Accident Management Strategy Research Articles

Machine learning-Based method for gas leakage source term estimation in highway tunnels

Incidents involving the leakage of flammable and explosive chemicals in highway tunnels can lead to disastrous events such as fires and explosions. The distinctive features of highway tunnels, characterized by their confined and elongated structure, intensify the complexity of managing such accidents, potentially resulting in significant casualties and property damage. Rapid detection and acquisition of leakage source information during a flammable gas leak in a highway tunnel are of utmost importance for effective accident management. This study introduces a predictive model developed using a machine learning approach. The model uses time series data from gas concentration sensors and anemometers installed within the tunnel to predict parameters such as gas leakage velocity, source location, and initiation time. A computational fluid dynamics simulation of gas leakage within the tunnel was constructed, and a dataset was generated using sample generation and preprocessing methods. The results demonstrate that this method exhibits robust predictive performance. Furthermore, a prediction strategy is proposed to enhance the model’s predictive accuracy and resilience to data noise. This model offers valuable insights for improving accident management strategies for gas-related incidents in highway tunnels.

Numerical simulation of in-vessel steam explosion for China third-generation PWR

In this study, in order to evaluate the Reactor Pressure Vessel (RPV) integrity under in-vessel steam explosion loads for China third-generation Pressurized Water Reactor (PWR), 2D and 3D numerical simulation of in-vessel steam explosion is conducted using MC3D code. The sensitivity analyses for in-vessel steam explosion are performed, including break area, water temperature, water level, melt temperature, and break height. The results show that the premixing stage cannot cause RPV failure and the possibility of the steam explosion blowing the RPV upper head off is almost non-existent. When the break area increases, the maximum pressure and impulse of the RPV inner wall increase obviously first and then generally decrease slightly. With the increase of water temperature, the maximum pressure and impulse first increase and then decrease. When the water level is higher, the maximum pressure variation is small. When the water level is very low, the maximum pressure decreases. The maximum impulse generally decreases with the decrease of the water level. When the melt temperature exceeds 2850 K, the steam explosion is not sensitive to the melt temperature. With the increase of the break height, the maximum pressure and maximum impulse first increase and then decrease. Furthermore, based on the results of sensitivity analysis, two cases of conservative conditions are selected, and the RPV integrity under in-vessel steam explosion loads for China third-generation PWR is considered. The result shows that the integrity of the RPV can be maintained under the in-vessel steam explosion loads for China third-generation PWR with a large margin. These studies provide some reference to improve the severe accident management strategy in China.

Traffic accident classification using IndoBERT

<span>Traffic accidents are a widespread concern globally, causing loss of life, injuries, and economic burdens. Efficiently classifying accident types is crucial for effective accident management and prevention. This study proposes a practical approach for traffic accident classification using IndoBERT, a language model specifically trained for Indonesian. The classification task involves sorting accidents into four classes: car accidents, motorcycle accidents, bus accidents, and others. The proposed model achieves a 94% accuracy in categorizing these accidents. To assess its performance, we compared IndoBERT with traditional methods, random forest (RF) and support vector machine (SVM), which achieved accuracy scores of 85% and 87%, respectively. The IndoBERT-based model demonstrates its effectiveness in handling the complexities of the Indonesian language, providing a useful tool for traffic accident classification and contributing to improved accident management and prevention strategies.</span>

Concrete ablation depth analysis of OPR1000 NPP during top flooding ex-vessel cooling

This study assesses the integrity of OPR1000 nuclear power plant (NPP) containment during top flooding ex-vessel cooling in the event of a postulated severe accident. Maintaining containment integrity during ex-vessel cooling is critical for preventing substantial release of radioactive materials to the environment. To analyze the feasibility of top flooding ex-vessel cooling of the OPR1000 NPP, the spread of molten corium in the dry reactor cavity is first calculated, followed by calculating the concrete ablation depth according to the reactor cavity flooding delay time. It is judged whether the leak-tightness integrity of the containment during ex-vessel cooling is maintained based on whether the containment liner plate (CLP) is damaged or not. In the first step analysis, the spread of molten corium in the dry reactor cavity is calculated using computational fluid dynamics (CFD) methodology for various cases, and results indicate a uniform deposition of approximately 143 tonnes of molten corium, with a thickness of about 33 cm, meaning that the thermal load is distributed uniformly in the reactor cavity. In the second step analysis, concrete ablation depths are calculated base on the delay time for flooding the reactor cavity. For the postulated severe accident, injecting coolant into the reactor cavity one hour after breach of the reactor vessel results in a concrete ablation depth of about 0.74 m, indicating no damage to the CLP and ensuring containment integrity. The maximum allowable delay time for flooding the reactor cavity for the postulated severe accident is evaluated to be 2.24 hours maintaining the leak-tightness integrity of the OPR1000 NPP containment. The research methodology and findings presented in this study are expected to aid in assessing the feasibility of top flooding ex-vessel cooling and establishing appropriate accident management strategies for nuclear power plants like the OPR1000 NPP.

