Core Meltdown Research Articles

High-pressure core meltdown and hydrogen generation during In-vessel and Ex-vessel phases of VVER-1000 nuclear reactor and effect of heat structures

Following unprecedented Fukushima Daiichi severe accident, which significantly amplify the production of hydrogen and the associated risk of explosions, particularly in accidents like station blackout (SBO) that jeopardize containment integrity and reactor safety, a thorough investigation and comprehension of the phenomenon known as high-pressure melt ejection (HPME) are necessary. This study examines how the SBO accident develops without operator intervention or access to emergency diesel generators through full nuclear power plant simulation, including containment, core, primary, and secondary circuits. Accurate time step selection is crucial for the simulation to produce reliable results for high-pressure melt ejection processes. While the process is faster and easier under Low Pressure Melt Ejection (LPME) conditions, time step tuning can significantly lengthen the (HPME) simulation (approximately one month for each scenario). After thorough examinations and analysis of various simulation findings, three cases were evaluated for the ex-vessel phase, illustrating the most likely scenarios of high-pressure molten material splashing into the cavity. The findings demonstrate that during the in-vessel phase, 578 kg of hydrogen is created. Additionally, during the ex-vessel phase, there is a significant increase in the containment’s temperature and pressure due to the high-temperature hydrogen and significant water and steam leakage. The temperature and pressure inside the containment rise well above the safety criteria due to the leakage of 1529 kg of hydrogen from the cavity and lead to hydrogen explosion due to MELCOR message. The calculated amount of hydrogen gas is within the explosion limit range of 18.3% to 59% of the containment atmospheric pressure volume. Moreover, the height and diameter of the cavity alter due to the direct heating and corrosion of the cavity walls, ultimately compromising the integrity of the plant in the alpha mode. Based on the results obtained, it can be seen that if a large portion of the molten material is deposited on the concrete surface (case III) instead of dispersed in the cavity atmosphere (case I), the rate of hydrogen production and therefore the total amount of hydrogen (and its related energy) produced will be increased and could lead to deflagration to detonation situation.

Improved Heat Transfer Capabilities of Nanofluids—An Assessment Through CFD Analysis

AbstractConventional fluids used in fission‐based water‐cooled nuclear reactors have lower heat transfer coefficients (HTCs) and thermal conductivity, which has led researchers to explore high‐performance fluids that can enhance heat transfer in routine operation and prevent core meltdown in the case of accidents. It is important to investigate a wide range of fluids that can help designers improve thermal hydraulic characteristics, such as HTC, critical heat flux, and minimum departure from nucleate boiling ratio (MDNBR). In this study, the effectiveness of nanofluids in enhancing heat transfer parameters, including thermal conductivity and heat capacity, was investigated. Four different nanofluids (Al2O3–H2O, ZrO2–H2O, Ag–H2O, and Si–H2O) with pure water as the primary coolant in an HPR‐1000 nuclear reactor were compared using computational methods. Due to computational limitations, only the flow channel among four fuel rods with the highest power density in the core was simulated using Eulerian computational fluid dynamics. The results of this study show that silver water (Ag–H2O) nanofluid outperformed other nanofluids and pure water. It had a higher average HTC and MDNBR, with a 67.15 % and 45.23 % improvement, respectively, compared to pure water. The fuel rod wall temperature was also reduced by 28.5 K with Ag–H2O compared to water. Comparison of current simulated results with literature data shows a good agreement.

Numerical simulation of the dynamic interaction characteristics between Wood’s metal and water under coolant injection mode

The accident of the steam generator tube rupture in the lead-bismuth-cooled fast reactor or the core meltdown in the light water reactor may cause violent heat and mass transfer between the melt and the coolant, potentially further damaging the reactor. Therefore, it is essential to study the corresponding mechanisms. In this study, numerical simulations are utilized to investigate the dynamic interactions between Wood’s metal and subcooled water in the coolant injection (CI) mode, focusing on jet behavior and critical cavity parameter changes. By analyzing these factors, this research aims to clarify the interfacial evolution and interaction mechanisms between the melt, water, and air, thereby tackling the problem of inadequate observation in experiments. A user-defined function (UDF) was employed to simulate the flow characteristics of the melt during the solidification process. This approach effectively addressed the limitation of Fluent’s inherent solidification model, where the fluid no longer participates in turbulent velocity calculations after solidification. In addition, an equivalence method was proposed according to the characteristics of two-dimensional simulations in Fluent. Comparisons with Cheng’s experimental results show that the equivalent model can accurately predict jet penetration characteristics while significantly reducing the demand for computational resources. According to this equivalence method, this research developed a new correlation for predicting the maximum jet penetration depth in the CI mode. The above results provide new perspectives for understanding the dynamic interactions between metals and coolants, which is significant for the application and optimization of technologies in the nuclear field.

