Nuclear Power Plant Accident Research Articles

Development of failure mitigation technologies for improving resilience of nuclear structures

After the Fukushima daiichi nuclear power plant accident, various countermeasures were taken for Beyond Design Basis Events (BDBE) in the system safety field. These included portable devices, additional backup facilities and accident management. They are different from approaches for Design Basis Events (DBE). In the field of structural mechanics; however, efforts were focused on strengthening to prevent failures for both DBE and BDBE in the same way. This approach will lead to limitless requirements for strength and expensive plants.As a breakthrough approach in structural mechanics for BDBE, we propose failure mitigation methods through the application of passive safety structures, where preceding failures release loadings and mitigate subsequent failures. When preceding failure modes have small impacts on safety performance, such as small deformation and crack initiation, and subsequent ones are catastrophic modes such as collapse and break, the passive safety structure improves safety and resilience. This idea is the utilization of passive characteristics of structures without additional equipment and electric power, allowing for simple and reliable plants.To demonstrate this idea, passive safety structures were applied to next-generation fast reactors, subject to high temperature and low-pressure conditions. In the case of loss-of-heat-removal accidents, high temperature conditions accelerate the creep deformation of structures. When deformation redistributes loadings and reduces stresses at important positions such as coolant boundaries, progression to creep rupture of boundaries can be mitigated. When an excessive earthquake occurs, plastic deformation and buckling become dominant, due to low pressure and, therefore, a thin-wall structure. The above-mentioned failure modes reduce rigidity and natural frequency. When the natural frequency becomes lower than the input frequency, vibration energy is hardly transferred to structures and the subsequent failures of structures, such as collapse and break, are mitigated.

Simulation of late phase phenomena with the system code AC2

Simulation codes like the AC2 system code package developed by German Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) gGmbH are used in international reactor safety research to investigate hypothetical severe incidents and accidents in nuclear power plants. Valuable impetuses for the development of prevention and mitigation measures for such events can be derived from the analyses. This article presents selected work by PSS on the development of simulation models for the late phase of nuclear accidents. Topical issues as the formation and thermodynamic interactions of debris beds in the lower plenum of the reactor pressure vessel (RPV), the release of fission products as well as the interaction of the molten core material with concrete in the reactor cavity in case of RPV failure are discussed. The analyses carried out provide important findings regarding the validation of the developed models. In addition, future development needs are derived for the consideration of further relevant phenomena.

A condition-informed dynamic Bayesian network framework to support severe accident management in nuclear power plants

In the nuclear industry, Severe Accident Management Guidelines (SAMGs) are developed with the objective of providing prescriptive actions for mitigating the consequences of Nuclear Power Plant (NPP) accidents. These guidelines are defined with respect to a set of expert-based prototypical severe accident scenarios and are symptom-based, i.e., they relate to the values of specific monitored plant parameters. Expert-based SAMGs might fail to deal with non-prototypical scenarios, i.e., scenarios not considered a priori by the experts. To address this issue, in this paper we propose a methodology based on Dynamic Bayesian Networks (DBNs) to support operators decision-making for avoiding the escalation of severe accidents to non-prototypical scenarios. The methodology is based on the integration of condition monitoring data into a DBN to allow it to i) infer the current Plant Operating Condition (POC), Initiating Event (IE), and NPP safety barriers Damage State (DS), ii) predict the grace time (GT) available to carry out mitigative actions, and iii) evaluate the effectiveness of the alternative candidate mitigative actions taking into account the safety barriers estimated DS and the predicted available GT. The proposed methodology is exemplified on a Loss of Coolant Accident (LOCA) in a WWER-1000.

Fukushima Daiichi Nuclear Power Plant Accident: Understanding Formation Mechanism of Radioactive Particles through Sr and Pu Quantities.

The Fukushima Daiichi Nuclear Power Plant accident released considerable radionuclides into the environment. Radioactive particles, composed mainly of SiO2, emerged as distinctive features, revealing insights into the accident's dynamics. While studies extensively focused on high-volatile radionuclides like Cs, investigations into low-volatile nuclides such as 90Sr and Pu remain limited. Understanding their abundance in radioactive particles is crucial for deciphering the accident's details, including reactor temperatures and injection processes. Here, we aimed to determine 90Sr and Pu amounts in radioactive particles and provide essential data for understanding the formation processes and conditions within the reactor during the accident. We employed radiochemical analysis on nine radioactive particles and determined the amounts of 90Sr and Pu in these particles. 90Sr and Pu quantification in radioactive particles showed that the 90Sr/137Cs radioactivity ratio (corrected to March 11, 2011) aligned with core temperature expectations. However, the 239+240Pu/137Cs activity ratio indicated nonvolatile Pu introduction, possibly through fuel fragments. Analyzing 90Sr and Pu enhances our understanding of the Fukushima Daiichi accident. Deviations in 239+240Pu/137Cs activity ratios underscore nonvolatile processes, emphasizing the accident's complexity. Future research should expand this data set for a more comprehensive understanding of the accident's nuances.

