From the viewpoint of energy security in Japan and the reduction of the environmental load, the continuous operation of light water reactors is essential. Since a pH adjuster with enriched Li-7 ions is required for water quality control on PWR, the development of Li-7 enrichment technology is one of the key issues. The multi-channel counter-current electrophoresis (MCCCE) method has been developed as a technology with a low environmental load. To put this method into practical use, it is necessary to understand the behavior of Li-7 ions in the channel flow and optimize the experimental condition to separate Li-7 and its isotope. In this paper, to understand the behavior of Li-7 ions in a single channel of the experimental apparatus, a numerical simulation method based on a computational fluid dynamics (CFD) code with a particle tracking method, TPFIT-LPT, was developed. In the method, the motion of each representative ion in the flow is evaluated using the particle as a mass point moving along the flow under an electric field. The difference in the isotopes was represented by changes in the magnitude of the added velocity. We also considered that although it is impossible to measure the behavior of each ion, it is important to measure the flow velocity of the bulk fluid for the validation of the numerical simulation. We developed a lab-scale experimental apparatus in which the single channel of the actual apparatus was simplified to measure the flow velocity by Particle Image Velocimetry (PIV). We set a pulsation flow condition on the lab-scale experiment, which is one of the difficult conditions for the numerical simulation and measured the velocity. As the result, we confirmed that the pulsation flow was reproduced. We set the measured data as the inlet boundary condition of the numerical simulation and conducted it. As the numerical result, we confirmed that the ions affected by the electric field moved upstream with pulsation. We also confirmed the effect of the electric field on the motion of the ions.

To explore different radiation heat transfer calculation model on temperature field distribution of argon space, and assess the accuracy of the calculation model, this paper established a pool type sodium cooled fast reactor argon space geometry model, respectively, by using the DOM radiation model and surface to surface radiation model for numerical simulation to get the temperature field under different radiation models based on the STAR - CCM + simulation software. The results show that when the absorption coefficient is 0, there is little difference between the DOM radiation model and the surface-to-surface radiation model. When the absorption coefficient is not 0, the larger the absorption coefficient is, the higher the temperature of the argon chamber is. Therefore, the selection of the radiation model is very important for the numerical simulation calculation of the temperature field of the fast reactor argon space.

About the boiling transition (BT) that determines the maximum thermal output of the BWR, it is considered that the spacers have significant effects on the occurrence of the BT. The occurrence conditions of the BT can be changed by devising the spacer shapes because it will affect to entrainment and deposition behaviors of droplets. In the light water cooled fast reactor: RBWR, thermal-hydraulics conditions are more challenging than in the current BWR. Then, the effect of the spacer on the BT should be sufficiently utilized in the RBWR. In the thermal-hydraulics design for the current BWR, large-scale tests were carried out and used to evaluate BT conditions. The RBWR is still in the design stage, and there is room to be changed many parameters. Then, it is not reasonable to determine the shape of the spacer only by large-scale tests but also by considering local effects on droplet entrainment and deposition.

On the other hand, by applying a two-phase CFD method with remarkable development in recent years, we can develop a model that can predict the effect of the spacers mechanistically. This research used the detailed two-phase flow simulation code TPFIT developed by JAEA to simulate annular dispersed flow in RBWR subchannels. In the occurrence of the BT, it is considered that the two-phase flow pattern is the annular dispersed flow, and we want to evaluate the effects of the spacer on annular dispersed flow in the RBWR subchannels. We performed numerical simulations of annular dispersed flow in the simplified subchannel of the RBWR. As a simulation parameter, we choose the existence of the spacer. The spacer in the simulation has a simplified shape and the same blockage ratio as the RBWR. In addition, we analyzed numerical simulation data and identified each droplet's occurrence and disappearance points. Based on these data, we evaluate the numbers of entrainment and deposition distribution in and around the spacer.

In this paper, the rewritten boiling model is embedded into Fluent in the form of UDF to perform numerical calculation of low-pressure boiling. In order to ensure the correctness of the rewritten boiling model, this paper uses the experimental data of Lee experiment to verify the correctness of the rewritten boiling model. The results show that the rewritten boiling model agrees well with the experiment. In addition, this paper only uses Fluent to carry out the numerical simulation of Lee's experiment, but has never obtained a numerical solution. Therefore, using UDF to rewrite the boiling model can improve the convergence of Fluent when calculating low-pressure boiling.

Most fluid flow in nuclear reactors is chaotic or turbulent because of high-temperature operations. And forced convection plays a leading role in normal operating conditions. Whereas natural convection becomes an important role when circulating pumps or fans fail. To deeply understand the natural convection phenomenon, studies of the phenomenon with simple geometries are prevalently devoted by many researchers. Also, this paper aims to investigate and report the flow properties of chaotic natural convection occurring in a square cavity. The investigation is performed by direct numerical simulation (DNS). The flow properties are focused on the chaotic flow patterns and overall heat transfer through the cavity at the Rayleigh number of 1010. The instantaneous and statistical results of the DNS are exhibited and compared with each other. Furthermore, the results are wished as a part of databases for validating new codes simulating chaotic flows.

Heat exchanger tubes of steam generator in nuclear power plants are prone to damage and failure under long-term flow-induced vibration, which threatens the safe operation of reactors. In order to study the flow-induced vibration behavior of heat exchanger tubes, a single rigid cylinder was taken as the research object, based on large eddy simulation (LES), the HHT-α method was used to solve the structural dynamic equation, and combined with the local dynamic mesh deformation technology, the flow-induced vibration calculation model of heat exchanger tubes under transverse flow was established. Firstly, through the calculation of flow around a fixed cylinder with Re=3900, the influence of the grid discretization on the prediction ability of flow field characteristics was studied, and the CFD model suitable for the flow-induced vibration calculation was obtained. Secondly, the CSD model suitable for flow-induced vibration calculation is obtained by solving the structural dynamics equations with different methods. Finally, based on the established fluid-structure interaction calculation model of fluid-induced vibration of heat exchanger tubes, the vibration response characteristics, the motion trajectory and the wake mode of three-dimensional cylinder fluid-induced vibration at different reduced flow velocities were studied. The results show that the flow-induced vibration model of the heat exchanger tubes can accurately predict the flow-induced vibration behavior of a three-dimensional cylinder under transverse flow. The relevant research is of significance for reference to establish the flow-induced vibration calculation model of tube bundle structures under transverse flow.