Effects of different load variation rates on the operational stability in an S–CO2 Brayton cycle coupled natural circulation lead-cooled fast reactor

The thermal inertia results in net heat absorption in natural circulation lead-cooled reactors during power variation processes, increasing the risk of core meltdown. In this study, a dynamic simulation model of a natural circulation lead-cooled fast reactor coupled with a supercritical carbon dioxide Brayton Cycle power generation system has been developed based on Vpower. The influence of thermal inertia on system operating temperature under different power variation rates was discussed. The results indicate that during power reduction, different decline rates only affect the temperature variation range of the system without impacting the final stable state. However, during power increase, different ramp rates not only affect the temperature variation range but also lead to varying degrees of system temperature rise. To address this issue, an optimized turbine throttling scheme is proposed. Compared to the original plan, the new plan successfully stabilizes the temperature at the rated value after increasing the system power. This study provides insights for the operation control strategy of similar reactor types.

Experimental and numerical research on the debris bed formation in severe accidents of nuclear reactors

The formation and coolability of the debris bed play a crucial role in the core meltdown process in nuclear reactors. Previous studies mainly focused on experiments involving interactions between small masses of molten material and water, with emphasis on the steam explosion process, lacking analytical data on debris beds and fragments. Therefore, the formation of the debris bed is studied by visual experiments, along with the relevant models are established and applied in this study. In the experiments, a large mass of zirconium dioxide is heated to a molten state in the crucible and then released into a water tank through the release device. The water tank is equipped with four glass windows and high-speed cameras to capture snapshots.Firstly, experiments on the formation of hundred-kilogram-scale debris beds were conducted using the prototype material, zirconium dioxide. The molten material temperature was 3000 K, the molten material mass ranged from 50 to 100 kg, the jet diameter varied from 10 to 20 mm, the water pool depth ranged from 0.55 to 1 m, and the water subcooling varied from 10 to 80 K. Visualized images of the debris bed formation were obtained, and an experimental database of debris bed formation was established, primarily including key parameters such as debris morphology, debris size, debris bed accumulation form, and debris bed porosity. The results indicate that the rise in the mass of molten material and the decrease in water subcooling increase the proportion of large-sized fragments. The experiments did not form a “cake” debris bed but instead achieved loose debris beds due to the small jet diameter. With an increase in the jet diameter and a decrease in water subcooling, the porosity of the debris bed will also increase. Subsequently, computational analysis codes were developed for the processes of jet break-up, debris bed accumulation, and formation, and finally, the deviation of the predicted accumulation morphology was less than 20 %. The results reveal that the influence of the jet diameter and mass of the molten debris on the deposition characteristics of the debris bed is very pronounced. Experimental research and code development hold significant reference value for the formulation of preventive and mitigative measures for severe accidents in nuclear power plants.

Study on the lower head failure in severe accidents Part II: Thermal-mechanical coupling behavior analysis on CAP1400 reactor lower head

Under severe accident conditions with core meltdown, large amounts of molten material relocate to the lower head at high temperatures. Significant thermal and mechanical loads may lead to failure of the reactor pressure vessel (RPV) lower head, spreading radioactive fission products. Lower head failure time and location are crucial for accident mitigation and consequence assessment. This study developed a thermodynamic failure analysis model for the lower head by investigating material failure mechanisms, coupling this into an Integrated Severe Accident Analysis code (ISAA), and conducting experimental validation and applied research. Part I mainly introduces the thermodynamic analysis module development, validation against OLHF experiments, and preliminary failure criteria analysis. Results indicate the model accurately predicts lower head failure behavior, and Larson-Miller life fraction and Hedl-Dorn parameter criteria accurately assess lower head integrity. Part II presents applied research after validation and preliminary analysis. Using the improved ISAA, melt pool behavior and lower head thermal–mechanical response were analyzed for a CAP1400 small break loss-of-coolant accident (SBLOCA), assuming multiple safety system failures. Results indicate the bottom of the lower head melt pool didn’t form a hard shell due to high heat release rate, only an oxide shell layer around the second layer. Decay heat from melt pool components cannot be dissipated, leading to lower head failure. Further External Reactor Vessel Cooling (ERVC) heat transfer model analysis and research are needed in ISAA. The two developed mechanical analysis models predict material failure at high temperatures, and the timing and location of lower head failure are consistent, at approximately 16,000 s at 25.38°±5.47° on the lower head. The predicted failure mode is similar to Part I validation, indicating both Larson-Miller life fraction and Hedl-Dorn parameter criteria provide consistent lower head integrity judgments.