Effectiveness re-evaluation on the intentional primary system depressurization during Zion-like Westinghouse PWR station blackout considering pressure dependence of radionuclides release

Previous studies have successfully resolved the issue of high-pressure melt ejection (HPME) followed by direct containment heating (DCH) during a total station blackout of Zion-like Westinghouse pressurized water reactor (PWR). This resolution is crucial, as the earliest occurrence of hot-leg creep failure can cause a decrease in the pressure difference between the reactor pressure vessel (RPV) and the primary containment vessel (PCV), falling below the cut-off pressure for HPME/DCH. As a recommended accident management strategy, intentional depressurization of the reactor coolant system (RCS) has been proposed. Depressurization leads to a delayed accident progression due to accumulator injection and maintains a pressure difference at the time of RPV failure below the cut-off pressure. In the present analyses using MAAP5, it was observed that depressurization before core heat-up, achieved by opening 2 power-operated relief valves (PORVs), resulted in the most delayed RPV failure, consistent with previous studies. However, present sensitivity analysis considering the pressure-dependent release of radionuclides from the fuel revealed significant changes in both the accident progression and the point of creep failure. In addition, the creep-failure point shifted from the RPV bottom to the RPV sidewall, and the pressure difference at that location exceeded the cut-off pressure. To prevent sidewall failure of the RPV, it may be advisable to employ a depressurization rate higher than that achieved by using 2 PORVs, even if it slightly accelerates the accident progression.

An analysis method to assess in-Calandria corium retention time in PHWRs

In Pressurized Heavy Water Reactors (PHWRs), during a postulated severe accident involving the core collapse, the core internals would fall onto the Calandria bottom and form a debris bed. Under such conditions, the core debris bed would be contained by the Calandria, which also facilitates its cooling by the Calandria vault water. In absence of any accident management action, the Calandria temperatures would rise, causing loss of material strength that may also threaten the structural integrity of Calandria. The gross structural failure of Calandria would limit its ability to further retain the core debris and would jeopardize the in-vessel accident management strategies. The knowledge of structural degradation of Calandria with time till its ultimate failure is important for formulating and assessing the severe accident management guidelines. Assessment of Calandria failure under accident conditions involves several important aspects such as complex thermo-mechanical loadings, temperature dependent elastic-plastic and creep modeling, flexibility of different parts and boundary constraints etc. In view of this, a systematic analysis method based on the sequentially coupled thermo- mechanical analysis has been devised and used for the structural assessment of Calandria. The failure assessment has been carried out as per the specified failure criteria for prominent failure modes including large unbounded deformations and plastic instability, large inelastic strains/ ductility exhaustion and creep-stress rupture. The paper covers various considerations taken into account during the analyses along with few case studies on Indian PHWRs.

Optimization strategy for SAM in nuclear power plants based on NSGA-II

Abstract The Severe Accident Management Guide (SAMG) is an important component of nuclear safety regulations. Many studies are being conducted to optimize severe accident management (SAM) strategies. To ensure the safety of nuclear power plants, decision makers need to monitor multiple parameters with security threats. Therefore, it is particularly important to search optimal SAM strategies under different numbers of mitigation targets. The Non-dominated Sorting Genetic Algorithm-II (NSGA-II) is an evolutionary algorithm that does not require derivative differentiation and is capable of population search. In this study, a nuclear power plant accident optimization strategy is developed using the Modular Accident Analysis Program (MAAP) in conjunction with NSGA-II. The strategy enables decision makers to consider multiple mitigation objectives in a complex decision environment. Focusing on the CPR1000, this study applies the optimization strategy to automatically search for optimal mitigation strategies for small break loss of coolant accident (SBLOCA) and station blackout hot leg creep rupture accidents (SBOHLCR). Comparing the optimization results with the basic accident sequence, it is found that the reactor pressure vessel (RPV) failure time is delayed from 72,702 s to 128,730 s under SBLOCA and from 23,828 s to 28,363 s under SBOHLCR. This study has also verified that the optimal SAM strategy obtained by the strategy through dual objective optimization has better mitigation effects than a strategy that only considers one objective. This optimization strategy has the potential to be applied to other types of severe accident management studies in the future.