Small miniature lead-cooled fast reactors use assemblies with hexagonal cross-sections as their fuel assemblies. There are gaps of a few millimeters between the two assemblies. Under normal operating conditions, the flow and heat transfer are usually ignored in these gaps. However, the heat transfer capability of the core by-pass narrow slit flow cannot be ignored under accident conditions. Although lead-cooled fast reactors have received less research, this narrow slit flow has been confirmed in studies using sodium-cooled fast reactors. As a result, its heat transfer capacity needs to be determined to assure core safety in the event of an accident. In order to simulate the narrow-slit flow in the core of a small miniature lead-cooled fast reactor and study the flow state, temperature, and velocity distribution of the narrow-slit flow in the core within the natural circulation flow range, a core model with only interstitial channels are constructed, and a hole is cut in the bottom of the enclosure. According to the results, the core exhibits clear temperature stratification in the height direction and a significant temperature gradient at 4% power; the average flow velocity inside the narrow slit declines as power increases, and when the core power is below 10%, fluid inside the narrow slit gap contributes more to the core cooling capacity; and a clear vortex or transverse flow appears inside the narrow slit.

The spent fuel dry-storage system of the HTR-PM600 in China is consist of six regions, with each composed of multiple wells. The spent fuel canisters are located vertically in the well. To accurately describe the flow resistance characteristics of the storage well, a 1/5 scaled model experiment is under considering. To completely align with flow similarity criterion, the characteristic velocity of the scaled model must be much higher than the full model, Which leads to an unfeasible pressure that a fan can provide. In this paper, the flow characteristic comparison of the full and 1/5 scaled models of the storage well was investigated to identity the characteristic velocity for the following scaled experiment using CFD code and flow self-similar theory. Firstly the flow resistance characteristic of single canister was studied with CFD code, and the canister was simplified to a simple item suitable for resistance calculation. Then a well model that conforms to the real situation was built up to simulate the flow resistance characteristic. With the calculation of different condition, a curve about the relationship of pressure and Reynolds number(Re) can be drawn. The characteristic velocity was chosen as low as possible in an acceptable range of errors to meet the pressure requirements of the fan. The result shows that within a specific interval of Re, the change of flow resistance characteristic is acceptable. The minimum value of the interval will be used in the following experiment.

The helical coil once-through tube steam generator (HCOTSG) has a high application value in integrated modular small reactors due to its compact structure and high heat transfer capability. In this paper, the effect of the number of spiral tube layers on the shell-side flow heat transfer is investigated by numerical simulation, and the velocity, temperature, and pressure distribution characteristics of the shell side of different layers of spiral tubes are compared and analyzed. The study shows that the spiral tube bundle shell side will produce complex vortex flow; with the increase of the number of layers, the flow distribution gradually uniforms, the flow velocity decreases, and the heat transfer capacity also gradually weakened, when the number of layers increased to 8 layers tends to stabilize. The pressure drop per unit length increases with the increase in the number of layers. This study will provide theoretical support to simplify the design of the structure of the HCOTSG shell-side thermal-hydraulic experiment body.

Radionuclides can be released into the terrestrial environment from nuclear facilities through particulate matter or gases. Radionuclide deposition in flora and soil is the initial step in their transfer into the terrestrial environment and food chain. Two main depositional processes remove pollutants from the atmosphere. Wet deposition refers to the transfer of a substance from the atmosphere to the ground in snow, hail, or rain. Dry deposition refers to the direct transport of gases and particulate matter to and absorption by natural surfaces such as vegetation.

The percentage of radionuclides that are first retained by vegetation is referred to as interception, and it is defined as the portion of radionuclides that are deposited by both wet and dry deposition. In any radioecological model, the fraction of radionuclides initially intercepted is a very substantial quantity, although the retained radionuclides are eventually weathered into the soil. This is because direct deposition might result in relatively high radioactive quantities in feed and meals. In this paper, a review of various mathematical models in understanding the processes involved in the transport of radionuclides following foliar uptake by plants is studied. However, in Estimating the concentration of radionuclides in food and assessing doses in humans, the systemic transport of radionuclides is critical as this is especially important for plants where only certain parts are used as food or feed, such as cereals and potatoes.

As stated in Management of Disused Sealed Radioactive Sources, DSRS management schemes typically include transferring to other authorized users or returning them to the manufacturers/original suppliers or sending them to a licensed storage operator or disposal operator for long-term storage, disposal, reuse, and recycling. Near-surface disposal, borehole disposal, and geological disposal are all choices.

The purpose of this study is to choose the preferred disposal options for DSRS through management options comparison and prove the safety of the preferred disposal options for DSRS through safety assessment implementation on the preferred disposal option to protect individuals, society, and the environment from radiation exposure. In line with the objectives, Borehole Disposal System is suggested after careful analysis of other disposal systems through management options. The safety assessment of DSRS disposal is carried out with the development of two scenarios that are the Basic design with time scenario and the radionuclide release to the biosphere as a defect scenario. The dose constraint of 0.3msv/yr is used as a benchmark for results comparison with a clear definition of the site characteristics, the inventory, the mathematical models for each scenario, and the disposal facility structure. The software Ecolego which is based on Matlab and Simulink is used to perform the safety analysis.

The results for the safety analysis after critical simulation with the software are mainly focused on the dose limit for humans, the maximum dose released from the radionuclides is 0.063mSv/yr for the basic design with time and 0.066mSv/yr for the radionuclides released to the Biosphere scenarios.

This report indicates that the inventory of DSRS can be safely and permanently disposed of using BDS.

Digitalization has the high potential to make decommissioning safer and more cost-effective. IAEA calls digitalization for decommissioning "Digital Decommissioning (DD)" ^[1][2][3]and conducts the development of DD. The DD proposed by IAEA manages information on decommissioning facilities using the latest IT technology and has BIM （Building Information Management) ^[4], which innovates the construction industry, as its core function and consists of three parts based on functions. The main part of DD consists mainly of BIM. One of the other two parts is the input part for the data-related decommissioning facility, and another is the output part for information related to decommissioning plan and practice, which is generated and edited by the DD. The information would play a very effective role in explaining to the regulator and promoting understanding among stakeholders.

The reliability of the information is premised on grasping the facts of the decommissioning facilities correctly. The input part for grasping the facts consists of four kinds of data input functions. these are the function of handling legacy data of the facility, the function of 3D data on layout and structure obtained on-site by optical scanning technology, the function of measurements and sampling of radiation dose and radioactivity, and the calculation function of radioactivity distribution in the facility calculated by calculation codes. The individual attributes of the data obtained by each function are closely correlated or the same among the four functions. Accurate and effective data acquisition is possible by effectively using relationships between attributes. Also, to utilize the input data through BIM, it is necessary to clarify the relationship with the BIM^[4] record format, which is called IFC (Industry Fundamental Classes) ^[5].

