as the simulant of metal (Zr/Fe) rich corium melt. The processes of melt jet fragmentation and debris formation was recorded by high-speed cameras. The contour and volume of the resulting debris bed were measured by a three-dimensional laser scanning system, and the debris particles were sieved for their size distribution. The experimental results revealed the effects of melt superheat and water subcooling on the characteristics of a debris bed, including the bed’s configuration and porosity, the particles’ morphology and size distribution. A preliminary comparison of debris bed characteristics between metallic melt and oxidic melt was also provided.

Influence of Passive Containment Cooling System on containment thermal hydraulic transients during a postulated severe accident

The nuclear reactor containment is an important safety system that acts as an ultimate barrier to prevent the release of radioactive fission products to the environment in case of a Severe Accident (SA). The containment system of advanced nuclear reactors is expected to be designed to meet the challenges posed by severe accidents. The response of containment to pressure and temperature build-up during a SA with core melt needs to be assessed to ascertain its structural integrity and leak tightness. The focus of this paper is on analysing the influence of a natural circulation based conceptual Passive Containment Cooling System (PCCS) on the containment pressure and temperature transients in a 220 MWe Pressurized Heavy Water Reactor (PHWR) and highlighting its role as an additional energy management feature. The scenario considered is a Large Break Loss of Coolant Accident (LBLOCA) along with failure of Emergency Core Cooling System (ECCS), moderator and calandria vault water cooling systems. A validated, lumped parameter computer code called Thermal Hydraulics of Containment (THYCON) containing a PCCS model is used for the analysis. The mathematical formulations relevant to the present analysis, solution methodology, validation and benchmarking exercises of THYCON and PCCS model with available experimental and numerical works is presented. It is then applied to analyse the influence of the conceptual PCCS on the long-term containment transients during a postulated SA. Parametric studies are conducted to elucidate the influence of location and operating conditions of PCCS. The key findings are presented and discussed in this paper.

Study on Selection of Internal Initiating Events of Small Natural Circulation Lead-Cooled Fast Reactor

The initiating events, as the starting point and basis for the deterministic safety analysis and probabilistic safety assessment of the lead-cooled fast reactor (LFR), provide important guidance on the design optimization and safe operation of reactors. This paper, based on the current design scheme of the small natural circulation lead-cooled fast reactor SNCLFR-100 and by reference to the experience of selecting initiating events of other advanced fast reactors, with the generalized “core melting” taken as the top target event, provides the internal initiating events of SNCLFR-100 by deduction in the master logic diagram (MLD) method, which form a relatively complete list of internal initiating events. This study thus can provide a theoretical basis for the safety analysis of natural circulation LFRs.

Theoretical Study on Heating Process of Micro-Al Particles in Laminar Flame

ABSTRACT In order to understand heating process of micro-aluminum particles in hot gas environment, preheating model and oxide film rupture models are established to analyze heating and rupture of oxide film procedure before Al particle combustion. Studying heating process of Al particles can better understand the heating phenomenon, including rupture and agglomeration of Al particles during heating. In oxide film rupture model, a thermal-stress-rupture (TSR) model is established for Al core in solid state, which describes extruding stress between oxide film and aluminum core due to different expansion coefficients. A molten-aluminum-state (MAS) model is established after Al core melts, which describes the relationship between pressure of liquid Al core and heating temperature. Results indicate that as diameter of Al particles increases, preheating time becomes longer. Rupture temperature of the oxide film is higher than the melting point of Al. It is better to use the rupture temperature to calculate preheating time. Additionally, the proposed heating model can predict preheating time of micro-Al particles more accurately than Sundaram’s model. For particle with diameter of 68.3/84/121.6 μm in Feng’s experiments, prediction deviation of this study is only 27.6%/3.8%/4.3%, respectively.