Case studies on radioactivity and decay heat contribution from coating materials of irradiated accident tolerant fuel for a typical PHWR

Seven coating materials such as Al2O3, NbC, Si, SiC, TiN, Y2O3, and ZrC have been considered as a possible coating material towards development of accident tolerant fuel for a typical pressurized heavy water reactor. For an effective severe accident management strategy in addition to spent fuel handling, storage and transport, it is required to know the contribution to the radioactivity burden and decay heat load from the irradiated coating materials of accident tolerant fuel. Under irradiation of coating material in PHWR flux condition, the decay chain of each activated radionuclide has been simulated using ISOGEN code using one group homogenized point cross section. In this paper, radioactivity and decay heat contributions from different irradiated coating materials have been quantified and compared. The minimum contribution to radioactivity and decay heat is observed in SiC after irradiation of 445 days of normal operation. Out of all the coating materials considered in the study, most radioactivity and decay heat were added due to irradiation of coating materials Y2O3. From reactivity load, source term, and decay heat perspectives, silicon-based coating materials are found to be the most suitable.

In-depth investigation of natural convection thermal characteristics of BALI experiment through Eulerian computational fluid dynamics code and comparison with Lagrangian code

In-vessel retention through external reactor vessel cooling (IVR-ERVC) is a severe accident management (SAM) strategy that has been adopted and used in many nuclear reactors such as AP1000, APR1400, and light water reactor etc. Some reactor accidents have raised concerns about nuclear reactors among residents, leading to a decrease in residents' acceptability and many studies on SAM are being conducted. Experiments on IVR-ERVC are almost impossible due to its specificity, so fluid characteristics are analyzed through BALI experiments with similar condition. In this study, computational fluid dynamics (CFD) via Reynolds-averaged Navier-Stokes (RANS) and large eddy simulation (LES) for BALI experiments were performed. Steady-state CFD analysis was performed on three turbulence models, and SST k-ω model was in good agreement with the experimental measurement temperature within the maximum error range of 1.9%. LES CFD analysis was performed based on the RANS analysis results and it was confirmed that the temperature and wall heat flux for depth was consistent within an error range of 1.0% with BALI experiment. The LES CFD analysis results were compared with those of the Lagrangian-based solver. LES matched the temperature distribution better than SOPHIA, but SOPHIA calculated the position of boundary between stratified layer and convective layer more accurately. On the other hand, Lagrangian-based solver predicted several small eddy behaviors of the convective layer and LES predicted large vortex behavior. The vibration characteristics near the cooling part of the BALI experimental device were confirmed through Fast Fourier Transform (FFT) investigation. It was found that the power spectral density for pressure at least 10 times higher near the side cooling than near the top cooling.

Critical heat flux model for heated horizontally oriented cylindrical vessel and its application to PHWR calandria under severe accident scenario

In-vessel retention (IVR) of molten corium is a key severe accident management strategy adopted in pressurized heavy water reactors (PHWRs) and is recognised as the only option to manage a complete core melt scenario. Since the PHWR calandria vessel is completely surrounded by a large pool of water, the vessel itself is expected to act as a ‘core catcher’. The success of IVR strategy depends up on whether the actual heat flux imparted by molten corium is less than the critical heat flux (CHF). In the present work, a hydrodynamics model is developed and applied to obtain the variation of CHF along outer surface of 220 and 700 MWe Indian PHWR calandria vessel and tube. The thermal margin available for success of IVR strategy is evaluated. The governing equations for external buoyancy driven boundary layer flow are derived and solved in non-dimensional form. The details of the mathematical model, benchmarking exercises, validation aspects and application of the model to PHWR calandria vessel and tube are discussed. Parametric studies are presented to bring out the influence of diameter, system pressure, subcooling of water, depth of submergence and slip ratio on the variation of CHF along the cylindrical vessel surface.