Investigation of molten fuel coolant interaction with simulated corium in sodium

In case of a hypothetical core meltdown accident in Sodium cooled Fast Reactor (SFR), corium interacts with cold pool sodium and results in Molten Fuel Coolant Interaction (MFCI). During MFCI, the corium fragments and resultant debris settle on the core catcher as a bed and continue to generate decay heat. The size and state of fragmented debris play a crucial role in post-accident heat transfer. An experimental study has been undertaken to investigate the fragmentation of simulated corium in sodium using real-time X-ray imaging. Experiments were conducted by releasing ∼400 g of molten mixture of alumina and iron at ∼2400 °C into ∼4 kg of sodium at various temperatures. The fragmentation phenomenon was acquired using a digital flat panel detector and the captured images were subjected to image analysis to extract information on melt fragmentation. Post-experiment, fragmented debris was retrieved and subjected to sieve analysis to obtain characteristics towards assessing the dominant fragmentation mechanism. Experiments were also conducted with water for identifying key differences in the fragmentation and debris characteristics. Non-energetic fragmentation, generation of fine debris, and relatively minor sodium vaporization observed in the small-scale experiments present a reduced potential for energetic interaction in sodium than that of water.

Research on thermal hydraulic characteristics of reflooding process based on core multi-channel model in system analysis code

There are four main stages in PWR Loss of Coolant Accident (LOCA): blowdown, re-irrigation, re-flooding and long-term cooling. The re-flooding stage is a critical phase to avoid core meltdown and achieve long-term cooling. RBHT reflooding experiment was simulated with the RELAP5 and COSINE program respectively.Thermal-hydraulic characteristics in re-flooding phase were analyzed using multi-channel model. Quench time is mainly influenced by system pressure and coolant subcooling. Later quench will result in a higher cladding temperature. Comparatively, the peak cladding temperature is less affected by the power of the heating rod and the re-flooding rate, which fluctuates within 7 %. The higher re-flooding rate allows quenching to occur as fast as possible and improves cooling efficiency. However, higher coolant sub-cooling and re-flooding rates may cause greater thermal stresses on the core and endanger core integrity. Based on the analysis of the re-flooding rate on the peak cladding temperature and quench time, the optimal re-flooding rate interval of 0.099–0.105 m/s was obtained.

The main structural-phase states of interaction between model corium of a nuclear reactor and a sacrificial material based on Al2O3 and a lead layer

This article presents the results of post-experimental studies of the interaction of model light water reactor corium with a candidate sacrificial material (SM) based on Al2O3 (corundum) and a lead layer. The purpose of the experiments was to study the features of the interaction of the sacrificial material with corium in the conditions of simulating a severe accident with core meltdown. An active interaction of corium with the sacrificial material was established as a result of the research. This is confirmed by the fact that part of the sacrificial material entered into physical and chemical interaction with the components of corium with the formation of liquid-phase reaction products and, as a result, creating different conditions for the crystallization of corium. The study of the phase composition showed that the microstructure of solidified corium in various areas after interaction with the sacrificial material is represented by a phase of the (U,Zr, …)O2±x type with different types of crystal lattice. At the same time, the phase analysis showed a close correspondence of the composition of the solid solutions to the initial ratio of uranium and zirconium in the model corium charge. This allows us to conclude that liquid zirconium interacted with aluminum oxide during the experiment. Thus, it was concluded that the proposed sacrificial material is promising based on the results obtained and the identified features of the interaction of aluminum oxide with the lead layer and the corium of a nuclear reactor.

