Over 10,000 m3 of slurry wastes were generated from Multinuclide Removal Facility (ALPS) since the accident. Simulated carbonate slurry was composed of CaCO3, Mg(OH)2 and NaCl and iron coprecipitation slurry was mainly composed of FeOOH and NaCl with SrCO3 and CsCl. To convert phosphates, simulated slurry was dissolved. FeOOH was dissolved in 2 M-HCl and ascorbic acid, and the other compounds were dissolved by adding HCl up to 3.5 M. Na2HPO4 was added to the solution. 4.8 M of NaOH solution was added to neutralize, and precipitation includes NaCl was obtained. When ratio of cation to phosphate (C/P) is less than 2, Sr and Ca in the filtrate were less than 0.01 %. It was rinsed by water to remove NaCl. Total amount of Sr and Ca were less than 0.1 %, Mg and Fe were less than 1.0 %. The rinsed precipitation was heated from 400 to 600 ºC. Heated precipitation at 600 ºC was crystalized and others were still amorphous. The heated precipitation was hot-pressed at 127 MPa with 300 ºC. The pellet was evaluated leachability based on ANSI/ANS 16.1. It was suggested leachability of the pellet on the condition ≤ 500 ºC and C/P = 1.4 was especially quite low. Although Cs wasn’t contained in phosphates, but over 99 % of Cs can be removed from dissolve solution by chabazite and CST. The solidification process from ALPS slurry was made by acid dissolution and Cs adsorption step, conversion and precipitation, rinsing, heating and hot press. This solidification waste was confirmed quite low leachability.

Tritium is very difficult to analyze mainly because its emission type (beta minus), its chemical and its physical behavior. However, during dismantling activities as well as environmental control, it is necessary to characterize it with low uncertainty.

Tritium is a radionuclide produced either naturally when cosmic ray interacts with nitrogen and oxygen atoms in the upper troposphere and lower stratosphere or during a man-made nuclear process when fast moving neutron with an energy of more than 4MeV reacts with nitrogen, measurable amount of tritium was released in the environment after the Fukushima accident.

The beta decay makes tritium difficult to characterize as when it undergoes beta minus decay to form the stable helium-3 atom. The electron produced by the reaction is only 5.7keV on average, one of the lowest energy beta particles. The low energy causes the beta particle to have low penetration, they can only travel an average of 6mm in air. Thus, characterizing tritium from a distance is extremely difficult.

Tritium is also hard to detect since it behaves like hydrogen and can form tritium gas or replace hydrogen molecule in other organic molecules. These molecules can easily be absorbed by the human body and become a hazard. While tritium is not dangerous in the vicinity of people, it will be if ingested. Furthermore, volatility makes a nuclide that need to be monitored carefully. Tritium permeates through many materials and porous substance including metals and rubber.

Tritium has a half-life of 12.312 years and is considered a short-lived radionuclide. However, its permeability, high coefficient of diffusion and chemical property similar to hydrogen making it difficult to detect makes it a concern for decommissioning projects.

Various ways of characterizing and measuring tritium exist, that will vary with the expected amount of tritium, the nuclides present and the sample, be it gas, liquid or solid. However, no technologies were crowned as objectively better for tritium characterization yet.

This paper presents different ways to monitor, characterize and measure tritium such as ionization chambers, proportional counters, scintillators, volumetric analysis, chromatography, and spectroscopy among others. We will as well, introduce the development in technologies that were made for the development of the M IONIX – TGN real time monitoring of tritium. Discussions on usual problem associated with tritium monitoring such as the elevation of radioactive wastes, response time and maintenance will also be presented.

One of the important challenges for the decommissioning of the damaged reactors of the Fukushima Daiichi Nuclear Power Station (1F) is the fuel debris retrieval. The generation and dispersion of radioactive aerosols during the cutting operations of fuel debris pose a significant radiation and health risk during the decommissioning process. Quantitative assessments of potential contamination with radioactive aerosols are fundamental to the development of a safety case for the decommissioning of 1F. Detailed studies on the generation of aerosols during cutting processes of inactive fuel debris simulants by various tools have been performed [1, 2, 3, 4, 5], underlining the influence of each cutting tool and its operating conditions, the chemical composition of cut materials, and the environmental conditions of the dismantling, such as in air for dry or wet conditions and underwater conditions. The size distribution, morphology and composition of particles released during cutting are equally significant when predicting their transport and deposition behavior through numerical simulations and assessing their radioactivity. Furthermore, quantitative characterization of radioactive aerosols is essential to develop and optimize technologies to capture and mitigate particulate contaminants dissemination.

In this context, the URASOL project (abbreviation for "Particle generation test using simulated uranium containing debris", a word coined from URAnium and aeroSOL), which is undertaken by a French consortium consisting of ONET Technologies, CEA, and IRSN for JAEA/CLADS, is dedicated to acquire basic scientific data on the generation and characteristics of radioactive aerosols from the thermal or mechanical processing of fuel debris simulant [6]. Heating process undertaken in the VITAE (acronym for VITi-AErosols) facility simulates some representative conditions of thermal cutting by laser [7]. For mechanical cutting, the core boring technique is implemented in the FUJISAN facility. Fuel debris simulants have been developed for inactive and active trials [8].

The aerosols are characterized in terms of mass concentration, real time number concentration, mass size distribution, morphology, and chemical properties. The chemical characterization aims at identifying potential radioactive particles released and the associated size distribution, both of which are important information for assessing possible safety and radioprotection measures during the fuel debris retrieval operations at 1F.

This extended abstract provides an overview of the research that was completed within the UK’s Robotics and AI for Nuclear research hub (RAIN). RAIN was established in the UK in 2017 with the aim of developing robotic systems to address challenges in the nuclear industry and to demonstrate the capabilities of these solutions in active nuclear facilities. The research within RAIN was focused on inspection of nuclear facilities, handling operations (with particular focus on the development of a remotely operated glovebox), and engagement with the nuclear regulator to ensure that the robotic systems being developed could gain approval for deployment onto radioactive sites.

Decommissioning work of Tokyo Electric Power Company (TEPCO) Fukushima Daiichi Nuclear Power Station (1F) is in progress. Internal investigations, sim-tests, and accident analyses are underway to understand the state of reactor pressure vessels (RPV) and primary containment vessels after the accident by TEPCO. During the severe accident (SA), various materials, including stainless steel control rods, B₄C, Zr-alloys channel box materials, and fuel rods, melt in the core region and fall into the bottom of the RPV. According to a possible scenario for 1F unit 2 based on the internal investigation, there is a possibility that most of the fuel oxides could not melt completely, and a metal melt pool could be formed. The metal pool could react with structural materials such as the Control rod drive (CRD), consisting mainly of a stainless steel housing and an Inconel stub tube, and cause the failure of the RPV. The reaction between the metal pool and the structural materials is important in understanding the RPV failure behavior. In this study, the ELSA (Experiment on Late In-vessel Severe Accident Phenomena)-1 test, which focuses on the damage caused by the eutectic melting of the liquid metal pool and CRD structures, was conducted.