Development of a surrogate model for quenching estimation of ex-vessel debris beds and its coupling with MELCOR

In the severe accident management (SAM) strategy for Nordic boiling water reactors (BWRs), a flooded reactor cavity is conceived to receive corium in case of vessel failure, with the hope that the discharged corium will fragment and form a coolable particulate debris bed in the deep water pool. The so-formed debris bed on the cavity basement is supposed to be very hot at the beginning and therefore its quenching is a prerequisite for long-term coolability. In previous study the coupled MELCOR/COCOMO simulation was employed to simulate quench process of ex-vessel debris beds in severe accident scenarios. Although it successfully extended the MELCOR capability, the calculation was dramatically slowed down by explosive computational cost of COCOMO. To overcome the limitation, the present study is to develop a surrogate model (SM) which can replace the mechanical code COCOMO and realize quick estimations of the quench process of ex-vessel debris beds. It was then coupled with MELCOR code for integral severe accident analyses of a Nordic BWR with cooling of ex-vessel debris beds. The SM was developed based on a database generated from COCOMO calculations of various one-dimension (1D) debris beds quenched in the reactor cavity, using artificial neural networks (ANNs). Finally, the coupled MELCOR/SM simulation was applied to safety analyses of postulated severe accident scenarios due to station blackout (SBO) in the BWR, where MELCOR performs integral analysis of accident progression while SM predicts the consequences (e.g. energy transfer) of debris bed quench. The simulation results show that the coupled MELCOR/SM simulation can predict the trends of containment pressure and pool temperature similar to those of the coupled MELCOR/COCOMO simulation. Compared with MELCOR standalone simulation, the coupled MELCOR/SM simulation predicted earlier pool saturation and containment venting.

A scoping study on remelting process of a debris bed in the lower head of reactor pressure vessel

Coolability and retention of core melt (corium) in the lower head of reactor pressure vessel (RPV) has been accepted as a severe accident management strategy to maintain the reactor pressure vessel (RPV) integrity of a light water reactor. To qualify the in-vessel melt retention strategy, lots of studies have been performed to investigate the natural convection heat transfer of a melt pool in the lower head. However, little work has been attributed to the precursory phase of the melt pool, i.e., the remelting process of a debris bed which is formed in the lower head at the very beginning of corium relocation from the core the lower head. The present study is motivated to conduct an experimental study on the debris remelting process. For this purpose, a dedicated test facility named COREM (COrium REMelting) is conceived and constructed, which features internal heating of electromagnetic induction and visualization of debris remelting dynamics. The test section is a semicircular vessel representing a slice of scaled-down RPV lower head, whose front and back walls are made of transparent tempered glass which facilitate visualization and induction heating of the debris bed. Fiber probes with multiple optical temperature sensors are mounted in the semicircular wall of the test section to measure its temperature distribution. In the scoping test, n-octanol and Wood's metal are selected as the simulant materials of metallic and oxidic components of corium, respectively, and their particles ware loaded in the test section to form a debris bed. This paper presents the first two scoping tests which have been performed with the COREM facility so far, under different mixing ratios of two debris materials. The measured data include the photography of debris remelting processes and the temperatures in the debris bed and the semicircular wall. Based on the temperature distributions, heat flux profile along the semicircular vessel wall is also estimated. The scoping tests well reproduce the dynamic process of debris remelting, with two distinct stages of fusion of n-octanol and Wood’s metal, i.e., melting successively from low to high melting-point debris particles. During the remelting process, the vessel wall temperatures increase with the polar angle firstly and then decrease gradually, leading to the highest temperature appearing in the middle of the lower layer of the stratified molten pool which is finally formed.

Backfitting of Dukovany NPP for design extension conditions

The contribution deals with the backfitting of Dukovany NPP for design extension conditions with emphasis on severe accidents. This NPP, located in Czech Republic, is equipped with 4 units of VVER-440/V213 reactors. During the last decades significant attention was paid to backfitting of these units. New safety systems aimed at mitigating the consequences of severe accidents have been designed and have been (are being) implemented. At first, special features of VVER-440/V213 reactor and containment equipped with bubbler condenser pressure suppression system are briefly described. The systematic process of plant backfitting for design extension condition is summarized further. Attention is paid to description of crucial severe accident management strategies that have been adopted and design and implementation of new safety systems that are necessary to fulfil these strategies. Than the requirements of Czech regulatory body (SUJB) set for practical elimination of early or large radioactive releases are discussed. Finally, the performances of severe accident management strategies are illustrated using the specific case of blackout scenario. For this purpose, an analysis was performed using ASTEC code developed by the French IRSN. The results of analysis are briefly described and discussed further. From the obtained results it follows that the high-pressure melting scenario is reliably avoided, corium is localised and stabilised inside the reactor vessel and the hydrogen in containment atmosphere is safely controlled by passive autocatalytic recombiners. Thus, stable plant state with preserved containment integrity is reached. Source term that has been estimated under these conditions indicates that the predicted releases are well below the criteria set by SUJB for early or large radioactive releases, and therefore long-term contamination of the environment is avoided.