Riverine 129I dynamics during high-flow events on the Abukuma River in Fukushima

This study presents the riverine 129I concentrations and fluxes during high-flow events. The river water samples obtained by a previous study were subjected to 129I analyses. River water samples were collected at the midstream of the Abukuma River for five and seven times during high-flow events in July 2018 and October 2018, respectively. Suspended solids and filtered water samples were measured for 129I/127I ratios using an accelerator mass spectrometer system and for 127I concentrations using ICP-MS. Aggregated mean values of dissolved 129I concentration and 129I concentrations in suspended solids were 0.11 μBq/L (n = 12) and 0.60 mBq/kg (n = 12), respectively. Corresponding values of 129I/127I ratios were in 2.2 × 10−9 (n = 12) and 3.7 × 10−9 (n = 12), respectively. These concentrations and ratios were comparable to those at the background level before the Fukushima Dai-ichi nuclear power plant accident. Positive correlations with dissolved 137Cs and Cl− concentrations suggested that the dissolved 129I concentration decreased due to dilution of the river water during the events. A positive correlation with total carbon content suggested that organic matter contents determine 129I concentrations in suspended solids. The total 129I fluxes during JUL18 and OCT18 were 9.9 × 103 and 2.2 × 104 Bq, respectively. Although dissolved 129I was predominant under low-flow conditions, the particulate 129I flux increased by one or two orders of magnitude during the peak water discharge phases. These results underline the importance of particulate 129I dynamics during high-flow events for quantitative evaluation of the 129I cycle.

DNA Damage and Repair in PBMCs after Internal Ex Vivo Irradiation with [223Ra]RaCl2 and [177Lu]LuCl3 Mixtures.

The combination of high and low LET radionuclides has been tested in several patient studies to improve treatment response. Radionuclide mixtures can also be released in nuclear power plant accidents or nuclear bomb deployment. This study investigated the DNA damage response and DNA double-strand break (DSB) repair in peripheral blood mononuclear cells (PBMCs) after internal exposure of blood samples of 10 healthy volunteers to either no radiation (baseline) or different radionuclide mixtures of the α- and β-emitters [223Ra]RaCl2 and [177Lu]LuCl3, i.e., 25 mGy/75 mGy, 50 mGy/50 mGy and 75 mGy/25 mGy, respectively. DSB foci and γ-H2AX α-track enumeration directly after 1 h of exposure or after 4 h or 24 h of repair revealed that radiation-induced foci (RIF) and α-track induction in 100 cells was similar for mixed α/β and pure internal α- or β-irradiation, as were the repair rates for all radiation qualities. In contrast, the fraction of unrepaired RIF (Qβ) in PBMCs after mixed α/β-irradiation (50% 223Ra & 50% 177Lu: Qβ = 0.23 ± 0.10) was significantly elevated relative to pure β-irradiation (50 mGy: Qβ, pure = 0.06 ± 0.02), with a similar trend being noted for all mixtures. This α-dose-dependent increase in persistent foci likely relates to the formation of complex DNA damage that remains difficult to repair.

Social networks and posttraumatic stress symptoms five to ten years after the Fukushima nuclear power plant disaster

Since the 2011 Fukushima nuclear power plant accident, multiple social network disruptions have been reported among the community in Fukushima, while posttraumatic stress symptoms among the residents have persisted. In this study, we aimed to explore the influence of time and social networks on the recovery of posttraumatic stress symptoms based on longitudinal data from community residents in Fukushima, following up five to ten years after the nuclear power plant accident. We conducted five questionnaire surveys quasi-annually, the targets of which were randomly sampled 4900 non-evacuee community residents. In this study, the data of 1809 respondents who participated in at least one survey were used (36.9% of the initial target). Setting posttraumatic stress symptoms as the outcome, we examined the interaction between time and social network size using a mixed model, adjusting for sociodemographic characteristics and disaster-related events. Their interaction was statistically significant, and the posttraumatic stress symptoms of those with small social networks persisted, while those with larger social networks recovered. Maintaining and promoting social networks may contribute to mental health recovery after a nuclear disaster.