A test sample simulating the CRD structure at the lower head was fabricated and loaded with Fe-Zr alloy as the simulated metal corium. The sample was gradually heated up to about 1400°C using the LIESAN (Large-scale Equipment for Investigation of Severe Accidents in Nuclear reactors) test facility, and in-situ observation was performed using a video camera.

The test results showed that the CRD structural material reacted with the metal debris and melted and failed at about 1050-1250°C, which was lower than the melting point of the CRD itself. It was also observed that the molten material flowed into the CRD, suggesting that the CRD structure was preferentially damaged during SA. From the cross-section observation of the sample and phase diagrams on its materials, it was found that the housing and the stub tube of the CRD melted and failed due to the Zr-Fe and Zr-Ni eutectic reactions with Zr in the molten metals, respectively. In the experiment, it took a certain amount of time from the start of the CRD reaction with the metal pool until it melted and broke, indicating that the reaction speed was also an important factor in the failure of RPV.

Robotic manipulations in contaminated environments are often either too complex or unstructured to be performed fully autonomously for decommissioning or clean-out processes, thus; remote robotic systems, where the human operator is in the loop for direct control and making decisions, are generally employed. Bilateral teleoperation systems enabling remote manipulation of the delicate materials in hazardous conditions with high dexterity and sufficient force feedback enhance operational safety within the nuclear industry.

Typically, operators need to carry out an extensive training programme in order to safely operate such a complex robotic system in hazardous conditions, e.g., approximately 6 months for Joint European Torus (JET) tokamak maintenance which is a highly structured and well-defined process. Training the operators, however, for the challenging conditions imposed by a contaminated environment, or after a disaster (e.g., low visibility, restricted motions in the confined spaces, limited interaction force) in a time- and cost-effective manner is difficult. For training purposes, creating physical mock-ups is not always possible due to the overall uncertainty within the actual operating conditions and using the bespoke robotic systems is not always possible due to the tight maintenance and operational schedules and limitations to the number of available systems.

This paper highlights the importance of Haptic Digital Twin (HDT), a complete system with physics simulation, haptic rendering, and local haptic device(s), can prepare the operators to remotely manipulate hazardous materials more economically by simulating different tasks, robots, and their surroundings in the hazardous environment. The operational readiness level of the operator can be increased by simulating different types of representative failure scenarios (e.g., actuator or sensor malfunctions, task setbacks, etc.) and different remote manipulation tasks can be developed, verified, and validated within the virtual platform. The proposed HDT can be applied to various decommissioning processes including the post-operational clean-out process (POCO) and cleaning out the Fukushima-Daichi power plant, in particular during the fuel debris retrieval process.

An articulated manipulator has been proposed and developed for retrieving fuel debris in Unit 3 of the Fukushima Dai-Ichi. This fuel debris retrieval (FDR) manipulator comprises a series of revolute joints and prismatic joints, i.e., realising an over 20 m maximum reach mechanism. Thus, it allows for collecting fuel debris from the post-incident vessel into a transportation carriage. It is a trivial approach to develop bespoke long reach mechanisms for specific missions in the nuclear industry. However, there is a lack of available tools for planning operations allowing for efficient and safe use of robotic systems comprising long reach mechanisms. This work analyses the technical challenges in operating long reach mechanisms, e.g., like the FDR manipulator for nuclear decommissioning, multi-purpose deployment manipulators for maintaining nuclear fusion reactors, and so on. The visions are provided describing the on-going research and development (R&D) pathway, which aims to establish a common pave of planning and de-risking operations using long-reach mechanisms in the nuclear industry. The associated progress of the R&D activities in the scope of LongOps project is reported.

To complete the decommissioning of the Fukushima Daiichi Nuclear Power Station (referred to as 1F), it is necessary to maintain damaged components and equipment in a safe state for medium and long term. The subcommittee of structural integrity of components was established in 2019 as a subcommittee of Review Committee on Decommissioning of the 1F of the Atomic Energy Society of Japan. The purpose of this subcommittee is to discuss the structural integrity evaluation standards for components and equipment that have partially lost their strength due to a severe accident, and to propose a concept on appropriate standards.

Target components and equipment are the reactor primary containment vessel (referred to as PCV) and its internal structure, e.g.: the pedestal of unit 1 to 3. Therefore, it is necessary to consider not only the required functions specific to the damaged components and equipment, which should be different from an operating plant, and also the magnitude of uncertainty in evaluating structural integrity. The subcommittee has been discussing concepts of structural integrity evaluation considering large uncertainties. This paper describes the concept on structural integrity evaluation for 1F.

Numerical simulation method for the dynamics of fixed-bed column systems is necessary for the safe and economic operation of the contaminated water treatment facilities in Fukushima Daiichi Nuclear Power Station. A two-phase homogeneous diffusion model (TPHDM) is one of the common models for the fixed-bed system simulation. Although breakthrough curve of the column with IE-96 zeolite was successfully calculated with TPHDM, reasonable breakthrough curve was not calculated when the column system equipped with a Crystalline Silicotitanates (CST) was simulated. This implied the necessity to improve TPHDM by modeling the phenomena not yet considered. It is known that electric double layer (EDL) is formed around the porous particle such as zeolite when it is in contact with liquid phase. Hence, a new kinetic model was developed where the effect of adsorption in EDL is explicitly modelled by introducing a new region for EDL between bulk and macropore regions. EDL works as a capacitor and its capacity was explicitly treated by defining a parameter f which is a ratio of the capacity of the capacitor to the total adsorption capacity of the adsorbent. The value of f was determined by the best fit approach. Consequently, the calculated breakthrough curves of the CSTcolumn and IE-96 column was in good accordance with experimental result.

We have independently developed the laser decontamination device consisting of a diffraction-limited single-mode CW fiber laser and a high-speed galvanometer. The surface of RI (radioactive isotope) contaminants in the nuclear reactor stainless steel, concrete, glass, and inorganic and organic materials was sublimated and instantly evaporated with the high-power density and high-speed scanning laser light, and to decontaminate below the detection limit. Recently, we have successfully been able to observe the sublimation and evaporation of this laser decontamination device with highspeed movie camera. By observing low-speed movie images below the detection limit and high-speed movie images, it can be seen that the high-power density and high-speed scanning is needed to realize the sublimation and evaporation without residual contamination. It can be seen that the high-power density and low speed scanning is conversely resulted the expanded melting in which the contamination remains. In the conference, the device design that realizes the decontamination below the detection limit will be reported. And also, the decontamination process using the sublimation and evaporation by observing high-speed movie will be discussed.

We have developed an automatic radioactive contamination inspecting system for vehicle entering and exiting the radiation controlled area. At interim storage facilities that store radioactive wastes released by the accident at the Fukushima Daiichi Nuclear Power Station, contamination inspection of articles taken out of the controlled area is mandatory, based on the Ionizing Radiation Hazard Prevention Regulations. Contamination inspection is also required for large trucks, but it is usually carried out manually using a portable survey meter. In this case, the inspection time will be very long, so it is expected that traffic congestion will occur at the screening site. In order to solve these problems, we have developed a “radioactive contamination inspecting system for vehicle” that enables quick and reliable contamination inspection by joint development of Mitsubishi Electric Plant Engineering and Mitsubishi Electric Corporation. In order to detect slight contamination on the surface of vehicles in areas with relatively high background dose rates such as interim storage facilities, we have newly developed a large-area radiation detector.