MELCOR 2.2-ASTEC V2.2 crosswalk study reproducing SBLOCA and CSBO scenarios in a PWR1000-like reactor part II: Analysis of containment thermal-hydraulics and ex-vessel phenomena

The Accident Source Term Evaluation Code (ASTEC) and Methods of Estimation of Leakages and Consequences of Releases (MELCOR) system codes are both used by Tractebel to perform Severe Accident (SA) analyses for the Belgian Nuclear Power Plants (NPPs).This paper contains the second part of the comparative study between these two integral software focused on the SA progression in the containment after the Vessel Failure (VF).The main phenomena characterize this phase of the accident usually called ex-vessel phase are the Molten Corium Concrete Interaction (MCCI) and ex-vessel corium/debris coolability, which determine the containment Thermal-Hydraulics (TH) response and can cause its late failure. Despite their importance from a safety point of view, chemical and physical processes behind these phenomena are not yet well understood, partly because of their extreme complexity and partly because of the limitations of current experimental capabilities in reproducing them. As result, models developed and implemented in the codes to describe the MCCI and debris coolability are extremely simplified because based on limited data availability.The benchmark study consists of a comparison of the results calculated by the ASTEC and MELCOR codes during the ex-vessel phase following two different scenarios a Complete Station Blackout (CSBO) and a Small Break Loss Of Coolant Accident (SBLOCA).For each scenario, two different Severe Accident Management Strategies (SAMS) are adopted, the first one is to ensure that the corium, once fully relocated in the cavity, has top-flooding conditions throughout the transient, while the second one is to keep the reactor cavity completely dry. Furthermore, for every case reproduced the reference and best-practice models are investigated.The results show that in case of dry cavity conditions, despite the different VF timing calculated by the two codes, the final results provide very similar indications even though some dissimilarity is observed regarding the amount of fission product (FP) released through the Containment Filtering Venting System (CFVS).In contrast, in the case of wet cavity conditions, the accident progression predicted is significantly different regarding number of venting operations, TH containment conditions, and amount of concrete ablated.The discrepancies observed, should be sought in the corium and debris coolability models, which given their parametric nature and oversimplification can lead to these results.

Investigation of small break LOCA with failure of the emergency core cooling system accident management in VVER-1000 reactor type

In this study, the thermohydraulic condition of the power plant is estimated as a result of the Small Break Loss of Coolant Accident (SB-LOCA) for Bushehr Nuclear Power Plant (BNPP) and the effect of control measures in postponing severe events as well as increasing the time available to enter control measures has been investigated. Two scenarios including control measures with and without operator action have been considered, considering the progress of the accident and the lack of effective intervention of control measures, both scenarios lead to the melting of the reactor core. The critical times for entering additional measures and restoring systems has been obtained from the progress of the accident for two scenarios that is used to develop an accident management strategy. The thermohydraulic calculation in this article deals with the estimation of time of critical point for accident management measures at a different phase of SB-LOCA.

Full scope 3D analysis of a VVER-1000 containment pressurization during a LB-LOCA by employing AutoCAD and GOTHIC code

One of the main goals of severe accident management strategies is to maintain containment integrity to prevent radioactive release into the environment. As a result, containment internal loads need to be investigated during a postulated accident. In the present study, short-term containment thermal-hydraulic (TH) parameters of a VVER-1000/V446 reactor are analysed during a LB-LOCA. To achieve this goal, in the first step as-built 3D structure of VVER-1000/V446 containment was modelled in detail by using AutoCAD. The AutoCAD model has been processed to be prepared for GOTHIC 3D input. Meanwhile, an equivalent GOTHIC lumped parameter (LP) model is also prepared to validate the modelling procedure and results against the FSAR. Finally, LP profiles and 3D TH contour results of GOTHIC code were presented and discussed. LP results show a close agreement with FSAR reference and can approve the accuracy of the simulation procedure. 3D contours present all-coordinate detailed TH parameters versus time inside the containment. Spatial distribution of TH parameters and minor short-term effects of spray as Engineering Safety Features (ESFs) to tackle the containment pressurization can be evaluated by employing these contours. 3D simulation results can provide advantages for the precise locating and installation of ESFs in design and operation to achieve the highest efficiency in case of containment accident.