Development of the Penetration Tube Failure Analysis Program in the Lower Head of the Reactor Vessel During a Severe Accident

For analysis of an ex-vessel severe accident, the corium melt conditions inside the reactor vessel are important at the time of the reactor vessel failure together with the reactor vessel failure mode. To determine penetration tube failure in the lower head of the reactor vessel during a severe accident, the Korea Atomic Energy Research Institute developed the PENetration Tube Analysis Program 2.0 (PENTAP 2.0) and carried out validation work based on experimental data that can simulate penetration tube heatup, rupture, penetration weld failure, and penetration tube ejection failure. A numerical simulation was undertaken to investigate the effect of the presence of melt in a tube, the expansion direction of the reactor vessel hole, and wall ablation on tube failure using PENTAP 2.0. The simulation results showed that the presence of melt inside the tube helps prevent tube ejection. When melt is not in the penetration tube, tube ejection is strongly dependent on the expansion direction of the reactor vessel hole.

Kinetic partitioning of major-minor cations between olivine and Hawaiian tholeiitic basalt under variable undercooling and cooling rate conditions

In order to elucidate the kinetic partitioning of cations between olivine and basalt, we performed undercooling (−ΔT) and cooling rate (CR) experiments at atmospheric pressure and QFM-2 buffer. Starting from the superliquidus temperature of 1250 °C, a Hawaiian tholeiitic basalt was cooled at the rates of 4 (CR4), 20 (CR20), and 60 (CR60) °C/h to the final target temperatures of 1175 °C (−ΔT = 35 °C; −ΔT35) and 1125 °C (−ΔT = 85 °C; −ΔT85). Results show that polyhedral olivine morphologies are obtained at -ΔT35, whereas strong disequilibrium skeletal and/or dendritic textures form at -ΔT85. The amount of forsterite in olivine decreases from to 85% to 78% with increasing both -ΔT and CR. A diffusive boundary layer also develops in the melt next to the olivine surface and its composition becomes progressively enriched in Ca, owing to its incompatible behavior with the lattice site. Residual melts are progressively depleted in silica and enriched in alkali from CR4 to CR60, but silica-rich melts are observed with increasing -ΔT. In terms of Fe2+-Mg exchange, olivines obtained at -ΔT35 are always in equilibrium with the diffusive boundary layer, comprising both the interface melt next to the olivine surface and the far-field melt where all chemical gradients cease. At -ΔT85, however, the Fe2+-Mg exchange indicates two distinct equilibration stages between olivine core and far-field melt, and between olivine rim and interface melt. Partition coefficients (Kd) of Mg, Fe, Mn, Ca, and Cr calculated at the olivine-melt interface preferentially change as a function of -ΔT rather than CR. From -ΔT35 to -ΔT85, KdMg, KdFe, KdMn, and KdCr remarkably increase, whereas the opposite applies to KdCa. Through the application of equilibrium partitioning models, we found that Mg, Fe, Mn, and Ca are incorporated into the olivine lattice site at near-equilibrium proportions. This generally good agreement with modeling data demonstrates that diffusive mass transport of cations in our experiments occurred under the conditions of local equilibrium at the olivine surface. In contrast, marked deviations from the expected equilibrium are found for KCr in response to the major influence of crystal field stabilization energy on cation incorporation.

Development of mechanistic cladding rupture model for integrated severe accident code ISAA. Part II: DVI line small-break LOCA in CAP1400

Cladding ballooning and rupture are the important phenomena at the early stage of a severe accident (SA). For example, the SA caused by the loss of coolant accident (LOCA), the decrease in primary loop pressure and the increase in core temperature will lead to the cladding ballooning and rupture. The cladding deformation affects the flow distribution and causes local flow blockage. Meanwhile, water vapor will enter the fuel gap from the rupture, thereby increasing the surface area of cladding that is oxidized by the steam. At present, the widely used integrated severe accident analysis codes cannot analyze fuel rods thermal-mechanical behavior at the early SA stage, and the judgment of cladding rupture is only based on simple parameter model. This paper integrates the developed FRTMB (core Fuel Rod Thermal-Mechanical Behavior analysis module) module into the integrated severe accident analysis code ISAA, so that the coupled system ISAA-FRTMB can analyze fuel rods thermal–mechanical behavior and judge cladding rupture. Part I introduces the need of developing the FRTMB module, verification of the module, and how the module interactively works in the severe accident analysis code. And evaluates the steady-state fuel rods thermal–mechanical behavior at different enrichment areas of the CAP1400 reactor. Part II focuses on analyzing fuel rods behavior during a hypothetical DVI Line (Direct Vessel Injection line) small break-out accident (break diameter d = 4 in.) of the CAP1400, and predicts the cladding rupture time and the corresponding failure temperature. Due to failure of components of the emergency core cooling system and assumed unavailability of several preventive and mitigative accident management measures (AMM), the accident develops into a severe accident scenario with core melt and reactor pressure vessel failure. The mechanical analysis results show the top of the fuel rod is the first to rupture, rather than the peak node. The primary factor affecting the fuel pellet strain is the deformation caused by thermal expansion, while the densification and swelling change little. These results demonstrate the applicability and reliability of ISAA-FRTMB in analyzing fuel rods thermal–mechanical behavior and judging cladding rupture during transient accidents.