The detector has high sensitivity to β-rays from radioactive cesium, but on the other hand, it has very little sensitivity to ambient γ-rays because of very little thickness of the detector. Furthermore, in order to efficiently inspect the entire vehicle, we have designed a movable inspection gate that installed five independent-moving radiation detector units. The detector units have several range finders to measure the distance to the surface of vehicles so that the inspection can be performed while keeping the distance from the surface of the vehicle constant. As a result, contamination on the surface of the vehicle can be detected quickly and reliably without radiation shield even in an area where the ambient dose rate is several μSv/h. For example, although conventional manual measurement method took about 10 minutes by 10 operators to inspect a 10-ton truck, the developed system can automatically inspect it within 4 minutes. The five systems were operated at the screening site of the interim storage facility from January 2019 to March 2022.

As part of the LONG Term OPerationS (LongOps) research and development program, Remote Applications in Challenging Environment (RACE), Sellafield Ltd and Tokyo Electric Power Company Holdings (TEPCO) have joined forces to develop the Next Generation Digital Mock-up (NG-DMU) for nuclear decommissioning and remote handling applications. TEPCO plans to conduct Fuel Debris Retrieval (FDR) in Fukushima Daiichi Nuclear Power Station Unit 3 prior to other units based on the “Mid-and-Long-Term Decommissioning Action Plan”. NG-DMU would be a valuable tool not only for efficient planning of the operations but also for use during the remote operations in FDR. In the early phases of the development, the elicitation of NG-DMU requirements occurs before the specific concept development of robotic equipment, which poses a significant challenge. Task analysis is used as an effective method to facilitate requirement capture for NG-DMU and support some aspects of the FDR implementation. Furthermore, the challenging environment of the Fukushima site complicates the Use Case Analysis even more. The uncertainty and complexity of the environment of the pedestal for supporting a Reactor Pressure Vessel inside the Primary Containment Vessel (PCV) and low maturity of the FDR strategy make the identification of the tasks and their sequence non-trivial. Site surveys provide valuable information about the environment. However, a complete representation of the environment is not available during task analysis, and certain assumptions have still to be made. For example, it is known that solid debris lumps and deposits remain at the bottom of the pedestal, and broken structures are piled up over it but the exact physical relationship between the structure and the debris is not yet established. Similarly, several assumptions have been made about the remote handling equipment which consists of a long reach arm, a multiple degree of freedom manipulator and an end-effector concept, referring to the promising devices for small scale FDR for Unit 2. This paper describes in more detail the task analysis process followed for the FDR to identify and prioritize tasks that can be supported by the NG-DMU and support its definition. The process involves several steps, from setting operational goals, breaking down tasks into manageable sets, with initial and final boundary conditions, etc. It is demonstrated that task analysis is a valuable tool for the development of a DMU that can be reconfigurable and easily adaptable to cover a range of operational needs or changes in the assumptions earlier in the development phase.

In Japan, at the middle of the year 2022, 24 Nuclear Power plants have been planned to stop and to prepare towards decommissioning. At these plants and at Fukushima Dai-ichi plant, safe and suitable management of radioactive materials is one of essential issues before decommissioning. For such a purpose, accurate radiation measurement is required however, target materials structures, or equipment are generally large, and heavy. Furthermore, they are not easy to cut into pieces or separate into parts. Classifying them efficiently, to not overestimate the amount of low level waste and put as much materials through clearance while operators are under as little dose as possible is a challenge. Considering these situations, radiation measurements should be on-site, able to characterize wide targets, keeping own shape and structures of the target, time efficient and accurate as possible.

To identify radioactive nuclides, it is necessary to determine the photons (γ-ray) energies from the target material accurately. Germanium detectors have been being utilized widely relied on their very high ability of energy resolution. However, to determine accurate radioactivity, not only photon counts but also geometrical efficiency between the detector and the target is required. Normally, the latter is not easy to evaluate accurately.

Conventionally, two methods are applied to determine geometrical efficiencies. One is the standard source method, the other is the numerical calculation method.

The standard source calibration method has been applied to this issue for a long time based on high reliability of the method. On the other hand, it requires costly and time consuming complex procedures. Moreover, it is often impossible to obtain an equivalent source that is representative of the target (in terms of geometry and density).

The alternative method is numerical calculation. General-purpose Monte Carlo codes, MCNP for example, have been widely used for this purpose however, calculation preparations require expert knowledge. In addition, due to the nature of Monte Carlo simulation, considerable computing time is required, increasing easily with the complexity of the model.

To solve In situ measurement problems, the In Situ Object Counting System (ISOCSTM) technique was developed. This technique has been already widely utilized around the world, especially it was well recognized by the NRC (United States Nuclear Regulatory Commission). ISOCS easily enables users to realize an accurate efficiency calculation in a quite short time.

This paper presents the abilities and capabilities of ISOCS. Furthermore, verification examples are shown through measurements with Eu-152 standard sources. Calculation values are also compared with those of MCNP.

The waste form development is a key issue for the decommissioning of the Fukushima Daiichi nuclear power plant. There are various radionuclides and wastes, some of which are troublesome nuclide in the waste management such as iodine-129 due to the extremely long-half life and mobility through the repository to biosphere. To immobilize the key nuclides, we propose the composite waste form that consists of primary wastes bearing objected radionuclides and a secondary metallic matrix which is the protection barrier for dissolution with high corrosion resistance. To demonstrate this concept, the representative composite waste form consists of silver iodide (AgI) and stainless steel was synthesized by powder metallurgy hot isostatic pressing (HIP). The powdery stainless steels were well sintered turning into the ingot matrix. No specific changes were observed for the AgI particles that they were individually dispersed and confined into the metal matrix. Based on the approved corrosion rate of stainless steel under an assumed repository condition, this composite waste form will immobilize the radionuclides for at least hundreds of thousands of years.

We developed a sodium iodide scintillator-type γ-ray spectrometer with improved energy resolution using an unfolding method for gamma ray (γ-ray) spectroscopy and dosimetry during emergencies such as radiation leakage. To improve the energy resolution, the response function was refined by weighting the energy dependence of the scintillation efficiency of the sodium iodine on the energy of the secondary electrons produced by the interaction with the incident γ-rays. Response functions were calculated using a Monte-Carlo simulation code EGS5 with a mesh width of 5 keV in the energy range of 0 to 3 MeV, assuming γ-rays irradiated parallel to the sides of the detector. The unfolding algorithm was an iterative approximation method, which is independent of the initial guess. As a result of repeated measurements using ¹³⁷Cs, ⁶⁰Co, ¹³³Ba, and ²²Na radiation sources, the energy resolution is about 0.01 MeV and the measurement accuracy of γ-ray fluence rate is almost constant at less than 10% in at 0.3 MeV and above.

