The general context of the article is to evaluate strategies that can be used to mitigate aerosol dispersion during the fuel debris or corium retrieval from Fukushima damaged reactors.

We propose to study various mitigation means, such as the spray scrubbing technology used to collect airborne particles, resin coating and local collection.

In the frame of the dismantling of Fukushima-Daiichi nuclear reactors, different projects were led to investigate the means to avoid radioactive aerosols dissemination during operations of fuel debris extraction. Indeed, these operations may produce a large quantity of aerosols, especially during the fuel debris cutting phase, but also during the phase of fuel debris transport in the event of a fall leading to the resuspension of radioactive matters. Furthermore, this resuspension phenomenon may also occur by the sole action of global ventilation of the PCV.

Knowing these risks, it was essential to evaluate solutions allowing to prevent the aerosols dispersion. For that, IRSN is involved in projects with ONET and CEA, and especially in the RESIN project dedicated to demonstrating the feasibility and the efficiency of fuel debris coating with different kinds of resin. These resins are of different natures and different viscosities, easing their applicability depending on the material to coat and its orientation. However, the covering of fuel debris spread on the floor may generate a potential increase of temperature under the resin which may lead to a loss of fuel debris cooling, and a potential fuel debris degradation and hydrogen production. Furthermore, the resin degradation may also occur, promoting the aerosol dispersion during cutting operations.

To answer this issue, a numerical study was conducted with the CFD code ANSYS CFX. It consists in evaluating the impact of the fuel debris coating by taking into account the residual heat of fuel debris especially just above the sumps which are full of fuel debris. A calculation was carried out to determine the temperature field at the interface between the sump and the fuel debris spread on the floor, and this temperature field was implemented as boundary condition in a calculation of heat transfer inside the pedestal by considering a layer of water and a layer of resin. Different parameters were modified to ensure the robustness of the calculation and the temperature values calculated.

The results presented in the article show that there is a strong impact of the transfer coefficient with the water just above the ground: for most cases, the temperature is low enough to avoid safety risks, but for some of them it can lead to a temperature of the water and the resin upper than 100 °C.

The 3D reconstructed image results of the inner pedestal region of Unit 3 at the Fukushima Daiichi Nuclear Power Plant underscore the critical need to understand the interaction between highly viscous corium and large-scale structures. This interaction is essential for accurately estimating the distribution of fuel debris within Unit 3. In this study, Fluid-Structure Interaction (FSI) solver is developed to calculate these interactions. Moving Particle Hydrodynamics (MPH) with an implicit pressure and viscosity solver is employed to model both the highly viscous fluid and structural components. The governing equations are modified to ensure that both fluid and structure maintain angular momentum, which is vital for accurately representing rigid body orientation. The structural material is modeled as a rigid body and calculated using Passively Moving Solid (PMS) methods. A penalty method, incorporating springs, dashpots, and friction sliders, is applied to simulate solid-solid interactions, similar to the Discrete Element Method (DEM). Phase changes between solid and fluid are modeled by adjusting the dynamic viscosity value. For verification and validation (V&V), several benchmark studies, including rigid body dam failure scenarios, were conducted. The results demonstrate the method's capability to accurately reproduce rigid body motion, both in terms of translation and orientation. Light rigid bodies were splashed by water, while heavier ones sank due to the density ratio. In melting scenarios, the method conserved angular momentum, successfully reproducing the rotational motion of complex rigid body geometries. Overall, the robustness and efficiency of this calculation method suggest its potential for simulating large-scale corium-structure interactions.

Meticulous planning is essential for debris removal from the damaged reactors at the Fukushima Daiichi Nuclear Power Station to avoid dispersing radioactive material into the environment. The debris retrieval requires a prior phase of cutting to extract debris pieces after pieces before further processing and long-term storage. Several cutting methods are considered due to the high heterogeneity of fuel debris, as well as the varying operating conditions on the different locations. Both mechanical cutting and laser cutting are contemplated to achieve this goal. These two techniques have been proven to generate dust, in the present case, from a highly radioactive material, that will disperse in the containment vessel and might deposit on the remaining structures and walls. The integrity of the containment vessels cannot be proven, and it is essential to ensure that the particles would remain inside the vessel and would be captured efficiently by the counter-measure methods.

To be able to predict the local aerosol concentration evolution, it is aimed to develop a transfer function that takes into consideration key input parameters, defined by the cutting method and by the atmospheric conditions, in addition to other continuous phenomena that will affect the physical properties of the aerosol and consequently the local concentration.

This present work is devoted to the preliminary study of the agglomeration dynamic of a simulant dust and its potential deposition on stainless steel surfaces.

Dust agglomeration is occurring substantially in a matter of tens of minutes or hours and should quantified within the scope of transfer function development, as it will affect the particle’s dynamics and the efficiency of mitigations measures.

Regarding dust deposition, the preliminary observations show that clean and dry stainless steel does not promote particle deposition.

In this paper, we describe the results of a feasibility study of a noise reduction method from images using deep learning technology for decommissioning work. Currently, remotely operated robots have been used for the decommissioning work at the Fukushima Daiichi Nuclear Power Station (FDNPS) due to the high radiation environment. we have been conducting research and development for providing clear images during operations by removing only noise from images containing noise to contribute to safe and secure decommissioning work. Since we do a feasibility study of the noise reduction method using deep learning, the main target is not the video, but rather images, which are components of the video.

We adopted the approach of building a learning model that can cope with various types of noise by training many noisy images in the deep learning process. In particular, a network called Noise2Noise was used in training to create a model to remove noise in the images. As a result of the noise reduction process, we confirmed the noise in the images was reduced. To quantitatively verify the effect of the noise reduction method applied to the images, a quantitative measure of image quality was calculated by Blind/Referenceless Image Spatial Quality Evaluator (BRISQUE), which is a kind of No-Reference Image Quality Assessment. As a result, the BRISQUE scores of the images improved after the noise reduction process. This result suggests that the processing was achieved without any loss of image quality.

Estimation of the characteristics and distribution of fuel debris in the Fukushima Daiichi Nuclear Power Station Unit 2 is of growing importance for safe decommissioning for the first retrieval. In this study, we focused boron included in the control rod as neutron absorber, and estimated the amounts of boron evaporation during the early phase of core degradation (between 18:50 to 22:50 in 14th March in 2011) and chemical forms of boron accompanied with fuel debris. First of all, possible chemical reactions under accident condition were investigated based on the review of literature and research report regarding the control rod degradation. Secondly, the influential reaction during degradation of control rod was chosen based on thermal hydraulic condition evaluated with the plant parameters. Finally, it was estimated that approximately 20% (from central region of core) and 55% (from peripheral region of it) of initial boron inventory could have been released to gas phase by the timing of core support plate failure. Chemical forms of boron accompanied with fuel debris were estimated as the metallic boride ((Fe, Cr)B_x) or the ferric borate oxide (FeB_xO_y).

This paper describes the development of an intuitive robot teleoperation interface. In robot teleoperation, time delays disturb real-time operational command transmission to the robot, and it makes maneuvering the robot difficult for operators. Under such delay conditions, the operators feel that their own efficacy of robot operation is reduced. For coping with this issue, we focused on operator’s subjective feeling, and we designed a robot teleoperation software-user interface which allows the operator intuitively maneuvers the robot under the delay. Moreover, we conducted preliminary experiment for surveying operator’s subjective feeling. The result of the experiment suggests the possibility that the loss of subject’s subjective feeling was reduced by the developed user-interface and the maneuverability of the robot is improved.