In this presentation, we would report on the development of both the data record format based on the analysis of the data attributes required by the four input functions, and the data interface based on the investigation and analysis of the relationships between the data attributes, with considering the regulatory framework and operator's management unique to Japan. In addition, we would also report on the extension of IFC to utilize these data in BIM.

This study simulated the long-term monitoring data of the hydraulic head around the Horonobe Underground Research Laboratory (URL) during the construction and operation to validate the hydrogeological model developed by the previous study. Compared to the previous study, the longer-term simulation and the additional information on the groundwater flow by the excavation progress during the simulation period enabled the reliable evaluation of the effective hydraulic conductivity of low permeable domain in the deep subsurface around the URL. The coverage of the ranges or reproduction of the trends of the longer-term monitoring data by simulation in this study validated the hydraulic conductivity model developed by the previous study. In addition to the validation of the model, the predictive simulation was also the scope of this study for the planned excavation of the URL to 500 m. The hydraulic head above the depth of 400 m was estimated to already reach the steady state now, whereas the hydraulic head below the depth would continue to decrease slightly. The inflow into the shaft was also estimated to reach the steady state now. These simulation results and the theoretical consideration of the observed hydraulic features in the field predicted that the significant change in the hydraulic head and the increase of the long-term inflow would not occur by the further excavation of the URL.

The purpose of this study is to predict the drying rate and the temperature and water distributions in the packed bed of porous particles in a water-containing state. We made a physical model of the drying process in a cylindrical bed and examined the heat and mass transfer resistances required for it. In this paper, the incorporation of these resistances into the CFD parameters and the comparison results between the predicted and experimental values of the evaporation amount were presented. The calculation result of the amount of evaporation reproduced well the experimental value. The wall heat transfer coefficient and the evaporation rate coefficient were found to improve the reproducibility of the calculations. Although the calculated temperatures underestimated the experiments, the tendency shown by the experiments was reproduced. The validity of the physical model was confirmed. Next, it was confirmed that the liquid saturation in the bed decreased from the periphery to the center of the bed. In addition, the liquid saturation on the vapor permeability had a large effect on the vapor flow distribution in the bed. Furthermore, it was found that the effect of liquid saturation on the effective thermal conductivity in the bed cannot be ignored.

The decommissioning of the Fukushima Daiichi nuclear power plant presents a great challenge in terms of waste management owing to the presence of various radionuclides and waste materials, including large amounts of radioactive sediment waste from the advanced liquid processing system (ALPS). The durability of the waste form is crucial for the safety of final disposal. Phosphate waste (ALPS phosphate) is a candidate material for stabilizing the carbonate and iron co-precipitation slurry wastes from ALPS. To enhance waste form durability, we propose using a composite form that encases primary waste, the ALPS phosphate in a secondary metallic matrix for protection from corrosion under a repository environment. The simulated ALPS phosphate (main composition of Ca:Mg:Fe:P=2:3:1:5, doped Cs, Sr, Eu, and Ce) was synthesized, using hot isostatic pressing (HIP), a powder metallurgy method, at 1000 °C and 175 MPa for 3 h and a composite waste form was obtained. Metallographic measurements and elemental analyses revealed that the powdery stainless steel was well sintered, and transformed into an ingot matrix, whereas the ALPS phosphates and their decomposition products were individually dispersed and confined within the metal matrix. According to a simple corrosion model, the composite waste form described is expected to have a long enough lifetime to ensure a safe, final disposal.

Advanced Thermal Reactor “FUGEN” is the heavy water-moderated, boiling light water-cooled, pressure tube-type reactor. The commercial operation of this reactor started in 1979 and terminated in 2003. We have been dismantling the facilities based on the decommissioning program since 2008.

Except for reactor core and heavy water system, the periphery system of FUGEN is similar to the BWR.

Therefore, the results of decommissioning in FUGEN will be reflected to the decommissioning of commercial reactors in the future. In Japan, we play an important role as a pioneer for decommissioning of the other reactor owned by electric power companies.

Since it is important to reduce the amount of low-level radioactive waste generated as a result of decommissioning, an attempt has been made since 2008 to apply clearance system to a portion of this waste.

Subsequently, decontamination and measurement data associated with the clearance were compiled and evaluated. As a result, we confirmed that the clearance operation of the turbine facility has started and has been progressing smoothly since then. We will obtain decontamination and measurement data in order to reflect them to the next own works and other decommissioning plants [1].

JENDL-5 released in 2021 includes enough cross section data for neutron activation calculations. Thus we have produced ORIGEN and ORIGEN-S libraries from JENDL 5 for activation calculations in nuclear facility decommissioning. We also produced the similar libraries from JENDL/AD-2017 for comparison. The ORIGEN and ORIGEN-S calculations for JPDR with these libraries demonstrated that the libraries were produced adequately.

The measurement of α/β radioactive surface contamination level is one of the necessary steps in the radioactivity monitoring work, and the measurement result is also an important basis for determining the contamination. For workplaces prone to radioactive contamination, routine contamination monitoring on the ground and equipment surface should be carried out regularly, which is very important to prevent contamination diffusion and reduce unnecessary exposure of personnel to radiation. At present, most of the α/β surface contamination monitors are portable (hand-held or vehicle-mounted) monitoring devices, but there are the following two problems: First, the detection area of portable monitoring device is small, resulting in low monitoring efficiency and time-consuming, which unable to meet the needs of the surface contamination monitoring in the entire area of radioactive sites. Second, after the nuclear accident, the level of surface contamination in radioactive sites is relatively high, and direct measurement of personnel will inevitably result in different degrees of radiation damage. Therefore, in order to solve the above problems, an intelligent robot for surface contamination monitoring has been developed. Based on an array of large-area surface contamination detectors, combined with positioning and navigation (SLAM) and wireless transmission technology, it has realized the autonomous, rapid and remote intelligent monitoring of surface contamination in the entire area of radioactive sites. The device not only reduces the unnecessary radiation to personnel, but also improves the level and intelligence of radiation monitoring device.

The numerous constraints associated with the management of Intermediate and High-Level Waste (ILW & HLW), with potentially complex physical-chemical compositions, led to investigate the implementation of in-situ conditioning treatment able to produce waste packages compliant with existing storage and disposal routes or on-site interim storage facilities. The issues involved in the management of this kind of waste can vary widely, and thermal treatment solutions seems logical given that these offer multiple advantages. Vitrification processes enable volume reduction, chemical stabilization of the waste, and efficient containment of radioelements in a glassy or glass-ceramic matrix. Moreover, vitrification processes have proven their adaptability to intermediate and high-level waste and can be flexible enough to accommodate a varied waste stream.

