Decommissioning Of Nuclear Facilities Research Articles

Toward Robotic Nuclear Decommissioning: Deep Learning-Based Object Classification and Pose Estimation from Partial-View Scans

This study introduces a novel framework for the robotic decommissioning of nuclear facilities, that focuses on object classification and six degrees of freedom pose estimation from partial-view three-dimensional (3-D) scan data. Addressing the challenge of precise robotic manipulation in environments where acquiring full-scan data is impractical, this framework leverages a deep neural network for initial pose estimation, subsequently refined by a modified iterative closest point algorithm. Our method demonstrates high accuracy in identifying scanned objects and estimating their poses from partial-view scans, validated through experiments with 3-D printed mock-ups. This advancement highlights the potential for significantly enhancing robotic automation in nuclear decommissioning and related fields.

Numerical Simulation on Dispersion of Hydrogen Leaked in Particle Layers of Glass Beads and Soil

In order to understand the dispersion of hydrogen (H2) leaked in packed beds of nonporous/porous particles in a partially open space practically, the dispersion of H2 in the particle layers of glass beads and soil was analytically studied using a computational fluid dynamics code to be compared with the experiments and to elucidate the effects of the particle layer. The packed beds in the partially open space can be considered as a basic model for all processes of transfer, treatment, storage, and disposal of radioactive materials containing fuel debris in the decommissioning of nuclear facilities after a severe accident. The H2 flowing out from a single leak point in the particle layer of nonporous glass beads was affected by buoyancy around the leak point and diffused directly above the leak point in an elliptical shape faster than in the horizontal direction. After that, when it reached the air layer in the head space above the particle layer, the H2 spread horizontally, formed a large concentration gradient near the boundary between the particle layer and the air layer, and further diffused in the air layer until the H2 concentration became about 1/3 or less of the concentration near the surface of the particle layer. When the particle diameters were 512 μm and 1200 μm, this tendency was more pronounced when the porosity was the same and the particle diameter was smaller. The calculations largely reproduced the experimental concentration distributions. When the particle layer was porous decomposed granite soil, the diffusion behavior of H2 in the particle layer proceeded in the same manner as in the case of the glass beads. However, a large concentration gradient was formed near the boundary between the particle layer and the air layer, and then H2 diffused in the air layer until the H2 concentration became below the lower combustion limit. It was suggested through sensitivity analysis that the air permeability coefficient had a large effect on the time course of the H2 concentration distribution. Based on this information, we further simulated H2 behavior in the vessel containing the H2 leaked particle layer. By inserting multiple vent pipes without considering H2 generation distribution and particle properties in the particle layer, H2 accumulated from one pipe was discharged by buoyancy without depending on the H2 generation distribution and particle properties in the particle layer and air flowed in from the other pipe. It was suggested that such a natural ventilation process would reduce the H2 concentration in the container.

A comparative study of different radial basis function interpolation algorithms in the reconstruction and path planning of γ radiation fields

Accurate reconstruction of radiation field and path planning are very important for the safety of operators in the process of dismantling nuclear facilities. Based on radial basis function (RBF) interpolation algorithm, this paper discussed the application of inverse multiquadric radial basis Function (IMRBF) interpolation method to the reconstruction of gamma radiation field, and proved the feasibility of reconstructing a radiation field with multiple γ sources. The average relative errors of IMRBF interpolation results were 4.28% and 8.76%, respectively, for the experimental scenarios with single and double gamma sources. After comparing the consistency between the simulated scene and the experimental scene, IMRBF method and Cubic Spline method were respectively used to reconstruct the gamma radiation field by Geant4 simulation data. The results showed that the interpolation accuracy of IMRBF method was superior to that of Cubic Spline method. Further, more RBF interpolation algorithms were used to reconstruct the multi-γ source radiation field, and then the Probabilistic Roadmap (PRM) algorithm was used to optimize the human walking path in the radiation field reconstructed by different interpolation methods. The optimal paths in radiation fields generated by multiple interpolation methods were compared. The results herein contribute to a comprehensive understanding of RBF interpolation methods in reconstructing γ radiation fields and their application in optimizing paths in radiation environments. The insights may provide valuable information for decision-making in radiation protection during the decommissioning of nuclear facilities.