Fukushima Daiichi nuclear power plant impact on regional economies from 1960 to 2010

Understanding the economic and social impact of nuclear power on a region is crucial; this is particularly important for Fukushima because the region relies on nuclear power and its sustainable recovery following the nuclear accident at Fukushima Daiichi. However, existing studies do not reveal any consensus on the economic impact of nuclear power plants. Further, long-term time-series data for conducting such an analysis are lacking; so, for this study, a long-term database from 1960 to 2010 was constructed (before the earthquake and nuclear power plant accidents happened in 2011). The data were analyzed to evaluate the effects of the nuclear power plants on employment in various sectors in Fukushima, using the difference-in-differences (DID) and staggered DID methods. Unlike previous studies, this study revealed that nuclear power surprisingly increased agricultural employment (892.62 more jobs) while reducing manufacturing employment (−466.40 fewer jobs) due to crowding-out effects. Furthermore, although employment in the construction sector also increased, the impact was temporary. These findings highlight the complex economic dynamics triggered by the presence of a nuclear power plant, which may have significant implications for long-term recovery strategies in the region. Despite the positive impacts on certain sectors, the negative effects on manufacturing underscore the need for careful consideration in future economic planning and recovery efforts. The study's findings provide valuable insights into the challenges and opportunities faced by regions hosting nuclear power plants, which must be considered for long-term recovery.

Methodology for analyzing dose consequence using atmospheric dispersion code A2CDOSE

Abstract The experience of Fukushima brings out the importance of the emergency response plan during or after a severe nuclear power plant (NPP) accident. To deal with the accident emergency response situation, the National Atomic Research Institute (NARI) in Taiwan developed an emergent radiological dose evaluation system with the Central Weather Administration (CWA) called “A2CDOSE”. A2CDOSE can evaluate the eight-day projected dose after a nuclear accident and provide the predicted dose consequences for decision-making. Nevertheless, a source term generation method was still needed during the accident. This study chose MELCOR as the comparison target for the source term simulation issue with RASCAL. A MELCOR calculation of Maanshan NPP was built and a 5 % containment leakage was set for both MELCOR and RASCAL source term calculations. It seems the calculations of noble gases were consistent between MELCOR and RASCAL. However, the Cs-137 and I-131 calculated by MELCOR were twice compared to the RASCAL results, which may have a chance to cause different calculations of public protective actions during an accident. After the source term analysis, several atmospheric dispersion cases were simulated by A2CDOSE and RASCAL for further comparison.

Improving the estimation of direct release rates and transport processes from the Fukushima Daiichi Nuclear Power Plant accident using higher-resolution oceanic dispersion model

A series of accidents at the Fukushima Daiichi Nuclear Power Station (F1NPS), following the 11 March 2011 earthquake and tsunami, resulted in the release of radioactive substances into the ocean. In particular, the release of large amounts of radioactive caesium has damaged the fishing industry, leading to voluntary restrictions on fishing and shipping. Oceanic dispersion simulations based on estimates of the pathways and fluxes of radioactive materials provide useful information for assessing the environmental impacts and formulating measures to mitigate the effects of the accident. For the direct release rate from the F1NPS site, an estimation method was developed using the results from nearby monitoring, and the seawater exchange rate was estimated in target volume using a numerical simulation. However, the influence of volume on the seawater exchange rate was not considered. Appropriate volumes must be considered when estimating the effects of future accidents. In addition, the directional coastal transport was underestimated in the simulations of the F1NPS accident because of the low resolution. To estimate the pathways and fluxes of radioactive material to the ocean and understand the distribution of the concentration of radioactive material based on ocean dispersion simulations, a study was conducted using a higher-resolution model. The horizontal resolution of the conventional ocean dispersion model Regional Ocean Modelling System (ROMS), was increased from 1 km to 200 m. The optimal settings of the seawater exchange rate were investigated, and the radioactive caesium transport process in the coastal direction was more accurately reflected. We found that the conventional volume for determining the seawater exchange rate, including the locations of release sources and observation points, is optimal. The validity of this finding was confirmed using experimental equations from previous oceanic tracer release experiments. To estimate future release rates, it an appropriate volume must be defined, for example, depending on the distance between the locations of the release sources and the observation point. In addition, improvements in the accuracy of the simulation of the coastal transport process were observed owing to the higher resolution, which increased reproducibility. However, with a horizontal resolution of 200 m, problems with repeatability near the harbours arose. A higher resolution, achieved using nesting or other methods, would be desirable to deal with releases smaller than those in the F1NPS accident.