In this context, the DEM&MELT In-Can process has been developed to match the requirements and constraints of challenging ILW and HLW. DEM&MELT is an innovative compact vitrification tool that can deal with a wide range of nuclear waste streams with different compositions and forms such as slurries, deposits, liquid effluents, ashes, adsorbents... DEM&MELT is flexible enough to accommodate uncertainties in waste composition and has been developed with a modular design adaptable to nuclear operators’ needs. The process allows a significant volume reduction in addition to safe radionuclides containment with moderate investments and operating costs.

A major milestone was achieved with the commissioning of a full-scale pilot in CEA Marcoule site and the implementation of several demonstration tests on various waste streams. The tests results confirm that the technology can achieve high waste loading and provides durable containment of the radionuclides as well as stabilization of the waste. The tests carried out also confirmed the simplicity, robustness, and versatility of the process.

This paper presents some results of demonstration tests performed on various waste with an emphasis on the process parameters, the waste loading achieved, the wasteform properties and the radionuclides volatility. The up-scaling methodology implemented to demonstrate the applicability of the DEM&MELT process to an identified waste, from laboratory scale tests through full-scale pilot tests up to the industrial design definition is also highlighted.

A monitoring and surveillance system is a mandatory element to ensure that a radioactive waste disposal facility provides and meets the required level of safety during both its operational and closure periods. Thanks to the technological advancement that has taken place in recent years, it has become possible to implement distributed wireless networks of lowpower and low-cost sensors to monitor parameters of interest in different scenarios that relate to sectors such as the civil, environmental, and industrial ones.

In this study an innovative approach for the identification and monitoring of the structural integrity of stored radioactive drums is presented. It consists of a network of radio frequency nodes equipped with solid-state gamma-ray and thermal neutrons detectors developed at the University of Pisa within the European project PREDIS. The nodes provide a unique identifier to the waste drums, and at the same time they allow cyclic measurements of the emitted radioactivity with a frequency and duration established by the user. Since nodes must be battery powered, a high level of hardware and software optimization has been performed to guarantee several years of battery lifetime minimizing human operators’ interaction. Data collected by the nodes are automatically transferred to radio gateways through LoRa technology for visualization, processing, and storage purposes.

The proposed solution demonstrates the possibility of collecting data automatically and passively from radioactive waste drums at a great distance without the use of mobile and/or mechanical scanning systems, enabling a much greater storage capability compared to common identification solutions based on radio frequency identification or near field communication tags.

The general context of the article is to evaluate strategies that can be used to mitigate aerosols dispersion during the fuel debris or corium retrieval of Fukushima damaged reactors. IRSN is involved in several projects led by ONET Technologies along with CEA to provide relevant information to analyze the risk of aerosol resuspension induced by fuel debris retrieval. The knowledge of the aerosol source term emitted during fuel debris retrieval operations is one of the key issues for the assessment of aerosol dispersion that can lead to the release of radionuclides into the environment. Such information are also necessary to define an efficient strategy to mitigate this risk. Various mitigation means could be implemented during the decommissioning of Fukushima Daiichi damaged reactors, depending on the operations such as cutting of fuel debris or metallic structures or investigation in Primary Containment Vessel (PCV) by robots. It is also important to consider accidental scenarios such as earthquake event, to define countermeasures limiting the consequence in terms of safety and radioprotection. We propose to study various mitigation means such as the spray scrubbing technology used to collect airborne particles and therefore limit their dispersion during the cutting operations. Resuspension of deposited particles may also occur during the decommissioning operations due to various type of stress, such as aeraulic, mechanical, vibrational and also during underwater operations. To address these particle resuspension issues, another mitigation means made by coating of resins is introduced.

In the process of nuclear energy application, a certain amount of radioactive solid waste will be generated, and some of them are not available for direct disposal for various reasons, so some temporary storage facilities need to be established for storing the waste. In order to save operation cost, sometimes there is centralized temporary storage of radioactive solid waste, and in this case, a large amount of radioactive solid waste is stored in the temporary waste storage and the source of the waste is relatively complex. Generally such facilities have a large number of waste drums (usually more than 200 drums) and strong radiation sources, significant sky backscattering, complex shielding schemes and other key elements to be considered in the shielding analysis. In this article, a nuclear power plant radioactive waste temporary storage facility is chosen as the research object, typical radiation source is selected and modeled using the Monte Carlo method to study and analyze the key elements of radiation shielding for a multi-source storage facility.

The results of the study show that the thickness of the outer wall of the facility, the height of the outer wall, the storage capacity, the radiation source items in the facility, the thickness of the top shielding, and the top shielding of the radiation source can all influence the shielding effect of the multi-source radioactive waste storage facility. Afterwards, this article combines the characteristics of multi-source storage facilities from the influencing factors, and conducts shielding optimization analysis for different treatment processes and plant structures. In the subsequent radiation protection design of similar shielding facilities, the conclusions of this article can be used for radiation protection analysis as well as optimization to ensure that the dose to the environment around the facility and the public meets the requirements.

To achieve decommissioning, it is important to dry wet adsorbent for contaminated water treatment in a container and to reduce the risk of leakage and spread of contamination. Accordingly, a vacuum drying apparatus using a one-third-scale cylindrical container was designed and manufactured, to understand the complex drying mechanism and determine the specifications of vacuum drying equipment. The apparatus consisted of the container, an outer cylinder with three heaters on its outer surface, a steam condenser and others. Drying experiments showed that drying proceeds from the outside of the container; initially the top of the container dries first, then the bottom dries more rapidly, and finally the bottom dries out faster than the top. Furthermore, the drying time becomes shorter with higher outer cylinder temperature and higher vacuum of the container. It was also found that the amount of condensation per unit time decreases gradually with time and the drying time could be related to when the rate of change over time of condensation amount decreases below a certain values, even when the initial saturation was changed. Using trends in this data, we suggested a method for determining the drying time. The method also uses ΔT, the difference between the heater temperature and the saturation temperature at the vacuum level, as a parameter to determine the drying time.

Applicability of temperature swing extraction technology employing monoamides was examined for uranium contaminated waste treatment procedure. Separation experiments on simulated target solution with three kinds of monoamides with different structure showed that Ce(IV) in the solution was selectively recovered by the temperature swing extraction operation. Based on the experiments, an appropriate monoamide for the procedure was selected.