Remote measurement and calibration of alpha-emitting radionuclides through the utilization of characteristic ultraviolet fluorescence

During the decommissioning of nuclear facilities and spent fuel reprocessing, conventional alpha particle detectors are insufficient for accurately measuring the emissivity of alpha particles due to their high radiation intensity and complex nuclide types. The air fluorescence method is developed to detect the α particle emissivity through measuring the intensity of ultraviolet fluorescence generated by the interaction between α particles and nitrogen or other inert gas mediums. This method offers advantages such as long detection range and immunity to γ radiation interference. The measurement results are primarily influenced by factors including α particle energy, detection distance, and gas medium composition, with photon noise and detection distance determining the detection limit. The obtained results indicate that: In the condition of measured in darkness, measurement time for 1000s, and distance for 1 m, The minimum detection limit for α sources is 7.6 kBq with a relative standard uncertainty of 10%.

A brief investigation of the dose field virtual simulation tools for reactor decommissioning and preliminary design for the HWRR reactor in China

The calculation and visualization of the dose field in the decommissioning of nuclear facilities is one of the important functions of the decommissioning virtual simulation system. The dose field simulation tools can provide radiation field distribution and play an important role in determining the decommissioning plan and protecting personnel during the engineering implementation process. This article investigates the development of dose field calculation and visualization in the reactor decommissioning virtual simulation systems. A preliminary technology plan suitable for the development of the decommissioning dose field calculation and visualization display programs of the first Heavy Water Research Reactor (HWRR) in China is proposed. The applicability of the selected scheme is analyzed. The functional requirement and development direction of the HWRR reactor decommissioning dose field tool are preliminarily determined. Furthermore, the reactor vessel of HWRR reactor is modeled, the dose field distribution is calculated and visualized based on the preliminary decommissioning code. This research can provide technical support for the development of the decommissioning simulation system for the first HWRR reactor in China.

FE modeling of the reconstructed Bituminated radioactive waste storage of Ignalina NPP

Decommissioning of nuclear facilities is a complex of various activities strongly related to safety. In 2004 and 2005, both reactors of Ignalina NPP were closed. One of the tasks that must be solved is what to do with the radioactive waste storage facility (Ignalina AE Building 158). Possible solutions: – complete destruction of the radioactive waste storage with the need to empty the canyons and store the waste in a special place or, as an alternative, turning the storage into a repository. Converting storage to repository also has many different options. The installation of protective barriers of different thicknesses or even the construction of a sarcophagus is possible. The cheapest way to form a protective barrier on an existing building structure. The problem is the relatively high weight of the barrier structure, which must be supported by the existing building structure.The article presents an analysis of the durability of the structures of the radioactive waste storage, assessing the influence of the self-weight of the building's reinforced concrete structural elements and the mass of the bituminous radioactive waste stored in the canyons, as well as the effect of additional loads acting on the floor and roof of the first floor during the planned conversion of the storage facility into the repository. Numerical simulation was performed using ANSYS computer code and the results were compared with the relevant standards and norms in order to assess the building's ability to undergo such transformations.Spatial volumetric geometric and calculation models of the construction structures of the 158 building were created. These models have been modified after considering future changes during the transformation of the building into a repository. Perform calculations of the state of stresses and strains of the landfill under the load of the self-weight of the structural elements and the hydrostatic pressure of the bituminous waste. On the basis of the results of the calculations, conclusions were formulated about the level of load on the construction structures of the building and the margin under the current operating loads.The developed numerical models can be easily extended to evaluate the effects of new building elements and additional loads on structures (specifically walls and ceilings) after the transformation of a waste storage into a repository.The article presents not only the idea of reorganizing the waste storage structure to repository but also the methodology of modelling such a structure by assessing the age of the building's soil and additional loads. This methodology can be useful and applied to the closure of other nuclear power plants when it is necessary to turn the radioactive waste storage into a repository.