Fukushima Daiichi (1F) Nuclear Power Station is preparing to begin Fuel debris retrieval (FDR). The FDR interventions should be carefully considered to maintain safety. First, the trial FDR will gather information on the situation inside the preliminary containment vessel (PCV). This information alongside data from previous investigations will inform the FDR plan moving forward.

The further expansion of FDR for Unit 1 and 3 is planned based on “Mid-and-Long-Term Decommissioning Action Plan”, and Unit 3 is prioritized depending on the viewpoint of securing the work area for accessing to the penetration X-6 and others.

The further expansion of FDR shall be continuously revised based on the information and experience that will be acquired in the future, and it is also important to integrate the results of studies, research and development toward FDR.

The LONG Term OPerationS (LongOps) project, a collaboration with Remote Applications in Challenging Environments (RACE) and Sellafield, is starting to progress forward the further expansion of FDR. Two key areas of the project are Long Reach Manipulators and Next Generation Digital Mock-Ups (NG-DMUs). FDR presents many challenges - confined space, high radiation, unpredictable conditions, etc. These challenges need to be anticipated by conducting key studies into the situation and clearly capturing the specific requirements for the NG-DMU in 1F. This will allow for flexible planning to respond to the challenges when they are encountered. The key studies are closely related to the solutions for the challenges and they should be considered from the perspective of the actual work, which plan to create new technology to address future issues in FDR.

This brief paper provides an overview of the LongOps research and development for the NG-DMU supporting FDR in 1F.

Focused on cutting of submerged objects an experiment investigation is performed on the cutting ability of water abrasive suspension jets (WASJs) under different working conditions. An in-house WASJ device with working pressure of 30MPa is used for experiments and the abrasive weight concentration is set to be 22.5 % approximately. The processing ability of the WASJ system is evaluated by cutting tests of aluminum alloy specimen both in-air and underwater conditions. Under submerged condition cutting tests are conducted by ventilating to the WASJ at different flowrates. Experimental results show that the kerf depth of air-ventilated WASJs depends on the standoff distance of cutting object and varies greatly with the ventilation flowrate. When the ventilation flowrate is increased to the range of 60~90 L/min the kerf depth of submerged cutting reaches to its maximum. Within the area of X/d 36 the relative kerf depth reaches to 1.0 and the cutting ability of air-ventilated WASJs is enhanced to the same level as that in atmospheric condition. The effectiveness of air-ventilation to submerged WASJ on its cutting performance is verified. Therefor, the feasibility to apply WASJs to dismantling of radiological structures of nuclear power plant under submerged environment is demonstrated.

For the safe decommissioning of the Fukushima Dai-ichi nuclear power plant, acquiring information on the composition of radioactive nuclides as well as their quantities are required. While an analysis based on a radiation measurement is one of the effective tools, the trace detection of the nuclides with a long half-life as well as the alpha radionuclides is not feasible. For this purpose, mass spectrometry is generally used as an alternative, which requires an ultra-high vacuum system, and therefore rapid measurement/continuous monitoring on site is rather difficult. Substitutionally, the methods based on highly sensitive laser absorption spectroscopy with an optical resonator have a large possibility to realize the rapid analysis of radionuclides with a compact and transportable apparatus. In this paper, our recent progress regarding the development of a glow-discharge cavity ring-down spectrometer for the real-time nuclides analysis for the aerosol particles generated by laser processing is presented.

IRSN has been involved with CEA for some years in projects led by ONET Technologies (OT) and funded by Japanese government (METI). These projects aim at evaluating a strategy for the retrieval of fuel debris fell over the floor of Fukushima Dai-Ichi (1F) reactor pedestal, by a laser cutting technology developed by CEA and adapted by OT for a potential use in the 1F site.

However, an important issue regarding aerosols dispersion and potential release into the environment is present around this well-proven technology and must be under control. To do that, IRSN has contributed in this project to provide quantitative information to assess the risk induced by aerosols generation and dispersion. Concomitant experimental and numerical simulations (CFD) studies were led in order to characterize the source term of aerosols from laser cutting of fuel debris simulants (Porcheron et al. 2020), to characterize their fate inside the reactor pedestal by the contribution of airflows coming from the laser cutting head (Gelain et al., 2019) (Gélain et al., 2018) and to propose solutions to help mitigating dispersed aerosols (Porcheron et al. 2019) (Sun et al., 2020) (Porcheron et al., 2021a) (Porcheron et al., 2021b).

The present article is dedicated to summarizing a global CFD study led to evaluate the impact of fuel debris laser cutting in a representative design of 1F2 reactor pedestal in terms of aerosols dispersion and mitigation means.

Given that laser cutting (as cutting in general) produces a large amount of aerosols, laser cutting head was first equipped with a local particle collection system close to the emission source. This system includes an extraction device (a flexible duct and filtration systems) allowing to catch a large part of emitted aerosols with a particle size distribution characterized by a mass median diameter of around 300 nm. However, even if this particle collection means is very efficient, a part of emitted aerosols is not collected and still dispersed in the pedestal. Hence, to collect them, IRSN proposed to implement a spray system, commonly used in French PWR to washout fission products in the event of a severe accident.

This article presents different results of CFD calculations showing the necessity of mitigation means to avoid high aerosols concentration in the pedestal which could lead to a release into the environment. These calculations give also quantitative information about the performance of intended collection systems (extraction and spray) implemented in a representative geometry of 1F2 reactor pedestal of Fukushima and their complementarity. Finally, they highlight the evolution of the aerosols emitted during a cutting scenario as well for the phases during which the laser cutting (cutting phase) is activated as the phases during it is stopped (non-cutting phase). This last part will allow to discuss about the options to help decreasing faster the aerosol concentration during the non-cutting phase.

CEA has manufactured a series of Fukushima Daiichi fuel debris simulants, either with depleted uranium oxide or with hafnium oxide as a surrogate of UO2. In ex-vessel compositions resulting from an interaction between corium and concrete, the oxidic phase density becomes lighter than that of the metallic phase, which segregates at the bottom. Three of these metallic phases have been mechanically cut at CEA Cadarache with handsaw and with core boring tool in FUJISAN facility.

It appeared that two of these metallic blocks were extremely hard to cut (one from a fabrication with uranium oxide, the other from a simulant block) while the last one was more easily cut. The similarities and differences in metallographic analyses (SEM-EDS and XRD) of these three metal blocks will be presented and discussed. This experience provides useful learnings in view of the cutting and retrieval of fuel debris from Fukushima Daiichi.