Coupled AC2-CFD simulations for a high-pressure core melt accident scenario

Abstract Even though very unlikely to occur, severe accident scenarios in nuclear power plants have to be analyzed. During high-pressure core meltdown scenarios in a pressurized water reactor the primary circuit should fail first. Previous analyses found that a free convection flow within the vertical steam generator (SG) tubes with a simultaneous stratified gas counterflow in the hot legs could arise. This phenomenon leads to higher thermal loads on individual SG tubes which might then fail leading to a containment bypass and the release of radioactive material into the environment. Lumped parameter system codes used for safety analyses do not provide the models necessary to simulate phenomena like mixing in three-dimensional flows and could not consider local turbulence effects. Computational fluid dynamic (CFD) codes provide such capabilities but are much more computationally expensive. Coupling of a system code with a CFD code can therefore be used to simulate such phenomena. The advantages of both approaches can be maximized by splitting up the simulation domain between the codes, depending on the expected flow conditions. The system code AC2 coupled with the CFD code OpenFOAM was used to simulate part of the severe accident transient. Free convection in the hot leg and the U-tubes of the vertical SG was observed in case of high-pressure severe accident sequences. The thermal load of individual SG tubes has been estimated from the results. These loads can be used as inputs for structural-mechanical analyses to estimate which part of the primary circuit would fail first.

CFD analysis of iodine mass transfer coupled with a chemical reaction in a filtered containment venting system

The removal of iodine from flue gases through FCVS during a core meltdown in nuclear reactors is important for environmental safety. In this study, CFD analysis for iodine removal in a venturi scrubber was done on Ansys Fluent using a mathematical model that coupled the mass transfer of iodine with the reaction between iodine and thiosulfate. A user defined function (UDF) was written to couple the mass transfer with the reaction. The UDF was hooked into Ansys Fluent, and the simulation was run for three different gas flow rates. Contours for the mass fraction of iodine, tetrathionate, and iodide, along with the hydrodynamics of the venturi scrubber, were analyzed at different gas flow rates. The mass transfer rate of iodine and its overall removal efficiency were observed to increase at higher gas velocities. The results satisfied the experimental data at a high gas flow rate of 340 m3/h but overpredicted the efficiency at lower flow rates of 240 m3/h and 290 m3/h.

Flow boiling heat transfer and pressure fluctuation characteristics in a divergent channel with inclined downward-facing heater

The core catcher cooling system is used in nuclear power plant to mitigate the core meltdown accident through flow boiling heat transfer. In order to better understand the flow boiling characteristics in the system, experimental research was conducted on an inclined channel. The flow channel is oriented 15° with a divergent downward-heating surface. The two-phase flow behavior and pressure fluctuations were capture by a high-speed camera and pressure transducer. The inlet mass flux and subcooling effects on flow boiling characteristics were deeply investigated. The inlet mass flux and subcooling conditions are within 100–400 kg/m2s and 1–30 K, respectively. The results showed that the boiling curves shift to the higher wall superheat with the increase of mass flux, and the effect of subcooling is weak. The bubble dynamic images by the high-speed camera showed that the bubbles continue to grow and coalesce during the sliding process along the wall. Large bubble blankets are formed periodically upon the boiling surface beyond a heat flux threshold, and they would undergo a rapid condensation process after departing the heating area, which leads to the liquid reversal and accompanied by strong pressure fluctuations. Further, beyond a certain subcooling and heat flux, two-phase flow instability phenomenon appears, which is characterized by complete condensation of bubble blanket accompanied by pressure shocks, meanwhile, the heat transfer coefficient decreases and the wall superheat increases sharply. The bubble blanket frequency could be obtained by the Fast Fourier Transform analysis of pressure, which is positively correlated with the heat flux and is negatively correlated with the mass flux and subcooling. Furthermore, a dimensionless correlation of reduction factor (RF) was proposed, which directly related the bubble blanket characteristics to the thermal–hydraulic parameters.

Development of a full-scale MATCH model in heat transfer performance simulation of RPV insulation structure with experimental validation

In a core meltdown accident, the reactor pressure vessel (RPV) insulation plays a significant role in the flooding process of the reactor cavity. The research on the function role of RPV insulation has attracted lots of attention whereas there are relatively few studies on heat transfer characteristics. Besides, the majority of previous studies on RPV insulation adopted a down-scaling or partial model instead of the entire model to decrease computational complexity. In this paper, a mesh assembly method is proposed to generate block-structured meshes and the heat transfer characteristics of the RPV insulation is carried out. Then a partial validation model has been verified against the experiments results. In addition, the effects of the air supply temperature, flow rate, and insulation installation gap on the overall model heat transfer performance of RPV insulation are investigated. The results demonstrate that the mean temperature of the insulation's outer wall (42.92 °C), the local temperature of the concrete (73.85 °C), and the heat flux of the insulation's outer wall (93.48 W⋅m‐2) are less than their limit values of 60 °C, 95 °C, and 235 W⋅m‐2, respectively. And the heat leakage from the supports accounts for over 90% of the total heat dissipation.