Evaluation of aerosol retention inside a steam generator under steam generator tube rupture accident conditions

The necessity has arisen to conduct risk evaluations of the steam generator tube rupture (SGTR) accident for domestic operating nuclear power plants in the Republic of Korea in accordance with a recent amendment to the Nuclear Safety Act. To evaluate the aerosol-type radionuclides released to the environment during the accident and to prepare mitigation strategies, tests conducted in experimental facilities that faithfully reflect the properties of domestic plants are essential. In this work, one such experimental facility was installed at the Korea Atomic Energy Research Institute with scaling work considering the geometrical shapes of a domestic Korean plant and the major aerosol removal mechanisms inside the steam generators. A preliminary test was first conducted to understand the amount of aerosol removal inside the pipes of the facility with analytical methods. Then, aerosol retention tests were performed in both dry and flooded conditions. Decontamination factors were evaluated in each test, and the results were analyzed using filter measurements, an electrical low-pressure impactor, and deposition mass in the tubes. The test results are believed to provide a good starting point to improve the related accident management strategies to increase the safety of domestic Korean nuclear power plants.

Potential Impact of Reactor Core Damage on Severe Accident Management Actions in Vicinity of Spent Fuel Pool

Fukushima Daiichi NPP accident showed that plant technical support center (TSC) in an extreme and rare external event (design extended condition (DEC)) can have a problem in the case of coincident loss of decay heat removal from the core (possibly resulting in significant core damage) and loss of decay heat removal from spent fuel pool. From the point of view of prioritizing severe accident management strategies it looks like the priority mitigation action should be to reestablish the emergency core cooling in the reactor pressure vessel. The reason is the longer time window available before the water inventory in the spent fuel pool would be evaporated and spent fuel exposed to overheating. However, if such actions would not be successful and reactor core would, consequently, be damaged, potential design basis leakage (or even greater leakage) from the containment to the fuel handling building (FHB) can affect already established TSC measures or operator accessibility to FHB, or it can jeopardize functioning of the systems, structures and components due to radioactive releases and presence of hydrogen (independently of the fact that containment atmosphere can be inerted by steam or that containment may be equipped with passive autolytic recombiners (PARs)). Paper describes an engineering evaluation of possible hydrogen presence in the containment annulus, its flammability and leakages through the penetrations toward FHB in the case of long term station blackout (SBO) without successful restoration of the core cooling in the reactor pressure vessel. SBO accident sequence progression and amount of produced hydrogen is evaluated by MAAP code.

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

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.

SAFETY OF UKRAINE’S NUCLEAR POWER INDUSTRY IN EXTREME OPERATING CONDITIONS

An urgent issue of the state (level) of safety of Ukraine’s nuclear power industry in extreme conditions is the situation of the Zaporizhzhia NPP, the largest in Europe, due to the station’s location in the war zone and difficulties in its management and operation, as well as the safety regulation of the station in the occupied territory. In the proposed article, the possibility of Zaporizhia NPP industrial site flooding is analyzed from the elaborated hydrodynamic model and justifying calculation; the conditions of ZNPP industrial site possible flooding are determined depending on the warhead power producing a corresponding destructive effect. Probabilistic approaches to assessing the Zaporizhia NPP safety objective state (level) in extreme conditions are insufficiently justified taking into account the lessons of the largest nuclear and radiation accidents. On the basis of a deterministic approach, the conditions of Zaporizhia NPP industrial site critical flooding due to extreme military action are identified. Flooding of the plant’s industrial site may cause two initial emergency events: a complete long-term de-energization of power units and violation of heat exchange conditions in spent nuclear fuel dry storage facilities. Prevention of flooding of the Zaporizhia NPP industrial site can be based on the construction of protective flood barriers on the cooling pond coast. Prospective directions to increase the accident management efficiency related to the nuclear power units’ passive safety systems modernization and improvement of operational symptom-oriented emergency instructions/manuals for managing severe (nuclear) accidents have been identified. An effective accident management strategy with complete long-term de-energization of WWER power units can be based on a comprehensive modernization of emergency recharge systems of steam generators with steam driven pumps and natural circulation circuits of passive heat removal systems from the pressurized reactor unit.