Quality and safety of products using GM soybeans in the diet of bulls

The production of high-quality and safe food products in areas contaminated by radiation due to the Chornobyl Nuclear Power Plant accident remains a pressing issue. The aim of this study was to determine the concentration and transfer factors of 137Cs, Pb, Cd, Cu, and Zn in bull products when including various high-protein feeds, such as narrow-leaved lupin and genetically modified soybean, in their diets. The study employed zootechnical methods (design and conduct of a scientific-economic experiment on animals), radiological and spectrometric methods (determination of the concentration of 137Cs, Pb, Cd, Cu, and Zn in feeds, longissimus dorsi muscle, liver, kidneys, and testes of bulls), and statistical methods. It was established that when using GM soybean in the grain mixture, the concentration of 137Cs in the muscle tissue of animals in Group II (experimental) decreased relative to the control by 2.57 Bq/kg, or by 40.5% with a statistically significant difference between groups (P > 0.99). Furthermore, the transfer factor of the radionuclide into the muscle tissue of animals in Group II compared to Group I also had 2.15% abs. lower. The content of Pb in the muscle tissue of bulls in both experimental groups exceeded the maximum permissible concentration by 2.52-2.78 times, while the concentration of Cd in the longissimus dorsi muscle and Cu in the liver of animals in Group II also exceeded sanitary-hygienic requirements by 10.0% and 2.0%, respectively. The introduction of 40% (by mass) of maize grain and 30% of GM soybean into the grain mixture instead of the same amount of wheat and lupin chop during the fattening of animals in the third zone of radioactive contamination contributed to significantly higher content and transfer of heavy metals into muscle tissue: Pb – by 10.3% and 0.10% abs., Cd – 10.0 and 0.03, Cu – 17.2 and 0.53, and Zn – by 17.2% and 2.40% abs. The results of the research can be used in agricultural enterprises and private households that produce livestock products in areas with a high anthropogenic load

System theory safety analysis of network malfunction in nuclear power plant distributed control systems

Abstract A secure and highly reliable communication network is a crucial component of the distributed control system (DCS) in nuclear power plants (NPPs), particularly for the protection system. However, there have been numerous reported cases of cyber-attacks in the field of nuclear power, and the cybersecurity threats faced by a DCS are increasing. Given that the DCS is a complex integrated system combining computers, communication, and control with a large number of interactions between components and external interferences, traditional linear safety analysis methods fail to meet its requirements. To minimize cybersecurity accidents caused by potential vulnerabilities, it is necessary to conduct systematic safety analysis on DCS cybersecurity accidents in NPPs. Therefore, this paper selected a specific case of cybersecurity accidents in a NPP for abstract modeling and reproduction. The causal analysis based on system theory (CAST) method was applied to conduct an analysis of the accident causes. It can identify existing vulnerabilities and potential factors impacting the safe operation of NPPs, and countermeasures and recommendations can be proposed. The reference for assessing potential risks and devising solutions can be provided for other DCS communication networks.

Development of a MELCOR Model for LVR-15 Severe Accidents Assessment

LVR-15 is a light-water-tank-type research reactor placed in a stainless-steel vessel under a shielding cover located in the Research Centre Rez (CVR) near Prague. It is operated at a steady-state power of up to 10 MWt under atmospheric pressure and is cooled by forced circulation. In 2011, the fuel was replaced, going from high-enriched uranium (HEU) to low-enriched uranium (LEU). After 2017, the State Office for Nuclear Safety (SUJB) asked CVR to evaluate the LVR-15 under Design Extended Conditions B (DEC-B). For this reason, a new model was developed in the MELCOR code, which allows for modelling the progression of a severe accident (SA) in light-water nuclear power plants and estimating the behaviour of the reactor under SA conditions. The model was built by collecting information about the LVR-15. Since the research reactor can have different core configurations according to the location of the core components, the core configuration with the most fuel (hottest campaign K221) was selected. Then, to create the radial nodalisation, the details of the core components were obtained and grouped in five radial rings and 27 axial levels. The simulation was run with the boundary conditions collected from campaign K221, and the results were compared with the reference values of the campaign with a negligible percentage of error. For the coolant inlet and outlet temperature, the reference values were 318.18 K and 323.5 K, respectively, while for the simulation, the steady state reached 319 K for the inlet temperature and 324 K for the outlet temperature. Additionally, the cladding temperature of the hottest assembly was compared with the reference value (353.72 K) and the steady-state simulation results (362 K). In future work, different transients leading to severe accidents will be simulated. When simulating the LVR-15 reactor with MELCOR, specific attention is required for the aluminium-cladded fuel assemblies, as the model requires some assumptions to cope with the phenomenological limitations.