Numerical analysis of melt migration and solidification behavior in LBR severe accident with MPS method

In Lead-based reactor (LBR) severe accident, the meltdown and migration inside the reactor core will lead to fuel fragment concentration, which may further cause re-criticality and even core disintegration. Accurately predicting the migration and solidification behavior of melt in LBR severe accidents is of prime importance for safety analysis of LBR. In this study, the Moving Particle Semi-implicit (MPS) method is validated and used to simulate the migration and solidification behavior. Two main surface tension models are validated and compared. Meanwhile, the MPS method is validated by the l-plate solidification test. Based on the improved MPS method, the migration and solidification behavior of melt in LBR severe accident was studied furthermore. In the Pb–Bi coolant, the melt flows upward due to density difference. The migration and solidification behavior are greatly affected by the surface tension and viscous resistance varying with enthalpy. The whole movement process can be divided into three stages depending on the change in velocity. The heat transfer of core melt is determined jointly by two heat transfer modes: flow heat transfer and solid conductivity. Generally, the research results indicate that the MPS method has unique advantage in studying the migration and solidification behavior in LBR severe accident.

Microstructure and mechanical properties of the LWR solidified melt prototype obtained by the out-of-pile experiment

Removing the solidified core melt from damaged reactors by drilling or cutting requires knowledge of both mechanical properties and its microstructure. The ingot of the light-water power reactor prototype corium with a different blend composition containing ZrO2, UO2 and Zr was obtained in this work. The microstructure has been studied and the mechanical properties of the light-water power reactor core solidified melt materials have been determined. All samples showed typical brittle fracture behavior.

Molecular dynamics study of melting, diffusion, and sintering of cementite chromia core–shell particles

For the additive manufacturing of chromium-oxide coated stainless steel powders, the melting, diffusion, sintering behaviors of the cementite phase in the stainless steel during heat treatment can affect the products’ physical and chemical properties, and thus, are worth studying. In this work, we use ReaxFF reactive force-field-based molecular dynamics simulations to investigate the melting, diffusion, and sintering at the molecular level through an atomistic-scale level representation for the cementite with a Cr2O3 coating passivation layer. Additionally, we also compare the Lindemann indices of the Cr2O3 cementite core–shell particle with those of a pure cementite particle of the same size, in order to understand the influence of the chromium-oxide layer on the cementite melting. Results indicate that the Cr2O3 shell layer does not change the melting temperature of the core–shell particle. However, the Cr2O3 shell layer’s oxygen atoms diffuse into the core and increase the impurity level in the Fe3C crystal, destabilizing the cementite structure in the core. We observe that the melting of the core–shell particle starts from the core while, in contrast, the melting of the pure-cementite particle begins from the surface. In addition, a comparison of mean squared displacements of hydrogen (H), oxygen (O), carbon (C), chrome (Cr), and iron (Fe) atoms in the solid (900 K) and liquid (2000 K) phases indicates that the atom mobility in the liquid phase is significantly higher than in the solid phase. H and O atoms in the particle exhibit a higher mobility level than Fe, Cr and C atoms. When the temperature is elevated from 900 K to 2000 K, the O content in the core and C content in the shell rises, while the O content in the shell and C content in the core declines. Furthermore, the sintering simulation shows that there are four stages for the sintering in the liquid phase: the two particles move randomly (Stage 1) until they make the first contact (Stage 2), the shell layers subsequently intermingle (Stage 3), and finally the particle cores fuse (Stage 4). The sintering in the solid phase (900 K) involves the former three stages. However, the temperature of 900 K is not sufficiently high to bring the two particles into Stage 4 during the time scale of our simulations.