Fuel debris retrieval at the bottom of the primary containment vessel (PCV) is one of the significant tasks for the decommissioning of the nuclear power plant and in particular for 1F. It is challenging for conventional manipulators to perform the retrieval process due to the presence of radiation, water leakage, and poor lighting conditions. We tackle those problems with the design and fabrication of a novel mechanical manipulator and its control and navigation algorithm. CVT (Continuous Variable Transmission)-based actuation improves the robot’s shock resistance. AI-based navigation algorithm enables semiautonomous navigation and grasping in the cluttered environment inside the PCV.

First, we investigated the shock-resistant mechanisms for the drive train and the gripper part. The drive train features CVT-VIA which utilizes the toroidal CVT for the adaptive gearing. In addition, the flexible debris gripper makes use of the spring joint instead of the conventional axial joint. The tuning of ligament allocation enables the designer to devise the spring joint with the arbitrary deformation characteristics.

Second, we developed a navigation system to overcome the obstacles. The human operator can control the robot end effector by clicking the target point from the vision input. The method features the ML-based approach to overcome the cluttered conditions.

Third, we validated the approach with a real-world experiment. The experiment was done at NARREC.

In conclusion, the decommissioning robot manipulator features the CVT-based actuation and a learning-based navigation system. Future works include the development of the whole manipulator with CVT-VIAs and the integration of the navigation system with the compliant actuators.

A work planning system for decommissioning the Fukushima Daiichi Nuclear Power Station has been developed. This system, designed to formulate robust plans without experience using typical work-centred decommissioning methods and processes, is a significant step towards improving work plans, including continuously dismantling equipment and related resources, such as remotely operated robots and human workers.

The processing speed for optimising the work schedule is proportional to the square of the number of activities. In an inexperienced project, such as the Fukushima Daiichi decommissioning project, the person-hour calculation formula for the dismantling quantities has yet to be determined. Hence, we need to examine how long hours it takes to complete each activity with a mock-up facility and physical simulation corresponding to several scenarios. Moreover, we must evaluate optimal task durations by combining potential activities after acquiring possible person-hours for a specific scenario. It has been proven that optimising a hundred activities per month for typical dismantling work took several hours. For this reason, optimisation processing applied to many activities simultaneously has the problem of being unrealistic.

We have introduced a novel model that effectively expresses the process's flow, addressing the abovementioned issue. We calculated the work time using a discrete event simulation (DES) calculation by adjusting the parameters related to processing throughputs. In DES, multiple robot operations for dismantling equipment by various operators can be modelled with a state flow diagram. This diagram includes waiting time for events, delay time, and storage capacity limitations for repeated activities. Based on this state flow model, DES can calculate the total durations of complex combinations of activities for multiple scenarios. Since the system state is changed only when an event occurs to reduce calculation time, it is expected that simulation calculation can be performed at high speed rather than a continuous model.

Analysis of samples collected from the Fukushima Daiichi Nuclear Power Station provides information on the conditions at the time of the nuclear accident. The particle collected from the X-2 penetration of the Fukushima Daiichi unit 1 contained a fine precipitate of monoclinic ZrO₂ in a cubic UO₂-based matrix. The purpose of this study is to experimentally confirm at what cooling condition such a phase separation could be obtained, taking into account the temperature and time required for such a phase separation to occur. In this study, four patterns of cooling process were applied to U_0.8Zr_0.2O₂ to confirm the phase separation and phase transition.

U_0.8Zr_0.2O₂ cooled from 2200 °C at 600 °C/min or 10 °C/min showed no phase separation nor phase transition by observation using a scanning electron microscope and energy dispersive X-ray spectroscopy and a transmission electron microscope and energy dispersive X-ray spectroscopy. On the other hand, U_0.8Zr_0.2O₂ annealed at 1450 °C had fine phase separation. The phase of precipitated (Zr, U)O₂ transitioned from tetragonal to monoclinic by cooling at 0.5 °C/min. It was obtained that the possible condition for fine precipitates of monoclinic ZrO₂ to exist in a matrix of cubic UO₂, as in the particle collected from the X-2 penetration of the unit 1.

The decommissioning of the Fukushima Daiichi nuclear power plant necessitates advanced robotic technology to mitigate radiation exposure to workers. This study introduces a Telescopic Active Scope Camera (TASC) designed to remove contaminated water within a Reactor Building Cooling Water System (RCW). The operation involves maneuvering a suction hose through a small hole on a perforated panel after navigating approximately 10 m of complex piping. Traditional continuum robots, such as the Active Scope Camera, face challenges in precisely controlling their tips to thread such holes. The TASC addresses this by incorporating a set of outer and inner continuum robots with flexible cilia-covered bodies and a mobile tip mechanism equipped with active wheels. This outer robot travels through intricate pipes, positioning the tip mechanism to align with the perforated panel while an inner scope camera (suction hose with camera) is pushed out to thread the hole. Vibrations from the outer robot enhance both propulsion in the pipes and the insertion of the inner scope camera. We developed a TASC prototype and experimentally demonstrated its capability to accurately thread an inner scope camera through perforated plate holes, evaluating its propulsion and pull-out performance under various design parameters.

Laser-cutting feasibility for the safe and efficient dismantling and retrieval of Control Rod Drive Housing (CRD-H) units from the damaged units at the Fukushima Daiichi nuclear power plant was thoroughly investigated in this paper. To facilitate this exploration, two distinct types of mock-ups were designed and fabricated: one representing an elementary CRD-H, and the other replicating the bottom of the vessel. These mock-ups also included crucial components such as the CRD-H flange, grid, and support bars. They exhibited various outer diameters, ranging from 152 mm to 168 mm and were modular in nature, composed of several cylinders with variable thicknesses, along with a central rod. Each cylinder within the mock-ups was designed to emulate the internals of a CRD-H unit. In certain configurations, zirconia pebbles and gravels were incorporated to simulate the conditions of damaged CRD-H mock-ups. To address the risk of falling objects during CRD-H segmentation, different scenarios were rigorously tested, including piercing for inserting a 10 mm diameter pin and U-shaped cutting for a plate insertion. This approach aimed to simulate field conditions and constraints, as accurately as possible and assess the effectiveness of lasercutting technology in handling the complexities of dismantling CRD-H units in Fukushima Daiichi's challenging environment.

Laser cutting tests were conducted on elementary mock-ups in DELIA(CELENA) facility at CEA Saclay, while remote handling operability with a robotic arm and cutting feasibility of the bottom of the vessel mock-up were assessed in HERA facility at CEA Marcoule. Results demonstrated the capability to cut 168 mm diameter 304L mock-ups, filled with zirconia pebbles and gravels, using 8 kW laser power in less than 40 minutes. Components of the bottom vessel, including a 152 mm diameter CRD-H mock-up of an equivalent thickness of 141 mm and the upper part of the flange, were efficiently cut using a 14 kW laser power in less than 15 minutes. Additionally, a hanger rod was cut at a distance of 300 mm, despite accessibility constraints, in less than 35 seconds. Furthermore, demonstrations of scenarios involving the cutting and removal of the grid, along with remote handling operations for pin and plate insertion, were also successfully carried out, showcasing the versatility and efficiency of the laser-cutting process in this intricate dismantling process.