We developed a microwave-enhanced laser-induced breakdown spectrometer (MW-LIBS) to produce strong emission intensities and a high signal-to-noise ratio (SNR) for spectrochemical and analytical applications inside the Fukushima Daiichi Nuclear Power Station (FDNPS). The MW-LIBS potentially solves the signal attenuation problem of fiber-coupled LIBS, which may experience deterioration in an intense radiation environment. This study's target material is Zirconium (Zr) metal used as the nuclear fuel tube cladding. Measurements of the emission intensity enhancement factor and SNR of the Zr are demonstrated for increasing microwave power. Results indicate enhancement in emissions of SNR of atomic Zr I and ionic Zr II increasing several hundred times higher compared with the standard LIBS. The interaction of the microwave and the plasma enhances the excitation and recombination cycle of the Zr I. Consequently, the ionization and molecular ZrO radical formation were also enhanced and sustained for a more extended period. The lifetime of the emissions lasted more than the microwave duration of 1000 μs. The microwaves also enlarged the plasma 20 times its initial volume.

The immobilization of the radionuclides produced in the decommissioning of Fukushima Daiichi nuclear power plant (1F) is a growing priority for waste management. Here, we have investigated the effect of spark plasma sintering (SPS) on the solidification process of wasteform mainly for Fukushima decommissioning to add more insight into this field. To develop the solidification process for wasteform, simulated ALPS precipitation waste was solidified. SPS technique successfully increased the density of ALPS phosphate significantly at 600 °C, and the value was three times higher than that of the normal heating sample; succeeded in significant volume reduction to 1/3. Crystallization to phosphate-type crystalline occurred at 600 °C by SPS, and this is likely the reason why significant densification occurred. Above results leads to suppression of volatile substances because of low sintering temperature and high-speed solidification process by SPS. Surprisingly, in leaching test for ALPS phosphate, the leaching of Ca, Mg, Fe and Cs was greatly suppressed by solidifying by SPS. Additionally, we demonstrated first the fabrication of hybrid waste containing ALPS phosphate powder as primary waste by encapsulating it in matrix material using SPS, which is expected to improve the safety and reliability of wasteform. Solidification of hybrid waste with several matrix materials could be achieved, and ALPS phosphate powder was stably maintained in matrix.

This paper describes a development of PHITSPlugin for the radiation behavior calculation. The developed plugin calculates a dose distribution in conjunction with Choreonoid which is a physical simulator. It was developed to contribute to estimate an integral radiation dose on the robots. We discuss a procedure for calculating the dose distribution. Also, we demonstrate to calculate the dose distribution by utilizing experimental examples.

In emergency response and decommissioning at Fukushima Daiichi Nuclear Power Station (FDNPS), remote controlled robot is required to work in harmful places instead of human workers. However, remote robot operation is not easy. In order to achieve a mission by remotely robot operation, the robot operator should understand what and how much he/she is capable to do by remote operation with the robot. Thus, a method for evaluation of robot capability is needed. Many of the missions by remote robot operation in FDNPS accomplish main objectives, while, few problems occurred in some missions. The experience from such problems in FDNPS have significant value for practical remote robot operation. The aim of this research is to develop a robot testing method based on cases of FDNPS remote operations with arm mounted robots. We survey FDNPS remote operations with arm mounted robots, and decide to develop a testing method to evaluate the robot capability for approaching a target object over an obstacle by a robot arm. We develop a test field for evaluation the capability and the test procedure to measure the metric to describe the capability using the test field. Moreover, we conduct actual tests and discussions with a team engaged in nuclear emergency response robot maintenance and training to make it more practical.

In the Organization for Economic Co-operation and Development/Nuclear Energy Agency/Committee on Safety of Nuclear Installations (OECD/NEA/CSNI), the Preparatory Study on Analysis of Fuel debris (PreADES) project was recommended by the Senior Expert Group (SEG) on Safety Research Opportunities post-Fukushima (SAREF) as one of several appropriate “near-term projects” to contribute to further international efforts to understand the 1F accident and support decommissioning efforts. As a result of the activity for five years, the PreADES project has summarized collected knowledge and expertise concerning fuel debris characterization and identified needs for fuel debris analyses that would contribute to 1F decommissioning. The project also aimed to improve the understanding of severe accidents and reactor safety assessments as well as creating appropriate and optimal methodologies for future sampling, retrieval, and storage of fuel debris. Consequently, the project provided important input for future international projects.

The activity of the PreADES project was carried out under the collaboration with many international organizations, and accumulated useful expertise. Following project tasks were discussed: 1) the Characteristic table that clarified characteristics of fuel debris for 1F, Three Mile Island Unit 2, and Chernobyl Nuclear Power Plant Unit 4 from various analyses and experiments, 2) Compiling informative tables for fuel debris analysis (the Analytical table, the Major issue and methodology table; the Hot testing analysis capabilities table). The first one identified analysis data to be obtained, and the last one indicated practical appropriate and optimal methodologies for fuel debris handling, 3) future international R&D framework.

Through these activities with many discussions, project partners re-realized that important aspects for the success of future collaborative initiative was to implement conditions for fluid dialogue and information exchange between involved organizations. As these organizations may have different priorities, such as interests in severe accident analyses or promotion of the decommissioning, it is important that information exchanges are established. It is only with such a constant open dialogue that more focused guidance and recommendations for the completion of debris retrieval.

Currently, post-PreADES project, “Fukushima Daiichi Nuclear Power Station Accident Information Collection and Evaluation (FACE)” project has been launched. In the project, actual sample analyses and international round-robin analysis activity with debris simulants are being discussed continuously.

The CLADS-MADE-04 is the next test in the series aiming at understanding of the melt propagation behaviour in the lower core region. In this contribution, recent results of the post-test analysis including microstructure of metallic debris investigated by Electron Probe Micro Analyzer (EPMA) are discussed. During the test, melting of the control blade happened with sudden wave of strong heat release relatively slowly (several cm/min) spread from the hottest area downwards along the degrading control blade and channel box consuming the walls made of Zircaloy-4. A significant damage happened with the sample supporting plate as well. The investigation of microstructure of such metallic debris would allow understanding of a mechanism of enhanced local core degradation. The nature of strong heat release and possibility of spreading to the surrounding materials is to be confirmed after thorough phase identification by EPMA. The difference between Fe-B eutectic debris and Zr-Fe eutectic debris will be outlined. It is especially important for understanding of the lower core plate melt-through and a possibility of a Zr-Fe molten material progression into the lower plenum.

The interaction between the BWR fuel cladding and fuel components was examined to estimate the conditions of fuel failure (temperature and location) in the early phase of Fukushima Dai-ichi Nuclear Power Plant (FDNP) accident. In the first series of tests, four types of specimens were fabricated using commercial fuel cladding and fuel components and heated up with or without high temperature steam to simulate the interaction at each location where the fuel cladding and fuel components would be in contact. In the second series of tests, two types of specimens with encapsulated helium gas were heated up in argon gas to examine the effect of internal pressure on the fuel failure at the plenum and the spacer locations in the steam starvation condition. In both series of tests, some of fuel claddings were pre-oxidized to examine the effect of surface oxide layer formed in the normal operating condition.