Various types of machine oil are used for analysis and utility equipment, and these organic liquid wastes are stored in the nuclear facilities and laboratories due to lack of appropriate treatment processes. Treatment of organic liquid waste is one of principal tasks, since radiolysis of organic material generates various hazardous products. Perfluoro oil, generally used in vacuum pumps, is difficult to be decomposed because of chemical stability. Calcination of fluorine compounds is possible to generate toxic and corrosive gas products. In order to achieve complete mineralization of the organic liquid wastes, application of subcritical water reaction was examined. In this study, effect of introducing a functional group into a perfluoro compound on its decomposition performance was experimentally evaluated. First, we carried out the transformation of perfluorohexane to perfluorohexyl iodide or perfluoroheptanoic acid based on reported procedures. Next, laboratory scale batch-wise decomposition tests with subcritical water on perfluorohexyl iodide and on perfluoro heptanoic acid were carried out. Analyses on degraded organic products remaining in the H₂O phase by atmospheric pressure ionization mass spectrometry and liquid chromatograph mass spectrometry were carried out. The decomposition products of each fluorine compound were identified, confirming that subcritical treatment is a promising treatment method.

Reducing the uncertainty is very important in making better decisions as to whether the measuring target object is radioactive waste or below the clearance levels. This could lead to a reduction of waste classified conservatively as radioactive waste.

One of the factors that overestimates radioactivity level is that the model of weight density distribution used as a conversion factor from measured radiation counts to radioactivity is different from the actual object. In order to solve this problem, this method virtually divides the object into some regions and sets the density distribution for each region.

The density distribution is set by measuring the amount of X-rays transmission in this method. The X-rays transmission amount depends on the thickness of the object in the region. The radioactivity conversion factor for the region is set by evaluating the radiation shielding effect based on the density distribution. The count rate of γ-rays is measured by γ-ray detectors for each region. And then, the activity concentration of the object is evaluated for each region by the weight, γ-rays count rate, density distribution, and radioactivity conversion factor.

The greatest merit of setting the density distribution is that there is no need to store the object in a uniform state to set the conversion factor, and pretreatment and storage work for measuring is greatly reduced.

A device equipped with a unit that sets the density distribution by measuring X-rays is being developed. This device is a box-type measuring one that measures and evaluates activity concentration by storing the object in a container.

More fundamental data is needed for setting the density distribution and the radioactivity conversion factor through experiments to improve the evaluation accuracy.

As a prerequisite for radiation monitoring, it is necessary to assess whether the detection device and analytical method can reliably monitor the radionuclides in a nuclear system or not. A key physical quantity is the minimum detectable activity concentration (MDAC) of the radionuclides to be measured with a certain level of confidence for a given monitoring system. For the high resolution γ spectrometry monitoring system, it is essential to determine the MDAC qualitatively and quantitatively. There are several methods for calculating the MDAC. In these methods, Currie method and ISO standard 11929 method, have been widely used. This paper discusses the similarities and differences between Currie method and ISO standard 11929 method in the term of the ranges of application form the perspective of the development, which can provide a reference for the radioactivity measurement in nuclear power plants.

The decommissioning process for Ohi Power Station Units 1 and 2 will take approximately 30 years, and will be divided into four phases, and according to our plan we will proceed decommissioning process step by step. This paper describes the specific construction status of the first stage. In the first stage, as preparatory work for future dismantlement, systems such as the reactor vessel and steam generator will be decontaminated, and radioactivity remaining inside the facility will be investigated. In addition, the secondary system equipment will be dismantled. Regarding the construction from the second stage onwards, we plan to apply for approval for changes to the decommissioning plan based on the results of the radioactivity survey in the facility that will be carried out in the first stage.

Waste such as grass and trees with radioactive cesium that was scattered widely due to the accident at Fukushima Daiichi nuclear power plant was buried in a landfill disposal site after intermediate treatment. Therefore, in order to clarify the current status of the radioactive cesium accumulated in the landfill disposal site, the radioactive cesium concentration in the depth direction was investigated. A hole with a diameter of 10 cm and a depth of up to 100 cm was dug in the landfill disposal site, and a small amount of waste was collected from the hole and the radioactive cesium concentration was measured with a germanium semiconductor detector. From the present study, the tendency of the distribution of radioactive cesium concentration in the depth direction was quantitatively clarified. In addition, it was found that since the distribution of the radioactive cesium concentration in the depth direction from the landfill surface was complicated, the tendency of that cannot be predicted using the conventional exponential function model used for general soil.

By using the metal fiber technology, silver zeolite technology, scrubbing technology, etc. established by filtered containment venting system, FCVS an air purification system that can be used under a wider range of conditions is expected.

The metal fiber filters and silver zeolite (AgX, etc.) that have a proven track record in Filtered Containment Venting System(FCVS) are nonflammable and can be used even at high temperatures. And also scrubbing technology is greatly improve the decontamination factor (DF) of radioactive substances. And also scrubbing technology is greatly improve the decontamination factor (DF) of radioactive substances. The system is expected to be used in a wide variety of ways such as by moving to an emergency evacuation site or a highly polluted place in the event of an accident to supply clean air for We manufactured the prototype of portable air purification system in 2020. As a result, we confirmed the high adsorption performance of organic iodine and the high collection rate of barium sulfate.

In this paper, we will propose a safer air purification system that can prevent the exposure of nuclear workers by using silver zeolite, which has excellent adsorption performance.

Through the research of Filtered Containment Venting System (FCVS), air purification methods have been developed, such as technologies of scrubbing, metal fiber, silver zeolite and mobility. Using these technologies, we have been studying the further improvement of air purification systems in the nuclear field. In this study, we conducted adsorption experiments using several metal fiber filters and evaluated the effect of the metal fiber wire diameter on the decontamination factor (DF). It was found that metal fibers with fine wires made a significant contribution to the removal of materials with minute particles.

Since various gases such as carbon monoxide (CO) are generated by MCCI (Molten Core Concrete Interaction), this is suppressed by injecting water into the lower part of the containment vessel. In order to install silver zeolite in the containment vessel and confirm whether it can play an auxiliary role in the hydrogen recombiner, we conducted joint research with PSI on the effects of CO and other substances on hydrogen reactivity. AgX exhibits a remarkable hydrogen-catalyzed reaction as before in a water vapor-free environment. Furthermore, it was confirmed that the concentration of CO₂ in the out gas increased in the presence of CO, indicating that it also has a catalytic effect on CO. The amount of hydrogen treated was calculated from the results of the AgX hydrogen reaction test. Under the conditions of sample temperature 170 ° C, residence time 0.092 seconds, air 45%, hydrogen 5%, water vapor 9%, and atmospheric pressure, it was found that the hydrogen processing capacity per AgX unit weight was 0.207 kmol / kg or more of hydrogen.