Experimental investigation on creep behaviors and life prediction across phase-transformation of thermal aged 16MND5 steel

Reactor pressure vessel (RPV) is one of the key equipment for nuclear power plant, and long-term exposure to high temperature is an important factor causing its performance deterioration. Under core meltdown conditions, RPV steel experiences high temperature creep due to the In-Vessel Retention (IVR) mitigation measures. This paper aims to investigate high-temperature mechanical properties, especially for the creep behaviors of 16MND5 steel after thermal aging at 650 °C for 2000 h. The results showed that the yield stress and ultimate tensile strength of the 16MND5 steel are decreased, and the creep rupture time is rapidly shortened after thermal aging. The microstructure was analyzed by scanning and transmission electron microscopy (SEM, TEM). 16MND5 steel showed a typical bainitic phase before thermal aging. After 2000 h of thermal aging, the micro-grains were coarsened, and finally changed into a ferritic microstructure. Based on Larson-Miller (LM), Orr-Sherby-Dorn (OSD) parameters methods and Kachanov-Rabotnov (K–R) model to predict uniaxial creep life after thermal aging, the prediction results are satisfactory, and all within the double tolerance band. This work reveals the creep damage mechanism of 16MND5 steel after thermal aging, and provides basic data for the implementation of IVR strategy in nuclear power plants after long-term service.

Model for core catcher ablation and cavity shape in the film regime

During core meltdown in a nuclear reactor the corium may propagates toward the lower head and hit the vessel or structural elements as a coherent liquid jet. This could occur for instance in sodium fast reactors (SFR) if discharge tubes and an in-vessel core-catcher are used as mitigation devices. This can result in the ablation of the core-catcher and potential loss of its integrity. During ablation, a cavity is formed within the solid. Therefore the ablation of a solid structure by a liquid jet is studied here, with a focus on the cavity shape to improve future core-catcher designs. During the cavity formation, liquid flows over the cavity being formed as a liquid film in the film ablation regime, and ultimately results in the formation of a liquid pool. In the film ablation regime, two types of cavity shapes have been identified previously. To describe the process of the cavity formation, a mathematical model valid in this film ablation regime is here proposed which links the local melting rate to the local curvature of the cavity. It is applied with two sets of assumptions applicable to the two types of cavity previously identified. For the first assumption set, the heat transfer coefficient varies along the liquid/solid interface while the liquid temperature is held constant, it is representative of laminar boundary layer growth in laminar film. For the second assumption set, the heat transfer coefficient is held constant while the temperature varies along the liquid/solid interface, this is consistent with a turbulent liquid film. The two modeling approaches are compared with experimental data acquired from the HAnSoLO experimental setup. Numerical predictions show promising agreements with experimental observations, particularly for the laminar boundary layer growth zone of the cavity.

KIT reactor safety research for LWRs: Research lines, numerical tools, and prospects

This paper describes the main research lines of the Karlsruhe Institute of Technology (KIT) relevant to Light Water Reactors (LWR) that permit both the safety evaluation of different reactor designs and the prediction of the radiological source term including its impact in the case of hypothetical severe accidents. The numerical tools for core analysis, nuclear power plant analysis of design basis and severe accidents, and the estimation of the radioactive dispersions of fission products after a core meltdown accident are described and selected results are discussed. The analytical investigations at KIT are complemented by experimental investigations at two facilities, namely the COSMOS-H/-L and the QUENCH facilities devoted to safety-relevant Thermal-Hydraulic (TH) phenomena and in-vessel severe accident early-phase for different kinds of fuel rods including Accident Tolerant Fuel (ATF). These data are crucial for the validation of the numerical tools used for safety demonstration. The paper is complemented by a brief description of the in-house tools under development for improved core/plant analysis based on the multi-physics/-scale coupling methodologies, where the in-house tools are combined with external system TH, Monte Carlo, and thermo-mechanic code to increase the prediction accuracy of the core/plant behaviour under stationary and accidental conditions. Selected results are presented and discussed. It can be stated that the long-lasting safety research activities at KIT devoted to LWR led to the development of a computation route combining external with in-house codes that is very much appropriate for safety evaluations and risk assessment; it includes unique multi-physics/-scale coupled tools for improved core and plant analysis of different reactor designs, e.g. PWR, VVER, BWR, SMR, and research reactors.