The first series of tests demonstrated that apparent eutectic reactions first started near Ni-based fuel components such as rod-spring and ferrule-spring at around 1400 K in the steam starvation condition but this temperature increased with steam supply increased. Barrier effect of oxide layer of fuel cladding was also found. In the second series of tests, mechanical failure and wall-through eutectic reaction occurred competitively at the spacer location, while only mechanical failure was observed at the plenum location.

We have proposed to apply a long-reach articulated robot arm for the internal investigation of the Fukushima Daiichi Nuclear Power Plants. This report describes the development of the articulated robot arm. To search at the bottom of the Reactor Pressure Vessel (RPV), we need to deliver measurement devices five meters above the X-6 penetration in the case of the No.2 unit. Thus, we develop a lightweight telescopic boom that can be mounted at the tip of the 10-m-long articulated robot arm “Super Dragon.” We increased the joint stiffness at the distal joint by introducing a torsional spring mechanism to achieve posture change of the telescopic boom from horizontal to vertical. We successfully reconstructed the surrounding environment using Structure from Motion by camera images obtained by the experiment.

It is necessary to accurately grasp the circumstances for the decommissioning plan of Tokyo Electric Power Co. Inc.'s Fukushima Daiichi Power Station. The decommissioning of Unit 1 will be carried out first. Since March of 2022, TEPCO has been investigating the inside of Unit 1's containment vessel using underwater robots. At the entrance of the pedestal that supports the reactor, it was confirmed that the reinforcing bars and the concrete around the inner skirt had become exposed. These are extremely important for earthquake resistance. If a seismic event occurs, the seismic force on the reactor will not be transferred to the bedrock supporting the power plant. According to our trial calculation, if an earthquake of 440 gal or more occurs, it is predicted that the pedestal base would fail and the reactor would collapse. First of all, it is necessary to urgently take countermeasures against the reactor falling over. This document proposes an emergency countermeasure plan. After that, we propose a decommissioning plan for the entire power plant that can be implemented with current technology.

The explosion that took place at Fukushima Daiichi Nuclear Power Plant Unit 3 (1F3) is supposed to be caused by a leakage of gas from the primary containment vessel (PCV). To deepen the understanding of the 1F accident scenario and improve the safety measures to prevent a similar accident from happening again, the relationship between the PCV->Reactor Building (R/B) gas-leak mechanism and the hydrogen explosion in 1F3 is estimated. The leak spots and the leakage behaviors (beginning time, flowrate, composition, etc.) were assumed according to the evidence shown in R/B inspections and PCV pressure behavior, respectively. Due to the uncertainties of these assumptions, a series of numerical analyses were performed using GOTHIC code by varying the leak spot and flow rate until the simulated leaked-gas distribution right before the explosion be interpretable to the observations from the explosion video and R/B inspections. As a result, the study suggested that; (1) most of the gas leaked from the drywell top head flange to the 5th floor (operating floor) through the shield plug, and then distributed gradually to the whole building through the hatch and the stairs. (2) As the continuous leakage proceed, the hydrogen and steam ratio in R/B increased rapidly, while the oxygen ratio decreased correspondingly. (3) Until the last moment before the explosion, an oxygen deprivation condition has been formed on the 5th floor. (4) Meanwhile, a proper portion of the hydrogen from the 5th floor and the residual oxygen was accumulated on the 4th floor, which makes it explosive. It may be one of the reasons that lead to the explosion observed on the 4th floor before the big eruption. (5) On the other hand, incomplete combustion might take place on the 5th floor due to oxygen deprivation. The remained hydrogen might keep burning after escaping into the atmosphere, and eventually led to the big eruption.

Neutron induced γ ray spectroscopy (NIGS) has been conducted for each material anticipated to be contained in retrieved fuel debris and waste from PCV units 1, 2, and 3 of 1F. At PFFF of JAEA, γ rays from capture reactions are measured from ²³⁹Pu and ²³⁸U contained in PuO₂ and UO₂ samples loaded in an accounting device, as well as fission prompt γ rays. Concurrently, γ rays from capture reactions in polyethylene, steel, aluminum tube, cadmium plates, which are components of the device, are identified. Additionally, capture γ rays are measured for samples of zirconium, copper, gadolinium, and concrete as well as de-excitation γ ray from inelastic scattering of ¹¹B in a boron carbide sample. Since we can identify those γ rays, material classification by NIGS is promising.

Fauske and Associates (FAI) is performing experiments and modeling for Tokyo Electric Power (TEPCO) for Fukushima fuel debris removal, potential processes, and interim storage. The purpose of our work is to provide a technical basis for TEPCO to select process steps and interim storage options. This paper discusses lessons learned from four recent pertinent projects for damaged fuel remediation: Three Mile Island (TMI), Hanford Spent Nuclear Fuel Project (SNFP), and Sellafield First Generation Magnox Storage Pond (FGMSP), and Sellafield Magnox Swarf Storage Silo (MSSS).

The diffusion of components of nuclear fuel debris formed during the Fukushima Daiichi nuclear accident to cooling water is the root cause of the contaminated water. This study aims to investigate the role of microorganisms in the dissolution process of fuel debris. Mixed powder of UO₂ and Fe(0) was used to simulate the fuel debris because it has been estimated that they account for more than 60 wt%. By exposing the UO₂/Fe(0) to microorganisms in a medium for some days, the dissolution phenomena of U and Fe were observed. It was found that microorganisms significantly accelerated the dissolution of U and Fe from fuel debris into the medium. The dissolution ratio of U and Fe was not in the same order of magnitude, i.e., heterogeneous dissolution. These results suggest that microorganisms can lead to the disintegration of nuclear fuel debris.

The metallographic condition and influential factor of liquefaction between as-received/pre-oxidized Zircaloy and SUS/Inconel was examined by an isothermal annealing experiment with diffusion couples. Consequently, the liquefaction was confirmed at a eutectic point of Zr-Fe/Zr-Ni system around 75at%Zr and the kinetics was controlled by the mass flux of alloy’s elements to Zircaloy/liquid interface (dissolution rate of Inconel was 1.93 times of that of SUS). The oxide layer on Zircaloy delayed the onset of reaction for SUS, whereas it did not for Inconel. This may be attributable to lower activation energy of Ni in α-Zr(O) compared to that of Fe.

In the case of performing 3D reconstruction from images called photogrammetry, we assume that it can be applied alignment and size adjustment of multiple reconstructions based on the object when the known objects exist in the reconstruction target area. We adopt QR codes as a known object. However, we could not reconstruct the QR code attached to the target in some preliminary experiments. In this study, we verify the shooting method so that the QR code attached to the reconstructed model can be detected with simulation. At first, we create the target object attached QR code. Next, we generate image data with a simulation of linear orbit shooting. We verify the QR code in the reconstructed model and performed 3D reconstruction from these images, and the shooting condition detectable QR code while changing the camera angle, QR code size, and shooting interval. The verification result shows that the size of the QR code affects QR code detection strongly. Moreover, in the case that the size of the code is large, the QR code can be detected by reconstructing the model from images where the QR code cannot be detected due to an angle problem. In contrast, shooting interval has weakly effect on detection. However, there is a QR code that cannot be detected by reconstructing results from large images. For these results, we clarify that it is important to choose images to be adopted in 3D reconstruction for QR code detection.