Methodology for estimating the contribution of forest fires in loss of offsite power events

Forest fires are disasters caused by natural or human factors and have environmental as well as economic impacts such as loss of biodiversity, agricultural damage, and property damage. According to statistics, the number of forest fires and the area damaged are gradually increasing in Korea. In addition, there are cases where the units 1, 2, and 6 of the Hanul Nuclear Power Plant (NPP) have already been affected by forest fires. Korea has recently initiated research on forest fires under multi-unit probabilistic safety assessment (PSA) project. This paper deals with the impact on the loss of offsite power (LOOP), which is one of the NPP incidents and accidents affected by forest fires. To estimate the frequency of LOOP induced by forest fires, the PSA methodology for earthquakes, a representative external event, was referred.

Distribution of 137Cs specific activity in river sediments of the Barents Sea basin (Nenets Autonomous Okrug, Russian Arctic).

This article focuses on the study of the distribution of 137Cs in the bottom sediments of Arctic rivers of the Barents Sea basin (using the example of the Nenets Autonomous Okrug, Russian Arctic). This research is relevant due to the poorly studied region and the significant number of radiation-hazardous facilities in the Arctic zone of Russia, both those currently in operation and those that are "nuclear heritage sites". The study of 137Cs specific activity in bottom sediments was carried out in the period from 2020 to 2023 in the rivers Chizha, Nes, Vizhas, Oma, Pechora (river delta), as well as the rivers Kolva and Usa (first and second order tributaries, respectively, of the Pechora River). A total of 199 samples were collected. In addition to 137Cs specific activity, the samples were analysed for sediment particle size distribution, organic matter content, carbonate content and ash content. The 137Cs specific activity mainly ranged from the minimum detectable specific activity to 5.4 ± 0.8Bq·kg-1. In the Nes River basin (Kaninskaya tundra), the 137Cs content in bottom sediments reached 36.0 ± 3.2Bq·kg-1 (in the case of lake sediments) and 22.9 ± 3.7Bq·kg-1 (in the case of river sediments), values that are higher than those of the North-West of Russia. Considering the large area of the study area (Kaninskaya tundra, Pechora river delta, southern part of Bolshezemelskaya tundra) and the similarity of physical and chemical parameters of the studied rivers, it is possible to assume the existence of a zone of increased radionuclide content in the Nes river basin. This may be due to the runoff from the Nes River catchment area, its hydrological features, and the accumulation of 137Cs in the small fractions of bottom sediments. The results confirm the conclusions of previous soil studies in the Nes river basin. The main sources of elevated 137Cs content are global atmospheric deposition and the Chernobyl Nuclear Power Plant accident.

Evaluating the Overall Quality of Online Information on Nuclear Power Plant Accidents in Japanese

Information on professionally produced materials is beyond the reading level of the average adult and is difficult to read and understand. Additionally, the readability of nuclear emergency preparedness manuals in Japan is not clear. The purpose of this study is to evaluate the understandability, actionability, readability, and quality of nuclear emergency preparedness manuals. A systematic quantitative content analysis of online resources was conducted from November 27 to December 8, 2023. The selection criteria were websites in Japanese related to nuclear disaster prevention and aimed at the general public. The PEMAT-P, jReadability, Journal of the American Medical Association benchmark criteria, and Global Quality Scale were used to evaluate each web page. The number of respondents who scored 70 or higher on the PEMAT-P was 71 (61.2%) for Understandability and 57 (49.1%) for Actionability. Regarding the difficulty level of the texts by jReadability, "Difficult" was the most frequent response with 82 responses (70.7%). The JAMA Benchmark Criteria was 35.3% in 41 patients for attribution, 80.2% in 93 patients for currency, and a mean GQS of 4.1 (SD=1.0). This study quantitatively evaluated the understandability, actionability, readability, and quality of nuclear disaster prevention materials for the Japanese public. The findings suggest that text pertaining to nuclear disaster prevention materials may not be easily understood because it is beyond the reading comprehension of the average adult.