Influence of Gate Depletion Layer Width on Radiation Resistance of Silicon Carbide Junction Field-Effect Transistors

Silicon carbide junction field-effect transistors (SiC JFETs) are promising candidates as devices applicable to radiation conditions, such as the decommissioning of nuclear facilities or the space environment. We investigate the origin of the threshold volage (Vth) shift and hysteresis of differently structured SiC JFETs. A large positive Vth shift and hysteresis are observed for a depletion-type JFET with a larger depletion layer width. With changing the sweep range of the gate voltage and depletion width, the Vth shift was positively proportional to the difference between the channel depth and depletion width (channel depth–gate depletion width). By illuminating the sub-band gap light, the Vth of the irradiated depletion JFETs recovers close to nonirradiated ones, while a smaller shift and hysteresis are observed for the enhancement type with a narrower width. It can be interpreted that positive charges generated in a gate depletion layer cause a positive Vth shift. When they are swept out from the depletion layer and trapped in the channel, this gives rise to a further Vth shift and hysteresis in gamma-irradiated SiC JFETs.

Minimum dose path planning method for virtual nuclear facilities based on navigation mesh

The operation, maintenance and decommissioning of nuclear facilities have radiation risks. When workers work in the radiation area, it is necessary to carry out radiation protection through path planning. In order to optimize the dose field path planning algorithm, a virtual verification platform for dose field path planning is established based on the Unreal Engine. The platform includes virtual environment, dose field, and personnel dose assessment modules, which can meet the needs of dose field path planning algorithm research and optimization. Considering the complexity and variability of the environment in the operation process, a new dose minimum path planning algorithm based on navigation mesh is designed. Based on this algorithm, the number of grids for path planning in complex or large scenes is significantly reduced, and the calculation time of the algorithm is reduced while ensuring the optimal dose. The simulation results show that the minimum dose path planning method based on navigation mesh can effectively reduce path planning time and ensure the safety of personnel during the decommissioning of nuclear facilities.

Principles of Multivariate Analysis in the Reltran Program for the Problems of Estimating Public Exposure Doses during the Decommissioning of Nuclear Facilities

The article represents the approaches introduced in the RELTRAN code developed under the FTP NRS-2 to modeling the radiation consequences for the population caused by possible releases of radioactive substances into the atmosphere during the planned work on the decommissioning of nuclear facilities. One of the tasks of the RELTRAN code is to perform calculations of the consequences of radioactive releases for the population for the tasks of justifying radiation safety. Several modeling options are considered: calculations based on average concentrations over several years, calculations based on average concentrations over several years, considering the schedule of the decommissioning process, case-study modeling approach based on historical meteorological series analysis and case-study modeling approach based on historical meteorological series analysis, considering the schedule of the decommissioning process. It was obtained that, depending on the duration of the planned process and depending on the season, the results differ significantly, for medium-term work, their seasonality becomes an important factor, at the same time, work with conditionally constant emissions lasting more than 6 months has a negligible uncertainty associated with time start of work.

The effect of sensitive and non-sensitive parameters on DCGL in probability analysis for decommissioning of nuclear facilities

In the decommissioning of nuclear facilities, Derived Concentration Guideline Level (DCGL) derivation is necessary for the release of the facility after the site remediation, which also needs to be implemented in the stage of establishing a decommissioning planning. In order to derive DCGL, the dose assessment for the receptors can be conducted from residual radioactivity by using RESRAD code. When performing sensitivity analysis on probabilistic parameters, secondary evaluation is performed by assigning a single value for parameters classified as sensitive. However, several options may arise in the handling of non-sensitive parameters. Therefore, we compared the results of the first execution of RESRAD applying probabilistic parameters for each scenario with the results of the second execution applying a single value to sensitive parameters among the probabilistic parameters. In addition, we analyzed the effect of setting options for non-sensitive parameters. As a result, the effect on DCGL were different depending on the application scenario, the target radionuclides, and the input parameter selections. In terms of the overall evaluation period, the DCGL graph of the default option was generally shown as the most conservative except for some radionuclides. However, it will not necessarily be given priority in the aspect of the need to reflect site characteristics. The reason for selecting a probabilistic parameter is the availability of the parameter and the uncertainty of applying a single value. Therefore, as an alternative, it can be consistently applied to distribution as an option for non-sensitive parameters after sensitivity analysis.