For decommissioning of the Fukushima Daiichi Nuclear Power Station, it is important to clarify the dose rate distribution inside and outside the containment vessel and the point cloud information of the structures. We developed a self-powered gamma-ray detector (SPGD) that can measure dose rates under high dose conditions. We devised a detector structure that was both compact enough to be mounted on a remotely operated survey vehicle (survey ROV) and sensitive enough to measure dose rates in low-dose-rate areas. To achieve high sensitivity, a multilayer structure was devised to increase the surface area of the emitter, insulator, and collector layers. An internal collector was provided to collect the electrons emitted by the gamma-rays, and the thickness of each layer was designed based on the electron range. We fabricated a prototype SPGD, 35 mm in diameter, 100 mm long and 0.36 kg in weight, as a compact, high-sensitivity detector. To evaluate the stand-alone performance of the detector, we conducted an irradiation test using Co-60. The measurement lower limit dose rate value was 0.06 Gy/h. The detector was mounted on the survey ROV in a mock-up facility simulating the structures around the pedestal, and noise evaluation tests during various operations were conducted to identify the noise source and implement countermeasures. As a result, we obtained the prospect that measurement in the assumed dose rate range was possible.

The containment atmospheric monitoring system (CAMS) readings for the drywell (D/W) and the suppression chamber (S/C) were obtained during the accident at the Fukushima Daiichi Nuclear Power Plant (1F) Unit-2, but the radioactive materials (RM) migration behavior within the D/W that caused the change in the readings is not clear. Therefore, an attempt was made to estimate the behavior of RM migration by analyzing the flow and radioactive source distribution in the D/W between 18:00 and 22:40 on March 14, 2011, when the core damage progressed.

The RM used to evaluate the migration behavior was assumed to be CsI aerosol, which was considered to be the representative radiation source during the relevant period. The deposition and entrainment of the aerosol to the D/W wall and internal structures were considered. The CsI aerosol distribution in the D/W obtained in this way was combined with the results of the preceding MCNP evaluation to estimate the D/W CAMS readings, and the amount of CsI entering the D/W was adjusted until it was roughly consistent with the actual measured value at 22:40.

As a result, the calculated D/W CAMS readings tended to increase due to the deposition of RM in the D/W atmosphere on the walls and other structures. It was suggested that the increase in the measured readings may have captured this effect. On the other hand, the rate of increase in the readings was underestimated. To quantitatively reproduce this rate of increase, further investigation is needed into the uncertainty of boundary conditions such as RM migration from the reactor pressure vessel (RPV) and the S/C, models such as RM deposition, entrainment, particle growth and their differences due to chemical forms, and the method of converting radioactive source distribution to CAMS readings.

The trapping behavior of cesium (Cs) on concrete, such as pedestal and shield plugs, was assessed through multiple deposition tests of CsOH aerosols on concrete main phaseCaCO₃ within a temperature range of 170oC to 570oC under humid conditions. The deposition rate of CsOH aerosols on CaCO₃ as a function of temperature was examined. X-ray diffraction analysis revealed the presence of water-soluble Cs₂CO₃ on CaCO₃ samples up to 420oC, but negligible detection above this temperature. This finding is supported by quantitative analyses of Cs weight gain and chemical analysis, demonstrating an increase in deposition rate up to 420oC, followed by a decline thereafter. The increased deposition rate is attributed to the chemical interaction between CsOH and CaCO₃, while the diminished rate beyond 420oC may result from the decomposition of CaCO₃ under steam, thereby reducing the available reactant for the chemical reaction with CsOH.

The activity ratio of short half-life noble gas fission products, ⁸⁸Kr-to-¹³⁵Xe (A(⁸⁸Kr/¹³⁵Xe)), has close relationship with the criticality (k_eff) in the primary containment vessel (PCV) or fuel debris canister, in which fuel debris are accumulated. In this paper, we investigated the time-dependence of the A(⁸⁸Kr/¹³⁵Xe) when the k_eff changes according to an incident such as the change of geometry in the PCV. Our calculation shows a slow change of the A(⁸⁸Kr/¹³⁵Xe) value with the maximum or minimum values before settling in a new stable value after several hours from the k_eff change. The applicability and the suitability of the A(⁸⁸Kr/¹³⁵Xe) time-dependent measurement is discussed for the monitoring of the k_eff in PCV.

Fuel Debris (FD) extracted from Fukushima Daiichi Nuclear Power Station (1F) reactors is expected to contain unsealed high level radioactive materials. Its handling from transportation to safe storage will pose several challenges in terms of safety and process feasibility. This paper focuses of how Orano’s experience and best practices in the design, operations, and maintenance of facilities handling high level radioactive waste (HLW) such as La Hague Reprocessing Plant can be leveraged to benefit TEPCO in the FD handling process on 1F.

Three main design requirements of HLW handling facilities are expected to apply to future FD handling facilities or equipment on 1F. The first requirement is to ensure the safety of the public and the workers operating these facilities (protection against ionizing radiation, containment of radioactive materials, management of radiolysis hydrogen generation, subcriticality and evacuation of thermal power). Then, high level radioactive FD will have to be handled in remotely operated confined spaces in which contamination management is a key point to ensure safe and reliable longterm operations. Finally, the installation of relevant analysis equipment is required to collect data on FD properties.

The following technologies and practices used by Orano for decades in its reprocessing facilities to transport, process, store and characterize radioactive materials, could be implemented on 1F to comply with the above requirements.

The PADIRAC shielded cask and airtight DPTE container is one example of a qualified, reliable, and simple solution used by Orano for moving safely small quantities of HLW while maintaining confinement during product transfer between the container and a containment structure.

Shielded and airtight hot cell which can be operated and maintained remotely is also one of the main contributors for safe handling of radioactive materials. Various solutions are used to ensure reliable management of contamination in handling of HLW, such as dynamic confinement using ventilation, contamination monitoring, container air-tight docking and the use of high-performance filters to ensure both hydrogen release and confinement of radioactive particles.

Finally, Orano has decades of experience and expertise in the operation of analysis equipment to quantify and characterize radioactive materials such as gamma/neutron dose rate measurement, gamma-ray spectrometry, gas chromatography for hydrogen measurement, and Active Neutron Interrogation, a non-destructive measurement of fissile matter in waste containers which will most certainly be required for the quantification of fissile materials in FD extracted from 1F reactors.

Operation errors have occurred in the decommissioning work using remotely operated robots due to the lack of information on the surrounding environment. Therefore, a system for constant observations inside the PCV is anticipated.

Goto et al. developed easy disposable monitoring robots to constantly observe inside the PCV and a modularized rail structure (Rail DRAGON) as a scaffold for them to move.

Since there are obstacles on the path of Rail DRAGON deployment, the joints must always be controlled to avoid obstacles. However, in the conventional procedure of connecting modules and pushing the whole rail, the power and signal line were directly connected to the power supply/control device, so the new module could not be connected without cutting the cable at once and interrupting the control. This research aims to propose a method for realizing a system that maintains joint angle control even when connecting a new module to the rail.

Through the experiment, we proved that joint control can be maintained during switching cables by constructing a detour of the power line and automatically judging the suitable moment when the fluctuation of angle signals is small.

Leaching tests of simulated fuel debris (CeO₂, (Ce, Fe)O₂, (U, Zr)O₂ and (U, Zr, Ca)O₂) were conducted by using the microchannel flow-through method. The pellet-shaped specimens were fixed with epoxy resin and polished. A Teflon sheet 0.16 mm thick with a slit 2 mm wide and 20 mm long was attached to the polished surface of the specimen, then various solutions, leachant, were passed through the slit at 6 μL/min as a microchannel. Samples of solutions, leachate, were collected every 3 hours, diluted with 2% nitric acid, and analyzed for Ce or U using ICP-MS to determine the dissolution rate.