Numerical study on thermal-hydraulics of external reactor vessel cooling in high-power reactor using MARS-KS1.5 code: CFD-aided estimation of natural circulation flow rate

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.

On water-ingression during top-flooding of corium melts

In the course of severe accident in a nuclear power plant, the corium mixture may flow down in the reactor pit and start a thermal attack of the basemat concrete. A simple way to terminate the melt progression may be to add water on top of it. Among the various physical processes that may participate in the quenching, water may penetrate into the solidified corium through cracks generated by the thermal stress during the solidification.The present paper aims at providing a clarification of the process though an analysis of the few dedicated experimental data available using proto-typical corium, namely the SSWICS 1-7 experiments. At first, a general reminder and analysis of the data is given. Then, the thermal-hydraulic aspects are investigated through the construction of a dedicated heat flux correlation and its use via the experimentally measured post-test permeabilities. The analysis is supported by 1D and 2D evaluations with the multiphase flow code MC3D-PREMIX, slightly modified for the purpose. It is concluded that the heat flux drastically decreases with the amount of added concrete material. Furthermore, 2D border effects are investigated and their importance highlighted in view of the experimental results analysed. These effects should explain the absence of observation of water ingression heat flux in the cases with large concrete amounts. Following, the paper proposes a model for the created permeability. Due to the complexity of the process and to the large uncertainties of the needed material properties in the considered situation with very high temperatures, a semi-empirical model is derived.Lastly, the model is adapted to the situation with internal decay heat, although no open experimental data is available to precisely support any model. A complete modeling is out of the scope of the paper, hence the focus is to provide hints for a first analysis of the impact of decay heat on its coolability.

Plant-specific assessment of the natural circulation-induced creep rupture characteristics in the RCS pressure boundary during the SBO accident

In nuclear power plants (NPPs), one of the important yet uncertain phenomena during a station blackout (SBO) accident is the possibility for a counter-current natural circulation flow in the reactor coolant system (RCS) pressure boundary, which could divert the core melt progression and/or the in-vessel fission product (FP) distribution and, in turn, lead to a heatup and eventual creep rupture. From the risk and accident management points of view, one of key concerns is to predict more explicitly whether such a creep rupture can occur in the RCS pressure boundary, and if it indeed can, also predict the most probable time and the most likely location. This paper presents plant-specific analysis results on the potential characteristics of natural circulation-induced creep rupture of the RCS pressure boundary that could potentially take place during an SBO accident of a reference pressurized water reactor (APR1400). MELCOR-CFD (Computational Fluid Dynamics) coupled analysis was performed to determine more precisely the relevant physical parameter that could greatly influence on the behaviour of natural circulation and the ensuing creep rupture. Multiple sensitivity analyses were also carried out to investigate the influence that particular MELCOR modeling schemes and plant conditions have in estimating the RCS creep rupture characteristics. The analysis results and relevant insights were summarized in terms of particular points of interest.

Determination of urgent protective action zone in Tehran research reactors-associated accidents

In accidents leading to the release of radioactive materials into the environment, it is of great importance to determine the urgent protective action zone (UPZ) in the emergency preparedness phase. The present study was conducted to determine the radius of this area for a low-power research reactor, taking into account the worst conditions due to a hypothetical partial core melting accident. The UPZ was calculated based on the IAEA recommended generic criteria for the public and emergency workers during an emergency exposure. Furthermore, according to the simulated predictable operational intervention level (OIL) in the emergency response phase, a new criterion for exposure via ground shine (OIL1γ) was proposed in UPZ determination for the individuals in the research reactor-associated accidents.