Experimental Simulation of Decay Heat of Corium at the Lava-B Test-Bench

During the development of a severe accident at a nuclear power plant (NPP), corium is formed—a melt of core materials. A distinctive feature of corium, due to the content of fuel elements in its composition, is the presence of decay heat, which makes a significant contribution to the nature of the interaction of the corium melt with the structural materials of the reactor plant. In this regard, the decay heat should be taken into account when conducting computational studies and physical experiments. For this reason, certain requirements are imposed on the methods of simulating decay heat in the corium prototype, which relate to both the uniformity of the volume distribution and its intensity. This paper presents the results of calibration experiments to substantiate the operability of the induction heating system of the Lava-B test bench, which is used to simulate decay heat in the study of processes occurring during an accident with the NPP core meltdown. So, in order to obtain optimal characteristics of the heating system, a series of experiments was conducted on heating the graphite block in the experimental section of the Lava-B test bench. In the experiments, the capacitance of the used oscillating circuit capacitor banks and the electrical power on the inductor varied. As a result of the analysis of the data obtained, the most optimal parameters of the inductor-load simulator system were determined. In general, the performed experiments confirmed the operability of the induction heater and the possibility of its use in experimental studies of the interaction of corium with the various structural elements of the NPP reactor core at the Lava-B test bench.

Experimental investigation on debris bed formation from metallic melt coolant interactions

Motivated by risk quantification of severe core meltdown accidents which may occur light water reactors, an extensive DEFOR (DEbris FORmation) test series with various molten oxidic materials was conducted previously at Royal Institute of Technology to investigate the characteristics of debris beds formed from oxidic melt-water interactions, which are important to the coolability of debris beds and thus the retention of core melt. Yet, little attention has been paid to metallic melt which exist in the core melt. The present study is intended to fill in the knowledge gap, i.e., to characterize metallic melt-water interactions when a melt jet falls into a deep water pool. Nine DEFOR-M tests are carried out to clarify the effects of melt jet diameter, free fall height and water pool depth on metallic debris formation characteristics. Molten tin of 20 kg is employed as the simulant of metallic melt. A melt sensor and a mass sensor are installed to detect the discharging period of coherent jet and the mass accumulation of debris bed, respectively. A high-speed camera is applied to record the process of jet breakup, debris fragmentation and solidification, and debris bed formation. The experimental results show that the water pool depth has a significant influence on steam explosion and debris bed formation characteristics. Moreover, the jet breakup length is sensitive to jet diameter and free fall height. Finally, the features of metallic debris beds are compared with those of oxidic ones in respects of debris bed's configuration and porosity, particles' morphology and size distribution.

A Study on Radiological Hazard Assessment for Jordan Research and Training Reactor

Numerical simulations of atmospheric dispersion and dose assessment were performed for the Jordan Research and Training Reactor (JRTR) to evaluate its radiological effects on surrounding population and the environment. A three-dimensional atmospheric dispersion model was applied to investigate the behavior of the radionuclides released into the air, and a dose assessment model was used to estimate the radiological impact on the population residing in nearby cities around the JRTR. Considering full core meltdown an accidental scenario, most of the source term was assumed to be released from the JRTR. Simulations were performed to calculate the air and deposition concentrations of radioactive materials for July 2013 and January 2014. The monthly averaged values of concentrations, depositions, and dose rates were analyzed to identify the most harmful effects in each month. The results showed that relatively harmful effects occurred in January 2014, and the total annual dose rate was estimated to be approximately 1 mSv outside the 10 km radius from JRTR. However, the impact of a nuclear accident is not as severe as it might seem, as the affected area is not highly populated, and appropriate protective measures can significantly reduce the radiation exposure. This study provides useful information for emergency preparedness and response planning to mitigate the radiological consequences of a nuclear accident at the JRTR.

Experimental simulation of selected design extension condition scenarios without core meltdown in the CODEX facility

Two DEC-A scenarios were simulated with electrically heated 7-rod bundles. The test parameters were selected on the basis of reference scenarios from VVER-440 safety analyses.The CODEX-SBO experiment simulated a station blackout event in two separated phases. In the first phase the arrival of cold water from hydroaccumulators stopped the further heat-up of the fuel rods. In the second phase the heat-up continued up to 1900 °C and the test was terminated with the injection of emergency coolant from the recovered active injection system.In the CODEX-NITRO experiment it was supposed that after emptying of hydroaccumulators nitrogen enters the primary coolant and the test was focused on the interaction of hot fuel rods with nitrogen. The presence of nitrogen at high temperature (up to 1700 °C) resulted in the acceleration of chemical reactions, including nitriding of zirconium components.