The (U, Zr)O₂ specimens separated into two phases when Zr was above 25%, while the (U,Zr,Ca)O₂) specimen was a single-phase up to about 50% Zr, due to the presence of a few % of Ca. The porosity of the phase-separated specimens was more than 10%, whereas the porosity of the single-phase specimens was only a few percent.

Single-phase specimens show smaller dissolution rates than multi-phase specimens. This is due to the formation of cracks and voids in the specimen by density change during phase transformation, which increases the contact area. The dissolution rate of UO₂ is larger than of CeO₂ even for the same single-phase specimen and porosity. This could be because uranium is in a highly soluble oxidation state, while cerium remains in a less soluble tetravalent state. In the case of uranium-based single-phase specimens, the dissolution rate tended to be lower with higher Zr composition. When the leachant was 1M NaCl solution, the dissolution rate of CeO₂ was slightly larger than that for pure water. Conversely, the normalized dissolution rate of uranium simulated debris was smaller for 1M NaCl solution than for pure water. This might be because uranium reacts with the NaCl component to form a protective film. SEM observation of the surface of the leached portion of the specimen before and after the leaching test showed elution of fine particles less than 1 μm in diameter and elution from the edge of the specimen. No significant dissolution of individual phases was observed in multi-phase specimens. Although the actual fuel debris initially contacted with seawater and then with freshwater, it is highly likely that particles of a few micrometers in diameter were dissolved. However, fuel debris larger than a few mm can be considered to dissolve only a part of the surface.

Airborne radioactivity must be monitored at nuclear facilities. The workers for nuclear power plants are at risk of internal exposure by airborne radioactive materials. In particular, artificial radioactive materials (uranium, plutonium, etc.) have severer damage on human bodies. Therefore, it is necessary to measure airborne radioactivity concentration of these materials. Radionuclides that emit alpha-rays exist in natural, ²¹²Po, ²¹⁴Po, and ²¹²Bi. Airborne particulate radioactivity monitor ideally should detect only the concentration of the artificial materials. The energy of the alpha rays is important to discriminate the artificial (4.0~5.8 MeV) and natural (over 6.0 MeV) radioactive materials. Therefore, it is possible to identify the two materials from the alpha-ray energy information and detect only the artificial materials.

A method for on-site radionuclide analysis is utilizing the energy spectra. Currently, ZnS scintillator or Si semiconductor detector are widely adopted to the detector of the airborne particulate monitor. The ZnS scintillator does not have enough good energy resolution to perform the analysis. Although the silicon semiconductor detector itself has sufficient energy resolution, alpha-ray energy spectra become distorted due to the energy loss of alpha-particle in the air and dust accumulated on filter paper. Therefore, it is difficult to analyze the original alpha-ray energy by either kind of detector in current methods.

This method is to estimate the source information from the detector response and obtained energy spectra. In this study, the detector response was created by radiation simulation, and the successive approximation method was adopted as an algorithm for estimating the source information.

We have fabricated an airborne particulate monitor equipped with an energy spectrum spectroscopy that applies the unfolding method. In order to evaluate the performance of this monitor, we performed a spectrum measurement using the radioactive source. From the measurement results, it was confirmed that the unfolding operation can obtain the radioactivity of the source within an uncertainty range. For further evaluation, the monitor collected dust and measured radioactivity continuously for a month. As a result of dust collection, it was confirmed that the decision threshold of artificial radionuclides had achieved less than 10^-7 Bq/cm³ under the natural radioactivity 10^-5 Bq/cm³. Furthermore, we confirmed that natural radionuclides ²¹²Bi, ²¹²Po and ²¹⁴Po (6.0, 7.7 and 8.8 MeV) can be measured separately.

The general context of this article concerns the assessment of radioactive aerosols release risk during fuel debris (FD) retrieval operations at Fukushima Daiichi Nuclear Power Plant (1F). This assessment is crucial for ensuring the safety of workers and the wider public throughout the decommissioning efforts at the 1F facility.

The Dust Acquisition (DA) project, led by CRIEPI and carried out as a subsidized project of Decommissioning, Contaminated Water and Treated Water Management, funded by METI, in collaboration with ONET, CEA, and IRSN, aims to evaluate the airborne release fraction (ARF) coefficient, a key metric reflecting the amount of radioactive material suspended in the air, particularly during the mechanical cutting of prototypic FD-simulants containing depleted uranium oxide.

The experiments were undertaken by ONET/CEA/IRSN on a dedicated facility named FUJISAN2 located at CEA Cadarache. The generated particles underwent comprehensive sampling and characterization, including size distribution, morphology as well as mass and number concentration, using dedicated aerosol metrology devices.

This article will present the experimental facility, including its aerosol instrumentation, the cutting sequences, and the methodology implemented to characterize aerosol physical properties and determine ARF coefficients, using gathered aerosol data measured by the different instruments.

Comparative analyses are then presented in terms of particles mass concentration between the cutting trials on the different FD simulants, to understand the influence of the sample properties on particle generation.

In order to safely remove fuel debris from the Fukushima Daiichi Nuclear Power Station (1F), it is necessary to quantitatively evaluate the radioactive airborne particulate generated by the cutting of nuclear fuel debris. We fabricated Uranium-bearing simulated fuel debris (SFD) with In/Ex-Vessel compositions and evaluated the physical and chemical properties of aerosols generated by heating the SFDs. Based on these results, we estimated the isotopic composition and radioactivity of aerosols produced when 1F-Unit2 fuel debris are heated (or laser cut). Plutonium, mainly ²³⁸Pu, ²³⁹Pu, ²⁴¹Am, and ²⁴⁴Cm were found to be the alpha nuclide, and ²⁴¹Pu, ¹³⁷Cs-Ba, and ⁹⁰Sr-Y were found to be the beta nuclide of interest.

From the standpoint of minimizing radiation exposure to both workers and the general public, it is imperative to gain a comprehensive understanding of the characteristics exhibited by the dust dispersed during the process of fuel debris retrieval. The dispersion behavior of dust is thought to be associated with material properties, simulated fuel debris samples were made with the intention of evaluating material properties. These were made dense enough to be suitable for such an evaluation and were exposed to high temperatures similar to those experienced by the actual fuel debris.

A series of simulated fuel debris samples were prepared by mixing uranium dioxide (UO₂) and zirconium dioxide (ZrO₂) in four different ratios (100:0, 90:10, 40:60, 15:85). Subsequently, the mixtures were subjected to sintering at 1700 °C, resulting in the production of pellet-shaped samples. Then the samples were observed under a scanning electron microscope coupled with energy-dispersive X-ray spectroscopy (SEM-EDS), underwent X-ray diffraction (XRD) analysis. Additionally, the grain size, density, Young’s modulus, Vickers hardness, the oxygen-to-uranium (O/U) ratio were evaluated. Furthermore, fracture toughness was calculated from the obtained data. The observed phases in the samples followed the pseudo-binary equilibrium phase diagram of UO₂ and ZrO₂, with some variations noted upon additional heating at 2400 °C. In this study, two main trends in the changes in material properties, other than grain size and O/U ratio as a function of the composition ratio were identified. A significant discontinuity was observed at higher ZrO₂ ratios. These observations indicate that the properties of the samples are influenced by the primary matrix component.

The CLADS-MADE-05 is the next test in the series of large-scale tests performed in JAEA with a purpose of understanding of mock-up assembly degradation and melt propagation behaviour accompanied by Cs and B containing aerosol release. A possibility of Cs and B aerosols interactions in RPV and formation of a new Cs bearing contaminating phase has been shown.