Influence of coupled factors on premixing and fragmentation of mild thermal interaction

After core melt jets drop into the water pool, they break up into several parts and premix with water, and subsequently fine-fragmentation may take place, leading to steam explosions. Besides, mild thermal interaction can also occur in fuel–coolant interactions (FCIs) during core melt accidents in light water reactors. However, this phenomenon is not fully understood owing to the various coupled mechanisms involved. Therefore, we conduct an experimental study to understand the influence of the underlying factors and thermal interaction mechanisms on the premixing and fragmentation processes of FCIs within a confined space. Based on the morphology method, the coupled effects of melt mass, water level, and melt superheating temperature are investigated, quantified in terms of jet breakup length, solidification time, and fragments size distribution. Based on the corresponding pressure and temperature histories, two types of influencing factors are studied in detail: the effect of melt properties coupled with the water level; and the effect of melt superheating temperature coupled with the water level. Finally, a new bubble dynamic phenomenon called the premix expansion is featured based on Rayleigh’s equation. These experimental data on mild thermal interaction are expected to be of significance in the development of models and validation of codes.

Thermo-mechanical behavior of an ablated reactor pressure vessel wall in a Nordic BWR under in-vessel core melt retention

The reactor pressure vessel (RPV) of a nuclear reactor is one of the key safety barriers preventing radioactive environmental releases during a severe accident. One of the promising strategies of severe accident management (SAM) is to retain the molten core having continuous decay heat inside the RPV by natural water cooling of the external vessel surface. The feasibility of such a strategy relies on complex safety analyses including accurate prediction of vessel thermo-mechanical behavior which can be assessed by mechanical stresses and strains.In this paper, we present the stress–strain response of an ablated RPV of a Nordic boiling water reactor (BWR) to dynamic thermomechanical loads set by expanding volumetrically heated molten pool inside the RPV cooled by water at the external surface. MELCOR 2.2.9541 severe accident code is used to simulate the in-vessel behavior and provides the input conditions for dedicated structural analysis of the RPV using ANSYS® Mechanical APDL 19.2. A creep model of the SA533B1 vessel steel is validated against uniaxial creep tests carried out by INEL (Idaho National Engineering Laboratory) and creep tests performed at CEA (French Alternative Energies and Atomic Energy Commission) as part of the OLHF (OECD Lower Head Failure) Project. Two generic severe accident scenarios are considered: (i) Station Blackout (SBO) and (ii) Station Black-out and Loss-of-coolant Accident (SBO + LOCA). In both scenarios, we found that the RPV has maintained structural integrity considering two failure criteria: stress-based and strain-based.

Investigation of ex-vessel core catcher for SBO accident in VVER-1000/V528 containment using MELCOR code

To mitigate consequences of core melting, an ex-vessel core catcher is investigated in this study. Instructions should be obeyed to cool down the corium caused by core melting. The corium destroys the reactor containment and causes radioactive materials to be released into the environment if it does not cool down well. It is important to build a core catcher system for the reception, localization, and cool down of the molten corium during a severe accident resulting from core melting. In this study, the role of a core catcher in the VVER-1000/v528 reactor containment during a station black out (SBO) accident is evaluated using the MELCOR1.8.6 code. In addition, parametric analyses of the SBO for (i) SBO accidents with emergency core cooling system (ECCS) operation, and (ii) without ECCS operation are performed. Furthermore, thermal–hydraulic analyses in dry and wet cavities with/without water are conducted. The investigations include the reduction of gases resulting from molten–corium–concrete interactions (H2, CO, CO2). Core melting, gas production, and the pressure/temperature in the reactor containment are assessed. Additionally, a full investigation pertaining to gas release (H2, CO, CO2) and the pressure/temperature of the core catcher is performed. Based on MELCOR simulations, a core cavity and a perimeter water channel are the best options for corium cooling and a lower radionuclide release. This simulation is also theoretically investigated and discussed herein. The simulation results show that the core catcher system in addition to an internal sacrificial material reduces the containment pressure from 689 to 580 kPa and the corresponding temperature from 394 to 380 K. Furthermore, it is observed that the amount of gases produced, particularly hydrogen, decreased from 1698 to 1235 kg. Moreover, the presence of supporting systems, including an ECCS with a core catcher, prolonged the core melting time from 16,430 to 28,630 s (in an SBO accident) and significantly decreased the gases produced.