Attitudes toward future child-rearing among male and female adults without children in different prefectures at 11 years after the Fukushima Daiichi Nuclear Power Plant accident

This study aimed to determine whether confidence in future childbearing and child-rearing in relation to radiation exposure differs among people without children in Fukushima Prefecture, its neighboring prefectures (Tochigi, Ibaraki, and Miyagi), and other prefectures, and among men and women at 11 years after the nuclear accident. In February 2022, a cross-sectional online survey was conducted using an Internet research company. A total of 985 people were surveyed: 240 from Fukushima, 305 from neighboring prefectures, and 440 from other prefectures. The proportion of women who were confident about future childbearing was low and the confidence was associated with radiation-related risk perceptions in both Fukushima and neighboring prefectures. The results suggest that the more people want to have children in the future, the more likely they will be able to discuss radiation, and the less likely they will be able to believe that radiation has a low impact on their partners. Therefore, it is necessary to create an environment where people can openly discuss radiation, and to examine factors that affect confidence in future childbearing and child-rearing by gender and prefecture.

Uncertainty-Based Validation Methodology and Experimental Analysis for External Control Room Human Performance Simulation: Application to Fire Probabilistic Risk Assessment of Nuclear Power Plants

In probabilistic risk assessment (PRA) of nuclear power plants (NPPs), human reliability analysis (HRA) is conducted to identify potential human failure events that could contribute to risk scenarios and estimate human error probabilities. Lessons learned from the 2011 Fukushima Daiichi NPP accident underscored that for PRA, it is critical to model external control room (Ex-CR) human actions. The state-of-practice HRA methods, historically developed for the main control room HRA, are limited in capturing the unique nature of Ex-CR human actions, such as location dependence (in addition to the time dependence) of human actions and spatiotemporal interactions of human performance with the surrounding physical environments, for instance, hazard propagation. To advance the Ex-CR HRA in the context of the fire PRA for NPPs, the authors’ team developed a simulation-based fire crew performance model using an agent-based modeling (ABM) technique. The ABM fire crew simulation was coupled with a fire progression model through a spatiotemporal interface using a geographic information system. This paper focuses on the validation of the ABM simulation, which is the key requirement for the simulation-based Ex-CR human performance model to be utilized in PRA. The existing validation approach, initially developed for physical models in the fire PRA of NPPs, is extended for validation of the simulation-based Ex-CR human performance model. Model uncertainty is used as a measure of model validity, which facilitates the incorporation of the validation result into the PRA. The degree of the model uncertainty is characterized by a lognormal error model whose parameters are quantified based on a pairwise comparison between empirical data and model predictions. The proposed validation approach is demonstrated using a case study of the fire PRA of NPPs. This study makes two research contributions: (1) it is the first to validate the simulation-based Ex-CR human performance model against empirical human performance data and incorporate the validation result into the PRA of NPPs, and (2) this study, for the first time, conducts a controlled experimental test to collect empirical data for fire crew performance at NPPs.

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.

Chemical implication of the partition coefficient of 137Cs between the suspended and dissolved phases in natural water

After the Fukushima Daiichi nuclear power plant accident, the terrestrial environment became severely contaminated with radiocesium. Consequently, the river and lake water in the Fukushima area exhibited high radiocesium levels, which declined subsequently. The partition coefficient of 137Cs between the suspended sediment (SS) and dissolved phases, Kd, was introduced to better understand the dynamic behavior of 137Cs in different systems. However, the Kd values in river water, ranging from 2 × 104 to 7 × 106 L kg−1, showed large spatiotemporal variability. Therefore, the factors controlling the 137Cs partition coefficient in natural water systems should be identified. Herein, we introduce a chemical model to explain the variability in 137Cs Kd in natural water systems. The chemical model includes the complexation of Cs+ with mineral and organic binding sites in SS, metal exchange reactions, and the presence of colloidal species. The application of the chemical model to natural water systems revealed that Cs+ is strongly associated with binding sites in SS, and a major chemical interaction between 137Cs and the binding sites in SS is the isotope exchange reaction between stable Cs and 137Cs, rather than metal exchange reactions with other metal ions such as potassium ions. To explain the effect of the SS concentration on Kd, the presence of colloidal 137Cs passing through a filter is significant as the dominant dissolved species of 137Cs in river water. These results suggest that a better understanding of stable Cs dissolved in natural water is important for discerning the geochemical and ecological behaviors of 137Cs in natural water.