This paper analyses the data on aerosol behaviour of Cs- and B- bearing aerosols when they interact in the gas line, and concentrates on the phenomenon of possible reactions on the liquid-liquid and liquid-gas interface when aerosols are interacting during the large-scale bundle degradation test.

For safety management of fuel debris retrieval in Fukushima Reactor, it is essential to figure out fuel debris dissolution and denaturation mechanisms in advance. Being submerged and cooled in sea water over a decade, interfacial dissolution and denaturation occurred persistently between fuel debris surface and cooling sea water. In this study, experiments of simulated fuel debris dissolution into various solution conditions were carried out, to investigate Uranium dissolution thermodynamic and kinetic behavior under various factors, together with dissolution mechanism difference and comparation between static batch and microfluidic dissolution of uranium.

In this experiment, dissolution experiments of simulated fuel debris (U_0.7,Zr_0.3)O₂ pellets and UO₂ pellets in various solution conditions were carried out through batch and microfluidic approach, using novel microfluidic device for micro scale dissolution. Dissolved uranium concentration was monitored by ICP-MS measurement, and surface conditions before and after dissolution were characterized by Raman, SEM and XRD.

Through the experiment, uranium dissolubility in various conditions and kinetics behavior were evaluated, and different mechanisms between bulk scale and microfluidics were speculated.

The environment inside Fukushima Daiichi is highly unstructured with unknown elements. There is not sufficient CAD data available of structure or equipment inside the reactor, and knowledge of contents may be missing or unreliable due to limited data following the disaster and subsequent activities. These factors limit options for object classification.

This paper will present the research work completed by LongOps programme to develop a dataset for an applied machine learning model to perform object detection and segmentation on the structures and contents inside the Primary Containment Vessel (PCV) (Tokyo Electric Power Company Holdings Inc. Official Website, 2023). To build the dataset, annotation, process to define boundary of an object and label it with corresponding class on an image or video, has been performed using past PCV investigation videos. This task was challenging by the presence of noise and haze caused by the water and propeller motion of the remotely operated vehicle carrying the camera(s). General classes were used to cater for the wide range of damaged equipment and structures. In total, 13 classes of components and structures including corrosion and fuel debris have been identified. To cater to the low number of annotations of some classes (also known as imbalanced dataset), new images have been created by adding noise, rotation, zoom, blurriness, and flipping to original images from PCV videos to improve the data imbalance between classes.

This study work also demonstrates the capability of detecting and segmenting components using a machine learning model. For this purpose, a Mask Region-based Convolutional Neural Network model has been used for the dataset that has been split into three categories: train, validation, and a test dataset to test the performance of the trained model. For safe operation of the robot(s), corrosion can be detected so that robot can be moved safely without further damaging the structures. In addition, fuel debris can also be detected in current trained model to help the operator in locating and removing them in a safe and efficient way.

Resonance ionization spectroscopy using narrow linewidth (~MHz) continuous wave lasers is an effective tool for element and isotope-selective measurements of trace radionuclides such as Calcium-41 (⁴¹Ca). In this study, triple (422.8 nm – 1034.7 nm – ~630 nm) and double (422.8 nm – ~390 nm) resonance ionization schemes of atomic Ca were developed, aiming for the separation of odd isotopes by (i) laser polarization-dependent selection rules, and (ii) large isotope shifts of odd Ca isotopes in the Rydberg levels. The measured ion yields of naturally occurring ⁴⁰Ca and ⁴³Ca were compared to estimate the separation factor of odd isotopes from ⁴⁰Ca. Our spectroscopic approaches could be a promising technique for the analysis of ⁴¹Ca in concrete waste from nuclear facilities, which will be investigated in the future.

In the in-vessel late phase of the severe accident (SA) at the Fukushima Daiichi Nuclear Power Station (FDNPS), fuel debris is considered to have relocated to lower plenum and then steel structures in the lower part of the Reactor Pressure Vessel (RPV) inside could have been severely damaged. To evaluate the phase status of the debris during the damaging process, it is important to understand phase equilibrium of the U-Zr-Fe-O system. In this study, phase equilibrium between metallic liquid and oxide solid of the Ce-Zr-Fe-O system was experimentally investigated as a surrogate for the U-Zr-Fe-O system in the temperature range from 1823 K to 1873 K. The experimental results indicated that oxygen solubility in metallic liquid decreases with decreasing the Zr concentration and increasing the Ce concentration. Considering the expected similarity with the U-Zr-Fe-O system, the U and Zr in metallic part of re-melted fuel debris may have been selectively oxidized to form a surface oxide layer.

This paper reports on the development of remote positioning technologies for the ITER Blanket Remote Handling System (BRHS). The international fusion experimental device ITER requires remote maintenance systems due to the high radiation environment inside the vacuum vessel. We present remote positioning technologies in radiation environments that are being verified and demonstrated as part of the development of the ITER BRHS, which will be used to replace and maintain the Blanket Module that are attached to the inner wall of the vacuum vessel.

The radiation environment in the ITER limitations on viable approaches for remote positioning. In this study, the authors developed remote positioning technologies that can be applied in radiation environments, such as virtual reality with structural simulator, robot vision using radiation-resistant cameras, and peg-in-hole operations with force feedback. We performed functionality testing using the full-scale prototype of a large manipulator and make a good progress on the testing. We are currently conducting verification testing for noncontact remote positioning using robot vision technology. In our future plans, we will conduct an integrated testing that covers operations from non-contact remote positioning to pegin-hole operations.

An investigation was conducted in which underwater robots entered the PCV basement floor area, and various sensors mounted on the robots were used to measure information about sediments and the state of damage inside the PCV. In this paper, we describe 3D mapping method of sediments by an ultrasonic sensor and the investigation results of actual plant.

One of the keys of Fukushima Daiichi Nuclear Power Plant (1F) decommissioning steps is the estimation of the Airborne Release Fraction (ARF) during the removal of nuclear Fuel Debris (FD), to evaluate the potential exposure dose for workers and the public and thus contribute to define the equipment and mitigation means for FD retrieval. The URASOL and the DA projects, respectively led by JAEA and CRIEPI in collaboration with ONET/CEA/IRSN, were proposed to obtain basic data on aerosols generation and characteristics from prototypic FD-simulants containing depleted uranium oxide cut by thermal or mechanical processing tools. The whole process developed by ONET/CEA/IRSN allows the manufacturing of specific compositions and supplying corium samples for cutting (Bouyer, 2024; Denoix, 2024), the realization of cutting tests and the on-line dedicated aerosols measurements as well as sampling aerosols (Alage, 2024), to conduct initial FD simulants and aerosols post-trial analyses (Journeau, 2022; Journeau, 2023).

This paper focuses on the initial FD simulant and aerosols chemical and microstructural analyses. Results on an ex-vessel composition of FD, named VF-U3, are given.

At the main entrance of Fukushima Dai-ichi Nuclear Power Plant (FDNP), a monitoring car detected neutrons during two periods, in the early morning of 13th March 2011 and from the night of 14th to the early hours of 15th March 2011 (TEPCO, 2022). This neutron detection at the main entrance has been an unresolved item that is essential to evaluate the accident progression in the early phase of FDNP accident (specifically, the early phase of core degradation). In the present study, a confirmatory examination was conducted by heating the slices of spent fuel pin in a hot cell to examine the release of actinide nuclides. The results indicated that, in the early phase of accidents, actinide nuclides were released in the temperature range below 1773 K and neutrons were emitted by its spontaneous fission.