A Based on the advanced FCVS technology, such as a gas-liquid mixing nozzle and a multi-layer metal fiber filter, and a silver zeolite filter for the purpose of removing organic iodine. we have developed a demonstration machine for a large-capacity air purification system that purifies a large amount of contaminated air. With this acceleration and centrifugal force field of water, almost all of the fine particles in the bubbles can be transferred to the liquid phase water and dissolved. We have developed a demonstration machine of an air purification system that can remove fine particles and viruses in the air of 100 m³/h with a single nozzle and we also developed a large-capacity air purification system that can purify 2400 m³ of air per hour with 24 nozzles. The FCVS system, which is high-performance and passively activated by using SRV, is thought to help eliminate obvious infringement of personality rights, avoid litigation outages, and promote public acceptance for nuclear power to restart. As a result of PRA, the loss of decay heat sequence called "TW" is reduced from 2.9x10^-6 to 3.8x10^-9. The total core damage frequency will be reduced almost 1/1000.

In this way, the merits of installing FCVS are great, increasing public safety, security and acceptance of nuclear power plants, increasing the corporate value of companies that operate nuclear power plants, and contributing to global environmental sustainability.

On the decommissioning of TEPCO’s Fukushima Dai-ichi nuclear power plant, fuel debris including water will be picked up from the bottom of the reactor vessels and packed into storage containers, and as a result hydrogen and oxygen will be generated by radiolysis of waters. Since hydrogen is flammable gas, it has a high risk of combustion and explosion. In order to reduce the hydrogen gas concentration and to secure long-term integrity of the storage container for fuel debris, Passive Autocatalytic Recombiner (PAR) was used. It consists of a spherical alumina as a base material and a small amount of platinum coated on the outer surface. Parametric experiments were conducted to clarify the controlling factors affecting the reduction of hydrogen concentration in the storage container with PAR, using a small-scale modeled experimental container. As controlling factors, the quantity of PAR, the installation position of PAR, the size of PAR, the amount of hydrogen generation, etc. were quantitatively evaluated. On the other hand, an analytical method based on the Arrhenius equation has been constructed to analyze the recombination chemical reaction between hydrogen and oxygen. Using this method, an analytical evaluation for the controlling factors was performed.

Single-crystal diamond (SCD), as an ultra-wide bandgap semiconductor material, has the advantages of high carrier mobility ,super radiation hardness characteristics, high thermal conductivity and high breakdown field intensity, making it an ideal material for the next generation of radiation detectors . The SCD film material is crucial to the radiation performance of the detectors, according to earlier investigations. In this paper, Metal-Semiconductor-Metal (MSM) SCD detectors were fabricated based on four different qualities SCD film materials manufactured by microwave plasma chemical vapor deposition (MPCVD) method, and their material characterization, electrical properties and radiation characteristics were examined. Atomic force microscopy (AFM), X-ray diffraction (XRD) and laser Raman scattering spectroscopy were used to analyze the surface morphology and lattice structure of several SCD films. The results of the characterization show that the two materials outperform the other two materials in terms of surface roughness and lattice defect density. The results indicated that the four detectors exhibited superior electrical properties, with resistivities over 1E13 Ω·cm, leakage currents below 2.5 pA at 100 V, and linear ohmic contact range of about ±0.3 V/μm. The charge collection efficiency and energy resolution of SCD detectors were measured by ²³⁸Pu α radioactive source. The charge collection efficiency of SCD detectors prepared by the two materials with better material characterization results was close to 100%, and the energy resolution was 2.5% and 2.1%, respectively Due to the considerable large surface roughness and many internal defects of the materials the other two detectors cannot be utilized for energy spectrum measurement, because of too small pulse amplitude and poor signal-to-noise ratio. Results can provide reference for selecting diamond film materials for radiation detectors and can be used to optimize the preparation process of diamond detectors.

Spent nuclear fuel (spent fuel), which is used nuclear fuel that has been exposed to radiation usually produced by nuclear reactors in nuclear power plants, is an inevitable product from the development of nuclear energy. Almost all of the fuel content is radioactive, and long systematic process are required for the safety management, which has always been an important global issue. In order to make sure that spent nuclear fuel should be safely managed, different countries developing nuclear power have established a complete policy and legislative system, so as to ensure that the whole process of spent fuel management is systematic, standardized and effective. In developed countries such as France, Russia and Japan, closed-cycle strategy is implemented with industrial-scale reprocessing plant under construction or in operation. At present, China has become the country with the largest scale of nuclear power under construction in the world. There will be a large number of spent nuclear fuel requiring properly and safely managed. The lessons-learning of how developed countries managing spent nuclear fuel arising is important for China. The authors suggest that it is necessary to combine the top-level design to the legal practice, so that there are laws to respect during all steps of spent fuel management, and responsibilities of all parties are clear.

One of the lessons learned from the accident at the Fukushima Daiichi Nuclear Power Plant is that an effective accident management system should be equipped on a nuclear reactor to prevent an accident from leading to severe consequences. In which filtered containment venting system (FCVS) is considered an effective system that can enhance the capability to suppress or prevent severe accidents by reducing the pressure, steam water, and flammable gas in containment vessels. The water inventory in a wet-type FCVS plays an important role that affecting the retention efficiency of FCVS. However, a high-level water of the FCVS may affect the performance of the liquid droplet separator due to the high entrainment rate. To overcome this, in this study, perforated metal sheet plates were installed on the upper part of the FCVS to decrease the entrainment rate. The void fraction measurements and entrainment rate estimation were carried out on a large-scale FCVS with and without the perforated plates. It showed that the perforated plates have significant contributions to decreasing the entrainment rate of FCVS, therefore, it allows to expand the water inventory (level) of the scrubber tank.