On 11^th March 2011, the Great East Japan Earthquake and Tsunami hit the Fukushima Daiichi Nuclear Power Station. Units 1–3 lost all power supplies, which set in motion a chain of events that led to severe core damages and releases of radioactive contamination into the environment. Since then, TEPCO has made many efforts to investigate the status of the reactors and containment vessels utilizing remotely operated vehicles (ROV) providing visual information from inside the Primary Containment Vessels (PCV), including Reactor Pressure Vessel (RPV) pedestal areas of Units 2 and 3.

Through the previous investigations of Unit 1 PCV, presence of large amount of foreign materials (deposits) was confirmed on drywell (D/W) floor. To clarify the distribution and nature of these deposits, as well as the state inside the PCV, set of 6 ROVs with a specific purpose, namely installation of guide rings (A), visual observation (A2), 3D mapping of deposits (B), measurement of deposit height (C), fuel debris detection (D) and sampling (E) have been developed by International Research Institute for Nuclear Decommissioning (IRID). Through January to June 2022, the first half of the investigations (ROV-A, A2, C) has been successfully completed, revealing significant degradation of the pedestal wall concrete in the vicinity of the personnel access port (hereinafter referred to as pedestal opening), deposit shelves attached to concrete/metallic structures, various morphology of deposits, and height of the deposits ranging between 0.2 to 1.1 m from the D/W floor in the investigated range.

These results are examined in the light of the assessments by severe accident analysis codes and experimental results, and discrepancies between the expected and observed states are pointed out. The volume of the deposits outside of the pedestal, their distribution and extent of damage to structures suggest that the current knowledge of severe accident phenomena, such as ex-vessel accident progression, molten core-concrete interaction (MCCI) and others, is incapable of providing comprehensive explanation of the state inside the Unit 1 PCV.

The radioactive nuclides dispersed without control to on- and off-site of Fukushima Daiichi nuclear power station (F1NPS) at its accident in 2011. A large amount of radioactive solid waste has been generated during its decommissioning. Radioactivity inventory of the solid waste should be estimated for R&Ds on waste management. Cobalt-60 is often selected as a key-nuclide for determining activity in solid waste generated at nuclear power plant, and is also detected in various F1NPS waste. In this study, radiochemical analysis data opened to public was investigated for correlation between Co-60 and nuclides of activation and fission products as well as actinides (H-3, C-14, Ni-63, Sr-90, Tc-99, I-129, Cs-137, Eu-154, Am-241, Pu-239+240). Cobalt-60 well correlated with Ni-63, Eu-154, Am-241 and Pu-239+240 in spite of differences in physical/chemical properties and process of contamination. This finding suggests a possibility to apply Co-60 as a key-nuclide of scaling factor method.

Cesium (Cs) penetration and leaching experiments on concrete of reactor containment structural materials are contributed to understanding of contamination mechanism by radioactive materials, which is one of six important issues identified by the Nuclear Damage Compensation and Decommissioning Facilitation Corporation (NDF). Accurately understanding distribution of Cs in concretes is important in order to decide how to proceed the Fukushima Daiichi (1F) NPP’s dismantling. The contaminated concretes are exposed to the environment changes, which means that humidity and Cs concentration around them may change depending on the climate and the plant conditions. Consequently, its penetration into them, and leaching from them repeated for long time. This complicated mechanism must be clarified.

In our presentation, the possibility will be showed that Cs penetration/leaching behavior in contaminated concretes exposed to the environment changes (humidity) like dry and wet could be complicated more. Also, the present plans to study these behaviors using INAA analysis and FEM simulation will be attached.

The analysis of small samples retrieved from the inside of the Primary Containment Vessel (PCV) of Units 1, 2, and 3 at the Fukushima Daiichi Nuclear Power Station (FDNPS) detected various types of U-bearing particles. Elucidation of the formation mechanism of these particles is expected to be a good practice of the real debris characterization. However, the current sample analysis could not accurately measure the oxygen concentration in the small particles, which is necessary for analyzing the debris formation mechanism. In this study, we attempted to analyze the solidification path of the particles by three-step thermodynamic approach. Step-1: The proper quasi-binary phase diagrams were drawn for the solidification path analysis. Step-2: The four prototypic solidification path patterns were identified based on the assessed phase diagrams. Step-3: The probable solidification paths of the U-bearing particles were discussed using these solidification path patterns and sample analysis results.

The following interesting tendencies of U-rearing particles were recognized: 1) Tetragonal phase, which was detected in all U-bearing particles in the sample analysis, is evaluated to stably exist only at around 1200 ℃ in the U-Zr-Fe-O system in the equilibrium condition. 2) Pattern II was reasonably identified as an only solidification path available for the Units 1 and 2 U-bearing particles, whereas pattern IV would be additionally possibly available for the Unit 3 particles.

From these thermodynamic analyzes, the following characteristic (or formation mechanism)s is speculated for these U-bearing particles:1) The debris in PCV are likely to have been solidified by gradual cooling from high temperatures (molten state) to intermediate temperatures (approximately at around 1200 °C) and solid-solid transition at lower temperatures may be limited. 2) Units 1 and 2 debris might be exposed to slightly hypo-stoichiometric conditions than Unit 3, and whereas Unit 3 debris might have a wider variation in the oxidation degree.

In order to estimate the reliable risk and safety required for retrieval of the fuel debris from Fukushima Daiichi NPP damaged reactors, it is important to understand the degradation of the fuel debris. Because groundwater merges into the contaminated water through cracks caused by PCV damage, microorganisms in the groundwater were introduced into PCV. This fact indicates that the fuel debris is in contact with the microorganisms in groundwater until the dismantle and the retrieval. We investigated the effect of siderophore-releasing microorganisms (SBs) on fuel debris. Fuel debris analogue pellet samples (FDAPSs) were formed by melting and solidifying CeO₂ (an alternative to UO₂)-ZrO₂ solid solution, and metallic iron. FDAPSs were contacted with two kinds of SBs on a membrane filter placed on agar medium for 50 days.The contacted pellet samples, the filter, and the precipitates deposited on the agar medium were analyzed by SEM-EDS, a secondary ion mass spectrometer (SIMS), Rutherford Backscattering Spectrometry (RBS), and Elastic Recoil Detection Spectrometry (ERDA). SEM-EDS analyses showed that the degradation products containing Fe were detected on the pellet samples, on the filter, and on the agar medium, indicating that Fe ions were dissolved, detached from the pellet samples, and migrated through the filter by the addition of SBs. Some portion of Fe were detected in the SBs cells. RBS and ERDA analyses showed that the degradation products were determined as Fe oxyhydroxides. Although Zr and Ce were detected in small amounts on the filter by SIMS analysis, dissolution of Zr and Ce may be limited. Therefore, presence of siderophore releasing microorganisms accelerate degradation of fuel debris in which Fe metal regions are preferentially degraded.