Radioactive wastes containing water, such as operational wastes caused by decommissioning of the Fukushima Daiichi Nuclear Power Station and secondary wastes generated from the Advanced Liquid Processing System (ALPS), need to be stored in a stable condition for the long term. For ALPS-treated secondary wastes with high radioactivity levels among the radioactive wastes, intermediate treatment is thought to be one measure to reduce the potential hydrogen generation risk and storage capacity. It was found that the characteristics of secondary wastes include concerns about hydrogen generation from radiolysis and the need to reduce the volume of the waste in consideration of potential changes in composition and storage.(Oniki, T., et al., 2019) It was found that in this study, the applicability of pyrolysis technology using steam, which is expected to provide high volume reduction, stabilization and control of nuclide volatilization, to resin-based adsorbents and chelating resins used as Ru adsorption resins was evaluated. （Parruzot, B., et al., 2023 ;Miyatake, K., et al., 2008）

Resin-based adsorbent and chelating resin, which are used as Ru adsorption resins in ALPS, were heated to 1000°C, by TG-DSC, TG-MS. XAFS was used to evaluate the effect of steam addition and the structure of the residue after heating. To confirm the effect of steam addition, heating in an Ar atmosphere was also performed for comparison. In both cases, not so much increase in weight loss was observed. In the case of the resin-based adsorbent, it was confirmed that the gas generation temperature during heating became lower with the addition of water vapor, and weight loss occurred from the lower temperature. It was also confirmed that Ru in the residue was reduced to metal in both Ar and water vapor atmospheres when heated to 500°C. In the case of chelating resin, it was confirmed that the gas generation temperature during heating became partly lower when water vapor coexisted. It was also confirmed that Ru in the residue existed as an oxide when heated at 500°C in an Ar atmosphere but was reduced to a metal when heated at 500°C in a steam atmosphere. This result suggests that heating in a steam atmosphere may inhibit volatilization by reducing Ru, although the weight reduction rate does not increase so much.

Deploying robotic platforms for characterisation into environments that are not well understood e.g., Fukushima Daiichi, presents numerous complexities and challenges, particularly regarding sensor integration and data capture. The deployment of these technologies necessitates reliance on a variety of sensors to gather information about the surrounding environment, presenting a unique set of challenges in terms of sensor network architecture, data collection, processing, and interpretation. Further, after every survey or inspection deployment, the data gathered must be easy to access, visualise, compare and query to serve as a single source of truth during the decommissioning planning. The objective of this paper is twofold: first, to present the outcome of the LongOps project in terms of integrating live sensor data from multiple sources into a cohesive digital mock-up. Second, to present the follow-up work related to the centralisation, contextualisation and visualisation of data captured after every deployment to enhance the efficiency of decommissioning planning and its safety.

Due to the large differences in melting temperatures of the different core materials (oxides and metals), estimating the thermal status (temperature, solid / liquid state) of the core materials at the time of the Reactor Pressure Vessel (RPV) failure of Fukushima Daiichi Unit 2 (from hereinafter, Unit 2) may provide a good basis for improving estimation of the current debris distribution. In another word, the thermal status of the debris in the RPV lower plenum at the time of the RPV lower head failure determines, which components of the debris are more likely to relocate to the ex-vessel (pedestal region). One of the preceding studies with accident progression analyses of Unit 2 indicated that most of the oxidic fuel (UO₂) was unmolten when the major core relocation to the RPV lower plenum (the core slumping) took place. The purpose of this study is to estimate the subsequent thermal status of the core materials at the time of the RPV lower head failure by referring to the recorded post-core slumping pressure histories using MELCOR code.

The following present understanding / modeling limitations were considered:

●Uncertainty of the injected water reaching the RPV was considered in two cases: Water injection tuned to reproduce the estimated time of the in-vessel water depletion from the plant data (estimated March 15, 4:30) (standard case) / no injection case.

●Uncertainty in the heat transfer coefficient between the debris and the lower head wall was considered in two cases: 0.1 kW/m² (standard case) / 1 kW/m² (high heat transfer case)

●Uncertainty in estimating when the Unit 2 lower head failed from the available plant data was recognized. In the analyses, detailed stress distributions of the structures and eutectic reactions were not considered and a simple creep failure of the lower head wall was modeled.

In all analysis cases, the differences of the analyzed timing of the lower head failure were smaller than the uncertainty of the timing estimated from the recorded pressure histories. The estimated in-vessel representative debris temperature at the time of the lower head failure (2270 – 2440 K) was sufficiently low with respect to the melting point of the oxidic fuel. Thus, the melting of uranium-containing oxidic fuel at the time of the Unit 2 lower head failure was unlikely. Such estimate is consistent with the muon imaging results, which indicated significant amount of the core materials still remained in the Unit 2 RPV lower plenum after the accident.

Following the decommissioning process at Fukushima Daiichi Nuclear Power Plant (1F), various radioactive wastes have been generated. Moreover, as the recovery of fuel debris progresses, the waste containing nuclear material will also be generated. The existing wastes have been categorized based on material type and radiation levels, and are currently stored above ground. Similarly, recovered fuel debris is expected to undergo appropriate waste treatment and storage. And eventually, all of them are supposed to be disposed of stably and safely.

The disposal methods for radioactive waste generated from 1F will be selected based on previously considered in the nuclear sector, taking into account the social and economic factors relevant to the characteristics of each type of waste. Radioactive waste disposal does not simply end with the burial of the waste package; it also requires the long-term management of waste data even after the closure of the repository. Particularly, radioactive waste originating from 1F possesses different types, of radioactivity, etc. compared to typically generated from the nuclear sector, and it requires highly transparent and continuous information dissemination to society. Therefore, it is essential to establish proper management techniques for the traceability data concerning the diverse characteristics of the waste, looking ahead to the disposal of various types of waste.

In recent years, distributed ledger technology, in particular blockchain, has attracted attention as a management method for sensitive data such as nuclear materials in the nuclear sector. Blockchain is a distributed management method that encrypts blocked data and has advantages such as ensuring data verifiability and transparency, data integrity (against tampering), and redundancy due to the distributed retention of information. Therefore, we have developed a Blockchain-based knowledge management system (commonly known as NEUChain-F) for secure and robust traceability data management of the various radioactive wastes originating from 1F.

The conditions for the Reactor Pressure Vessel (RPV) boundary failure and the behavior of the post-failure core material relocation to the pedestal in Fukushima Daiichi Nuclear Power Plant Units 2 and 3 were evaluated by integrating plant data, internal investigation results, and existing researches.

As a result, it is estimated that melting of Control Rod Drive (CRD) penetrations on the RPV wall occurred in Units 2 and 3, causing a leakage of core materials consisting mainly of molten metals. It is also suggested that local failure of the RPV side wall could have taken place at some time during this period. After this moment, cooling of remaining debris inside the RPV by injected water seems to have been established in Unit 2. In contrast to this, in Unit 3, in-vessel debris cooling is likely to have been insufficient allowing further debris/RPV heat up and resultant massive debris relocation to the pedestal.

This paper introduces the development of a 3-D information representation method to assist remote robot operators in the Fukushima Daichi Nuclear Power Station (FDNPS). This research aims to make remote robot operation easier and improve the safety and efficiency of FDNPS decommissioning. The proposal method is to overlay grid lines on the camera image. This method helps the operator's spatial awareness without increasing workload because its usage is the same as using camera images. The experiment evaluates the implementation of this method for ground robot moving operation, and its results show that this method improves operator distance awareness and reduces workload.