Information of Cs distribution is important for decommissioning of the Fukushima Daiichi Nuclear Power Station (1F). Several experimental studies confirmed the Cs retention on stainless steels by chemical reaction at high temperatures (commonly > 800 oC), but the Cs retention on non-metallic materials, such as concrete and thermal insulators, was not fully understood though they are used with large quantity in light water reactors. This study demonstrated that Cs might be deposited and retained on the concrete structure where the temperature was not so high during the 1F accident. It was revealed that the CsOH/concrete interaction at around 200 oC resulted in the formation of water-insoluble Cs-(Al,Fe)-Si-O deposits and water-soluble phases, i.e., cesium carbonate hydrate and possibly cesium silicate, if Al and Fe are not present. CsOH might be trapped on concrete by chemical reaction with CaCO₃ to form Cs₂CO₃ hydrate, and with aluminosilicate and SiO₂(quartz) to form Cs-Al-Si-O and Cs-Si-O deposits, respectively. This output will be useful for elucidating the trapping mechanism that caused an extremely high radioactivity on concrete shield plugs at 1F, and for developing an effective decommissioning practice for concrete structure.

Solidification treatment is planned for filter sludge and spent resin waste generated from nuclear power plants. Ordinary cement is used to solidify waste. However, when cement is used for solidification, it has been confirmed that radioactive substances are easily eluted due to the influence of cellulose and resin contained in filter sludge and the like.

If the solidification treatment is performed using chemical treatment or heat treatment, secondary waste is generated or a dedicated treatment facility is required, which complicates the solidification treatment.

We conducted a test using a geopolymer (SIAL^®) that has been used to treat radioactive waste from nuclear power plants in other countries.

From 2018 to 2019, We have conducted laboratory-scale experiments using SIAL^® to solidify filter sludge and spent resin. The unconfined compressive strength was 6.4 MPa or more, satisfying the unconfined compressive strength of 1.47 MPa, which is the burial standard. In addition, most of the distribution coefficients of major nuclides satisfied the set values for burial. Based on the above, we have confirmed that there is no problem with the solidification and the ability to contain radioactive materials. we are planning to consider the formulation adjustment of the solidification material and the addition of additives.

In 2020, we conducted a cold test simulating filter sludge and used resin in order to investigate whether even drum-sized waste can be solidified. The unconfined compressive strength is 4.2 MPa or more, and we have confirmed that there is no problem in solidifying even drum-sized waste.

Based on the results of solidification of filter sludge and spent resin, we thought that it could be applied to other wastes, and in 2021, we conducted a solidification test on incinerated ash. As a result of conducting a full-scale test using general incineration ash, it was confirmed that drum-sized waste could be solidified. However, when the incineration ash contains a large amount of amphoteric metals such as aluminum, the reaction between the amphoteric metals and alkalis generates air bubbles, which lowers the strength of the waste. We are planning to consider pretreatment before incineration.

In the future, we plan to investigate the applicability of SIAL^® technology to hot tests of incineration ash and sulfate (concentrated waste liquid). We will continue to collect data to confirm that the solidification method is suitable for burial.

Cobalt-based alloys are used as a hard-facing for valve seats installed in nuclear power plants. In high-temperature, high-pressure service environments, cobalt leaches out due to erosion, corrosion, wear and cracking. Since leached cobalt becomes 60Co due to radiation in the reactor, we have developed a method to improve the corrosion and wear resistance of cobalt-based alloy valve seats using directed energy deposition(DED).

In this development, powdered cobalt-based alloys are melted by a laser and bonded to the surface of the base material.

The objectives of the research were to evaluate the material properties, and to conduct an evaluation through an actual performance test and an endurance test using a 600A motor operated gate valve.

The material properties were evaluated as follows: the erosion depth was less than one-tenth that of conventional valve seats in high temperature water corrosion tests, the plane strain fracture toughness KIC was about three times that of conventional valve seats.

In the actual performance test and durability test using a 600A motor operated gate valve, there was no significant damage to the seat surface even after 100 operations at 288°C in a saturated steam differential pressure environment, and amount of the leakage was less than the allowable leakage, and the performance was confirmed to be equivalent to that of a conventional valve seat.

The core of an advanced gas-cooled reactor (AGR) is constructed with a large number of nuclear graphite components. These components form channels for fuel and control rods and also restrain the core by incorporating a keying system. The structural integrity of these components is of prime importance for safe operation of the AGR. Fuel channels are built-up of cylindrical bricks which are exposed to fast neutron irradiation and radiolytic oxidation during the life of a reactor and as a result the fuel bricks experience highly non-linear changes in deformation and material properties. These highly non-linear deformation and material property changes generate a complex stress state within the fuel brick and its interacting graphite components. The behaviour and assessment of graphite components is, generally, conducted using time integrated finite element (FE) modelling and analysis. The deformation and material properties which are used in these types of assessments are calculated as a function of fast neutron irradiation, irradiation temperature and weight loss, to reflect radiation ageing of nuclear graphite over the reactor’s life. This paper provides insight to complex finite element analyses conducted using COMSOL Multiphysics to predict keyway root cracking and seal ring groove wall cracking of graphite fuel bricks.

The regulation for nuclear power plants (NPPs) has been enhanced after the Great East Japan earthquake and the accident at TEPCO’s Fukushima Daiichi Nuclear Power Stations in March 2011 to take countermeasures against beyond-design-basis events. The importance of seismic probabilistic risk assessment (PRA) has drawn much attention to improve the seismic safety of nuclear facilities against earthquakes that exceed the design input ground motion. The realistic seismic response of the equipment and piping in NPPs for fragility assessment in seismic PRA needs to be evaluated. In particular, as piping systems have plant-specific complex configurations, the arrangement and stiffness of piping support structures have a significant impact on the seismic response characteristics of the entire piping system. By contrast, the current seismic design procedure adopts an evaluation method assuming an elastic response. Hence, this study aims to develop an elastic-plastic response analysis method that can estimate the realistic response of piping systems, including piping support structures.

The authors have conducted loading tests on piping support structures and optimization studies of elasto-plastic property models to develop an elasto-plastic response analysis method for piping systems, including piping support structures. In this study, the applicability of the method is confirmed through a simulation analysis of the elasto-plastic response for the piping support structure loading test previously reported. As the results, a good correlation is found for the ductility factor and the damage status between the test and simulation analysis results. Therefore, the ductility factor is concluded to be effective as a damage evaluation index for piping support structures.

Hydrogen is one of the promising major energy sources in the future. A thermochemical water-splitting iodine-sulfur process (IS process) is one of candidates for the large-scale production of hydrogen using heat from nuclear energy or solar power. Severe corrosive environment which is thermal decomposition of sulfuric acid exists in the IS process. To achieve an industrialization of massive hydrogen production system, one of the key factors is the development of structural materials for the severe corrosive environment. A hybrid material with corrosion-resistance and ductility was made by a silicon powder plasma spraying and laser treatment. To confirm the production characteristics of a container using the hybrid material, the container which has a welded part, a chamfer, a curved surface was experimentally made. The substrate of the container made of Hastelloy C276 superalloy includes TIG welded part. A corrosion test of the container was performed in 95 mass% boiling sulfuric acid until 500 hours to evaluate the corrosion characteristic of the container. The container had excellent corrosion resistance in the condition of 95 mass% boiling sulfuric acid.