At Fukushima Daiichi Nuclear Power Plant Unit-2 (1F2), three reactor pressure peaks occurred after the forced depressurization by SRV opening at around 18:00 on March 14, 2011. Based on previous investigations, it is estimated that during the first pressure peak from around 20:00 to 21:20 on March 14, the core temperature increased and hydrogen was generated by zirconium-water reaction.

To estimate the chemical form of the released radioactive materials from the fuel and the core structure damage behavior, the temperature and the hydrogen/water vapor ratio in the core during this period was studied. The factors that affect the temperature and the hydrogen/water vapor ratio in the core were modeled. Based on this model, the distributions of temperature and hydrogen/water vapor ratio in the core were evaluated between the end of forced depressurization at 18:40 and the start of the second pressure peak at 22:40 on March 14th.

Based on the evaluation results, it was found that around the first pressure peak, zirconium-water reaction progressed mainly in the lower part of the core, which led to depletion of water vapor. On the other hand, the upper part of the core tended to become a reduced atmosphere of almost exclusively hydrogen, which condition is consistent with that for the formation of insoluble Cs particles.

Based on above discussion, new hypotheses regarding accident progression scenarios according to the observed events were proposed, such as the possibility that a leakage and blockage of the in-core instrumentation dry tubes occurred around 20:00 on the 14th, and that the neutron sources such as Cm were released from the reactor vessel to D/W through this leakage path during this period.

In the framework of the OECD/NEA ARC-F project, predictions of long-term MCCI progression by eight different computer codes have been compared for the configuration of the FUKUSHIMA DAIICHI UNIT 1 sump. Indeed, the severe accident at Fukushima Daiichi Nuclear Power Station occurred on March 11th 2011. Nevertheless, the effective cooling of the fuel debris was not restored before March 23rd. Therefore, Molten Core Concrete Interaction (MCCI) most likely occurred for more than 10 days without the presence of water (i.e., under dry conditions). Such a long period of MCCI is rarely studied.

This paper focuses on the detailed analyses of the results obtained with the TOLBIAC-ICB tool during the two steps of this benchmark. The first step consisted in a preliminary rough comparison of results allowing then to draw new converged axes of research. Results of TOLBIAC-ICB V3.4 mod2 is described and explained in relations with implemented physical models and results obtained from others codes. The second step presented simulations with more agreed initial data and boundary conditions. The detailed MCCI phenomenology is depicted and physically discussed based on the long-term evolutions of temperatures, detailed pool and crust compositions in term of chemical components, 2D ablation front, gas releases… All this advanced analysis raises the question of the validity of some models of current, intensively used, MCCI tool for very long term transients.

A technique to simply judge the presence or absence of nuclear materials in fuel debris generated in the Fukushima Daiichi Nuclear Power Plant (1F) accident is important from the viewpoint of sorting the debris from radioactive wastes. In the case of uranium(U)-containing waste from fuel cycle facilities, the U amount may be determined by the detection of gamma-ray at 1001 keV emitted by 234mPa, a daughter nuclide of 238U (Suzuki et al., 2011). However, it has been pointed out that such gamma-ray measurement techniques cannot be used for fuel debris that contains a large number of fission products (FPs), which emit various energies of gamma-ray. In this study, we focus on the spontaneous prompt fission gamma-rays (PFGs) directly emitted from nuclear materials. PFGs exists in a higher energy region than that of FPs and can avoid the interference by them. We aim to measure PFGs in simple ways.

The understanding of long-term stability of fuel debris (FD) is one of the most important issue for reduction of radiation exposure due to micro particle generated from FD during Fukushima Daiichi decommissioning. The objective of this study is to investigate the probability of long-term aging of FD under decommissioning environment. Nine types of samples which simulate FD considered to exist in the Primary Containment Vessel (PCV) were fabricated, and underwent air exposure tests and water immersion tests to evaluate the probability of micro particle generation due to chemical aging under the environments exposed in the PCV interior, or during the retrieval and storage operation. The experimental results revealed particle generation occurred on specific FD material type and environmental condition.

In order to retrieve the fuel debris from the Fukushima Daiichi NPP safely, risks such as α particle dispersion must be considered. In this paper, a new method for a large-scale fuel debris retrieval by using Geopolymer (hereinafter referred to as GP) studied is summarized.

This paper also includes a new disposal concept for various radioactive wastes and the barrier material performance as wastes when using GP. As a result of the evaluation, it was concluded that the method using GP might be effective in the process from fuel debris retrieval to waste disposal.

For the decommissioning of Fukushima Daiichi Nuclear Power Station (1F), it is necessary to analyze the state and element composition including radio isotopes of the debris. Both radiation measurement and ICP-MS analysis are excellent methods for elucidating radioactive species and composition. Both the methods are limited to be applicable to bulk samples. As for microscopic analysis, it is very difficult to analyze simultaneously isotope ratios, element distribution, and mixing states. Existing surface analysis methods, such as SEM-EDX, TEM-EDX, TEM-EELS, are generally reliable for element analysis and quantification. However, they cannot perform isotope-selective analysis. In order to elucidate the properties of debris with high accuracy is useful to evaluate the risk of re-criticality. Moreover, the evaluation of fission products needs to understand the distribution and dynamics in the reactor.

It is an extremely difficult task to realize the analysis of solid-local isotope ratios and trace elements. There are few options other than mass spectrometry to isolate and detect isotopes of target elements, and solid-local isotope analysis is limited to secondary ion mass spectrometry (SIMS). However, in SIMS, the problem of isobaric interference may occur, and it is not possible to accurately analyze the isotope of interest. Therefore, we are developing an isotope micro-imaging technique. This isotope micro-imaging technique is based on SIMS, but element-selective ionization with resonance laser ionization is newly employed, called R-SNMS. We have performed the analysis of radioactive Cs imaging for a small particle. In this paper, we report on the micro-imaging of each isotope of Nd from a simulated debris sample made of stable isotopes, assuming an analysis of 148Nd, which is important for understanding the burnup of fuel debris.

This paper introduces the initiatives for the decommissioning of the Fukushima Daiichi Nuclear Power Station.

In particular, we introduce the implementation status of onsite preparations, etc., and future issues and initiatives, which are currently focusing topics, such as Fuel debris retrieval, Waste management, Contaminated and treated water management, Fuel removal from spent fuel pools and Analysis strategies for promoting decommissioning.

11 years have passed since the accident of TEPCO Fukushima Daiichi Nuclear Power Station (NPS). All units at the site are being kept stable. The monitoring data indicates that the environment impact on the site and surrounding area have been significantly reduced.

Things are going back to normal in the surrounding area of Fukushima Daiichi NPS. The lift of evacuation order is progressing. Infrastructures such as railway and town offices are being restored. Former residents are beginning to return to formerly restricted areas. Local festivals are starting to be held again. Industries including fisheries are also revitalized.

Import restrictions of food by other countries and regions are being lifted. Last year, Singapore and the United States lifted the restriction. The United Kingdom and Indonesia also lifted this year.

Reconstruction and decommissioning in Fukushima is progressing concurrently and steadily.