Comprehensive study to obtain necessary data for evaluation of airborne release behaviour during fuel debris retrieval of Fukushima Daiichi is underway by the authors with domestic and international collaborations since JFY2021 as subsidized projects of Decommissioning, Contaminated Water and Treated Water Management, titled “Development of Safety System (Acquisition of Dust Dispersion Data)”, funded by the Japanese Ministry of Economy, Trade and Industry.

As a huge number of experiments are necessary to obtain reliable data for five cutting methods against various operational parameters, this project was planned to proceed in several phases. Dust dispersion data for dry cutting with disk cutter, chisel, core-boring and laser were assessed during JFY2021 to JFY2022, while those for wet cutting with same methods and AWJ are under measurements since JFY2023. In addition, computational analysis to simulate airborne release and transport behaviour has been developed for future analysis of aerosols behaviour within PCV and reactor building.

The experimental setups for tests consisted of rectangular wind tunnel to cut samples under 0.1 m/s airflow to simulate inside PCV and samplers to collect particles to measure total weight and particle size distribution. Due to smaller aerosol generation than reliable detection limit, cutting tests with core- boring and chisel were carried out within small acryl box instead of wind tunnel. With a correction to same flow condition according to computational analysis, the measured data with different cutting methods were compared each other.

As the fuel debris simulants, non-radioactive materials were used to measure the dependency on cutting method and cutting parameters while materials containing uranium are served for the absolute measurements. The dust dispersion data for dry cutting with disk cutter, chisel, core-boring and laser have been measured for cold simulants while the data with simulated fuel debris containing uranium was measured for disk cutter.

A long-reach robotic system is developed for fuel debris retrieval operations at Fukushima Daiichi. The system has a hyper-redundant kinematic design resulting in high dexterity to access cluttered decommissioning environments inside the primary contaminated vessels. The long-reach system is in a slender design to allow for ease of accessing narrow ports and navigating through congestion. Similar to other long-reach systems for nuclear decommissioning operations, the long-reach system design results in significant flexibility, causing deformation and vibrations. The kinematic hyper-redundancy makes it challenging to plan motion sequences, taking into account space constraints and other safety-related considerations. Motivated to realise safe and efficient fuel debris retrieval, it is necessary to tackle essential technical gaps in the control and operation of the long-reach manipulator. In response, a series of research and development projects have been conducted to reduce risks in controlling the long-reach system and to provide operation teams with assistive features to plan operations, by leveraging state-of-the-art from other sectors. This addresses the key technological gaps paving the way towards retrieving fuel debris in engineering practice.

This study proposes an estimation method of radiation source distribution in high-dose environments using a mobile robot equipped with a non-directional detector and lightweight shielding, exploiting shielding attenuation. One challenge with radiation source distribution estimation using non-directional detectors is the decreased accuracy when measurement points are unevenly distributed. By leveraging the rotational motion of lightweight shielding and capitalizing on shielding attenuation, we propose a method to enable the non-directional detector to acquire directional information. The obtained radiation measurement results were used to explore the most likely radiation source distribution through maximum likelihood estimation. Through simulation experiments, the effectiveness of the proposed method was confirmed, demonstrating an improvement in the accuracy of radiation source distribution estimation.

The accident of the Fukushima Daiichi Nuclear Power Station has caused radioactive releases over Japanese lands and into the Pacific Ocean. In the days following the initiating events, a new form of radioactive Cs pollution has been collected and characterized. These particles, referred to as Cs-bearing microparticles (CsMP), are made of an amorphous silica matrix, in which many elements are imbedded, including radioactive Cs. These particles are a significant source of Cs pollution resulting from the accident and a fraction of them, called type-A CsMP has traveled over several hundreds of kilometers due to their micrometric size.

The generation mechanism of these particles has remained uncertain since their discovery. The inclusion of a significant concentration of radioactive Cs suggests a damaged core related interaction, and the silica matrix implies an Si-rich source involved at the origin of the particle generation. Therefore, among other possibilities, it has been suggested that CsMP could result from molten core concrete interaction (MCCI) after molten core relocation on the concrete pedestal.

To investigate this proposed mechanism, a small-scale experimental set-up has been designed to collect and study the aerosols generated during MCCI. Prototypic corium, containing depleted uranium oxide and stable fission product elements has been reacted with basaltic concrete at about 2000°C through induction heating, and the resulting aerosol have been collected. The first phase of the experiments, carried out in pure nitrogen atmosphere has produced micrometric spherical particles made of amorphous silica, highly similar morphologically to type-A CsMP. The chemical composition shows significant similarities, such as the matrix composition and the Si-relative concentration of Na, K, Rb, Al, Sn; but also some discrepancies like the low concentration of Fe and Cs and the high concentration of Ca. For the second phase, the protocol has been modified to inject a 20% steam - 80% nitrogen gas mixture in the test section and try to reach a more representative containment vessel atmosphere condition. These experiments have proven more challenging to carry out due to the reactive atmosphere, but micrometric particles have been also observed. While the overall composition was somehow similar, the Fe and Cs concentration appeared higher and Ca concentration lower.

Ion exchange technology has been used to decontaminate radioactive waste liquids from various sources of the nuclear industry for decades. Both organic and inorganic sorbents are in wide use in both operating nuclear power plants and also in plants under decommissioning. Fortum’s selective NURES® (NUclide REmoval System) ion exchange materials have been successfully used to treat waste liquids produced after the Fukushima accident in addition to the usual treatment of waste streams produced by normal operation of nuclear power plants and decommissioning liquids. In Fukushima ALPS system, two of the four ion exchange materials in NURES® product portfolio were used, CsTreat® and SrTreat®, to remove radiocesium and radiostrontium, respectively, from the waste liquids.

Cesium selective inorganic sorbent, CsTreat®, is based on ferrocyanide structure and provides high decontamination factors and capacity even in complex waste matrices for cesium removal. However, a concern over the potential release of cyanide from the ferrocyanide based materials has been raised especially in connection to the solidification of the used sorbent. As one of the challenges in the overall process of nuclear waste treatment is to secure a final stable state of the waste matrices, the disposal technique needs to be investigated thoroughly to assure a safe state of the waste matrix for decades. The used ion exchange materials are often solidified in various matrices and the chosen technology depends on the waste type, local regulations and power plant preferences. For NURES® materials, Fortum has developed a solidification technique called LOCKIT® which has been tested and optimized for CsTreat® solidification. To prove the final safe state of the LOCKIT® solidification product, the release of cyanide from CsTreat® needs to be assessed. In this work, the potential release of cyanide was investigated in leaching environment and in temperature gradient experiments. The tests aimed to simulate the conditions the solidified material would be exposed to in the final disposal environment. In addition, the temperature gradient tests investigated the potential release of cyanide in elevated temperatures which are relevant during the actual solidification process. The results presented provide information on the stability of the LOKIT® solidification process for ferrocyanide ion exchange material and the long term safety in regards to the potential release of cyanide.

Decommissioning of a nuclear installation generates radioactive liquid wastes that are more complex compared to the liquid wastes being handled in the operational phase. Different types of wastewater involved in the dismantling processes and decontamination processes challenge the capability and performance of the existing liquid waste treatment facilities. The ultimate goal of any radioactive liquid waste management is to immobilize the radionuclides contained in the liquid and ensure a safe and stable final state of the radioactive waste matrix. Processes such as evaporation, drying, precipitation and ion exchange are typically used and combined depending on the waste composition. The case-by- case approach is considered the best strategy where the process is optimized for the specific waste matrix. Volume reduction, i.e. to minimize the amount of wastes that ends up in a final repository is prioritized in the design and planning of the treatment system. A significant volume reduction helps mitigate the risks involved in the unknown final disposal costs in many countries.