After the 2011 Fukushima Daiichi Nuclear Power Station accident, the importance of seismic probabilistic risk assessment has drawn much attention to improving the safety of nuclear facilities against ground motions exceeding the design ground motion. Since the seismic force acts on multiple components simultaneously, evaluating the joint damage probability of multiple components considering the response and resistance correlation for detailed analysis of scenarios under severe ground motion is crucial. However, the building’s conventional simple sway-rocking (SR) lumped mass model cannot sufficiently express the three-dimensional (3D) detailed seismic response characteristic, including the local response and damage. There was no choice but to express the effects of correlations between damage events in a simplified way, such as perfect correlation. Therefore, by utilization the seismic response analysis method using the building’s 3D finite element (FE) analysis model that can express local response and damage and obtaining the building’s detailed floor response related to equipment damage, a more realistic evaluation of the correlation of equipment damage is expected. This paper focuses on the equipment response related to equipment damage in nuclear facilities, compares the correlation coefficients of the maximum acceleration and floor response spectra of the equipment installation position obtained using the building’s conventional SR and 3D FE models, and clarifies the correlation characteristics of equipment response.

Most studies on local damage of reinforced concrete (RC) panel structures subjected to projectile impact have focused on normal impact, and only few studies have focused on oblique impact. Thus, we conducted impact tests under different impact conditions including oblique impacts so as to confirm the different impact behaviors of the RC panel structure. This study aimed to develop an analytical method by examining the test results and analytical conditions, and validate the proposed method through comparison of the test results.

We focus on the scabbing damage which is one of the local damage modes of RC panel and intend to investigate the relationship between the damage criterion and scabbing occurrence based on oblique impact test results due to soft projectile with hemispherical nose shape.

With the advantages of complete exemption of contact, wear contamination and lubrication, excellent endurance, and well-controlled performance, active magnetic bearings (AMBs) have been the ideal solution to support the rotating machines in the High Temperature Gas-cooled Reactors (HTGR). However, the nonlinear factors introduced in the AMBs-rotor system can induce the rotor with complicated behaviors, which have negative effects on the safety and reliability of the reactors. This paper utilizes numerical methods to study one common nonlinear behavior, namely nonsynchronous vibrations. Time-domain responses, Poincaré section, and power spectrum density are obtained based on the numerical integration. Furthermore, the global dynamics of the AMBs-rotor system is elaborated based on the generalized cell mapping digraph method. By analyzing the numerical results, it is revealed that period-3 motion arises in the AMBs-rotor system with a large increase in the maximum rotor displacement. Nonsynchronous vibrations are generated on account of the multiple co-existing solutions, trivial and nontrivial solutions, caused by the nonlinearity of the electromagnetic force. On specific operating conditions, nontrivial solutions are exhibited and the nonsynchronous vibrations occur accordingly. Nonsynchronous vibrations can be prevented by eliminating nontrivial solutions through the control parameter modification.

ASME B16.5 and B16.47 flanges are extensively used in nuclear pressure vessels and pipping systems. With the ubiquitous use of these flanges in harsh environment, their integrity and leak tightness need to be analyzed globally. In particular, identifying the flange classes and sizes that are more effected by the level of the tightening is of interest to the piping engineer in order to take the necessary action to not overstress these flanges. In the absence of an appropriate design method that takes into account the flexibility of the complex statically indeterminate bolted flange connection and the interaction between the different joint elements, some class and size flanges are compromised and can suffer failure.

This paper proposes the use of an accurate analytical model that is validated using the numerical FE method to analyze all ASME standard flanges. The study focuses on parameters such as flange rotation and stresses in the flange, gaskets and bolts. The most critical size and class flanges and their highly stressed locations will be revealed. Some guidance on the level of bolt-up stress that should not be reached for specific classes and sizes will be given in order to avoid overstressing the flange or leakage failure.

After Fukushima nuclear accident, there are higher demands and requests for the development of severe accident analysis software since the nuclear regulatory agencies set more stringent and specific rules of the nuclear power plant (NPP) in severe accident emergency and training. This research introduces an initialization and coupling method, which captures and continues an accident transient by MAAP5.03 code. Integrated with the existing full-scope simulator (FSS), the implementation of severe accident simulation can be extended through the method. Thus the FSS can perform more abundant and complete accident scenes in the personnel training and accident drills. Moreover this method provides us a predictive way to the NPP operating state or accident processes by limited parameters, together with the accident detection and diagnostic analysis procedures. To ensure MAAP’s calculation effectively recognizing and following the transient tendency, the architecture of MAAP was analyzed and a set of simulation parameters were selected. The comparative results of the steady state and a LBLOCA accident state for 3rd generation passive NPP were performed and some specific quantitative assessments was defined to verify the functionalities derived from the initialization and coupling method. We also examined the prediction and evaluation of accident management for the plant under the MSLB conditions. The results show MAAP can adequately capture and smoothly continue the accident transient state. And also it demonstrates the feasibility to externally revise the severe accident analysis programs for extensively simulating and analyzing demands for the NPPs.

Numerical simulations of dispersed solid particle behaviour inside a scrubbing pool are presented. The goal is to evaluate the decontamination factor of the particles during the pool scrubbing process. The basic phenomena of pool scrubbing are described. The setup used for the simulation validation is presented. Then, the boundary and initial conditions of the PECA experiments, which were used for simulations, are presented. The subgrid model for particle decontamination is presented and the calculation results are evaluated and compared with the PECA experimental results.

Based on the lessons learned from the accident at the Fukushima Daiichi Nuclear Power Plant, and the revision of the national government's basic plan for disaster prevention, and trends among other electric power utilities, Electric Power Development Co., Ltd. is going to implement accident management education (AM education). As part of the education, AM simulator has been developed to visualize plant conditions at the time of an accident in order to promote understanding of severe accident response.

AM simulator visualizes plant behavior during a severe accident, allowing the user to confirm the details of plant conditions and the uncertainties of physical phenomena and operations. This will promote understanding of text-based AM education.

In addition, we are considering using the AM simulator to analyze plant behavior during SA in advance and to expand the database which is necessary for predicting the progression of events during AM.