In this work, we share Fortum’s experiences in managing radioactive liquid wastes in recent nuclear decommissioning projects. Mobile treatment systems incorporating an ultra- selective ion-exchange process is the most suitable candidate for purifying a diversely-sourced liquid waste. The maximum volume reduction originates from the fact that the radionuclides are selectively captured and subsequently conditioned for disposal. Here, we present a recent liquid waste treatment system delivered to a central Germany nuclear power plant with the aim of purifying 1300 m3 of high salt evaporator concentrate containing mostly Co-60, Cs-137, Cs-134 and Sb-124 radionuclides. Through a straightforward step- wise filtration process the radionuclides are intercepted by the selective ion-exchange columns, enabling a free-releasable output liquid after the treatment. In addition, we present recent developments in further treatment of decontaminated waste liquids to minimize e.g. boron releases. NURES® technology considers the entire radioactive waste treatment from liquid generation to final disposal, and by knowledge and experiences as an operator, ensures the safe waste treatment throughout the process.

One of the keys of Fukushima Daiichi Nuclear Power Plant (1F) decommissioning steps is the estimation of the Airborne Release Fraction (ARF) during the removal of nuclear Fuel Debris (FD), to evaluate the potential exposure dose for workers and the public and thus contribute to define the equipment and mitigation means for FD retrieval.

The Dust Acquisition project, led by CRIEPI in collaboration with ONET/CEA/IRSN, aims to evaluate the ARF , a key parameter reflecting the amount of radioactive material suspended in the air. Specificity of this project is to study disc cutting in air of prototypic FD-simulants containing depleted uranium oxide. ONET/CEA/IRSN has designed, installed, and qualified a dedicated facility named FUJISAN2 located on the PLINIUS Platform at CEA Cadarache, aligned with the specifications and cutting conditions set by CRIEPI, facilitating data comparison between Japan and France sides. This paper presents the whole process developed by ONET/CEA/IRSN for this study with the capacity to manufacture specific compositions and supply corium samples for cutting, to perform cutting tests on FUJISAN2 and on-line dedicated aerosols measurements as well as sampling aerosols, to conduct material analysis of initial FD simulants and aerosols.

In Japan, after the Fukushima Daiichi Nuclear Power Plant accident, many nuclear power plants should be considered operation period and life spans, including decommissioning. To realize safe and appropriate decommissioning of many nuclear plants, suitable management of radioactive materials produced in the reactors is one of the most important issues.

Classification of radioactive materials into clearance level and higher levels is also quite important as for the total cost of decommissioning.

Also, the accurate and cost-effective measurement for radioactive materials classification is highly desired. The standard source method has been applied to radioactivity measurements for the efficiency calibration of the measurement object. However, the objects to be measured in decommissioning are generally large and heavy. Therefore, this method requires shredding and molding. But these preparations are limited or cost prohibitive. Another method uses numerical calculations. For example, MCNP^TM is a general purpose Monte Carlo code utilized worldwide for these efficiency calculations. However, this code requires experience and knowledge of a trained user. Also, quite long calculation time should be expected. The In Situ Object Counting System (ISOCS^TM) technology was developed to solve this problem. This technology is already widely used around the world.

Prior to the Fukushima Daiichi NPP accident, ISOCS had only been applied to a few cases in Japan, such as the measurement of primary loop recirculation (PLR) piping in nuclear power plants. However, since the Fukushima NPP accident, the demand for on-site and off-site radioactivity measurement has rapidly increased, and the number of situations that make it difficult to create the efficiency calibration using a standard source has increased due to the high throughput and the large size of measurement targets. This is especially true for the measurement of removed soil and incinerated ash, off-site of Fukushima Diichi NPP. In addition, MCNP calculations are typically not adequately fast for real-time modeling. On the other hand, the time required for efficiency calibration using ISOCS is only a few seconds. At the Fukushima NPP off-site, ISOCS was also applied in the TRUCKSCAN and BULKSCAN systems. At the sites of other decommissioning projects, ISOCS has been applied to measurements of turbines, contaminated pipes, condensers, concrete wall of turbine buildings and more. In this report, we show the results of these various projects and the ISOCS benefits for cost, accuracy, and uncertainty.

The decontamination and decommissioning (D&D) of nuclear facilities and post-accident areas require the characterization and sorting of all materials based on their radioactive content prior to removal, reuse, or storage. Performing direct or in-situ measurements of large, bulky objects is a cost-effective strategy. It eliminates the need to disassemble, cut, crush, pre-treat, or re-package them into smaller items. These costly activities are also potential sources of additional exposure to workers through excessive contact and handling.

Calibrations based upon radioactive standards and sources are often time-consuming or difficult if the objects are large, have irregular shapes, or if the distribution of contamination is not homogeneous. This problem gets worse if the items are numerous or if these parameters change from item to item. Mirion Technologies’ In Situ Object Counting System (ISOCS) is a suite of software tools that can be used to generate efficiency calibrations based upon mathematical representations of the measurement geometry. These representations are created from models generated and benchmarked at the detector factory using Monte Carlo methods (MCNP) and NIST-traceable sources. ISOCS provides a large range of templates for most nuclear waste components in nuclear facilities, and it has an extensive history of benchmarking, validation, and regulatory acceptance in several countries.

Regardless of the calibration method, any deviation or mismatch in the distribution of the contamination compared to the calibration can introduce biases in the reported results. Recently, Mirion has developed an advanced type of general ISOCS template called SuperISOCS. With SuperISOCS, it has been demonstrated that complex scenarios and objects can be accurately represented and validated by adding up primitive template shapes. These have included entire glove boxes and its contents, stratified waste tanks, and thousands of truckloads containing sacks of soil of various sizes and fill heights.

This work presents the latest studies using SuperISOCS in D&D applications as a fast and effective tool to perform efficiency calibrations and uncertainty estimations of complicated geometries. The benefits of the ability to set whether any surface or sub-item has contamination will be demonstrated. Examples of comparisons between SuperISOCS and MCNP of conduit and pipes with arbitrary bends, and complex machinery, such as turbine assemblies, will be highlighted. Cost-benefit evaluations of these scenarios will provide some guidance to stakeholders and end-users on when approximations can be made using general templates and when SuperISOCS calibrations are recommended.

Thermochemistry prevailing between gaseous CsOH and concrete main chemical phase CaCO₃ at temperatures up to 570°C was investigated with various scenarios using the thermogravimetric method. The aim was to elucidate the decreasing behavior of cesium (Cs) trapping on CaCO₃ observed in the transpiration method. A quasi-two-compartment platinum crucible was developed to realize co-measurements of both CsOH and CaCO₃ during thermal treatment. Post-test X-ray diffraction was conducted to identify the chemical compound formed on the CaCO₃ precursor. The early presence (timely sensitivity) of CsOH near the heated surface of CaCO₃ was found to play a key role in the trapping (in the form of Cs₂CO₃). Such a factor is crucial because, otherwise, the Ca(OH)2 would predominate the surface upon CaCO₃ decomposition, leading to no reaction with CsOH.

Multi-particle model was developed to enable the prediction of dispersion behavior of non-spherical dust. Developed method was used to calculate the terminal velocity of various cuboid shapes. Shape factor of each cuboid shape was estimated from the calculated terminal velocity and conversion between the aerodynamic diameter and actual particle size was enabled. Numerical simulation of ARF measurement test was performed and the result was compared with experimental data to validate the developed method.