Numerical simulation and experiments study on laser ablation decontamination for strontium and cesium from contaminated 316L stainless steels in spent nuclear fuel reprocessing

Radionuclides, 90Sr and 137Cs in this study, generated during spent fuel reprocessing will contaminate the structural material (316 L stainless steels), affecting the safe decommissioning of nuclear facilities. Laser ablation is a promising surface decontamination technology. In this study, the optimal laser parameters (the laser power of 16 W, the scanning speed of 175 mm/s, and the pitch spacing of 10 μm), used for laser ablation decontamination experiments for contaminated 316 L stainless steel in NaOH and HNO3 containing Sr and Cs which simulated the environments of spent nuclear fuel storage pond and Plutonium Uranium Redox Extraction (PUREX) process, were determined by numerical simulation with COMSOL Multiphysics. Laser induced breakdown spectroscopy (LIBS) was used to evaluate the decontamination performance of laser ablation, and no contaminant elements (Sr and Cs) were detected on the steel surface after laser ablation. Moreover, physicochemical technologies (SEM-EDS, XRD, and XPS) were used to characterize the surface chemical states and morphologies of contaminated steel before and after laser ablation. It was found that on the surface of the contaminated steel after laser ablation, δ-ferrite phase was detected, and the passive film was re-formed.

SHARE: Stakeholder based analysis of research for decommissioning

The H2020 EU-funded SHARE project (Stakeholders- based Analysis of REsearch for Decommissioning) is a forerunner to the establishment of a framework for collaboration on research activities related to the decommissioning of nuclear facilities. SHARE aimed to provide an inclusive roadmap for decommissioning research, in both technical and non-technical areas, in the EU and abroad, to enable stakeholders to improve jointly safety, reduce costs and minimise environmental impact. SHARE has been built on a consultation process considering the needs and perspectives of different stakeholders all across the decommissioning value chain. The project also considered existing and emerging innovative technologies solutions as well as the international best practices in the field of decommissioning. After a three-year process, the project provides a Strategic Research Agenda and a Roadmap built on the participation of the international Stakeholder community in a multi-step process including a questionnaire survey, a state-of-the-art review, a gap analysis and multiple workshops. As the final output, the SHARE roadmap effectively set the framework for organizing the priorities identified in the SHARE SRA.

Application of robotics during the decommissioning of nuclear facilities: state of the art and development prospects

The article deals with the application of mobile robotic systems (UGV) during the decommissioning of nuclear industry and energy facilities. It is shown that, in contrast to the recovery from accidents, when usually the work is carried out in an unknown unstructured environment, which implies damage to buildings, blockages, flood zones, etc., decommissioning is carried out in a rather complex, but well known environment according to a pre-developed plan. The general requirements for such UGVs are discussed, based on the features and practice of their application.

Optimal Decision Support System under the Decommissioning of Nuclear Facilities

Optimal Decision Support System under the Decommissioning of Nuclear Facilities

Effects of Vent Size and Wind on Dispersion of Hydrogen Leaked in a Partially Open Space: Studies by Numerical Analysis

To understand the dispersion of hydrogen (H2) leaked in a partially open space practically, which can be considered as a basic model for all processes of transfer, treatment, storage, and disposal of radioactive materials containing fuel debris in the decommissioning of nuclear facilities after a severe accident, this paper uses a computational fluid dynamics code to study analytically the effects of vent size and outer wind on H2 dispersion. The paper adopts the experimental Hallway model, which has a H2 release hole on the ceiling, one vent on the roof (Roof vent), and one vent on the side (Door vent). Air flows in the model (room) from the Door vent while H2 is discharged outside from the Roof vent. The discharged (outflow) amount of H2 increases in conjunction with the air inflow when the size of the Roof and/or Door vents is increased, and then vice versa. The effect of wind depends on the direction to the Door vent: Wind from the same direction as the Door vent promotes H2 discharge while wind from the opposite direction suppresses it. The dispersion behavior characteristics of indoor leaked H2 are clarified for comparing model tests with the same Froude number and different scales. It is found from the analysis results of comparing model tests with the same Froude number and different scales that when H2 leaks into the room and diffuses to the air, the flow generated by the buoyancy of mixed gas creates the stack effect, which causes natural ventilation by drawing in air from the outside through the vent. In addition, it is speculated that the H2 concentration decreases after its leak by quickly mixing with air that flows in from the vents and reaches the floor due to the Coanda effect, which is the effect of the free jet being drawn to a nearby wall.

Cyto-Genotoxicity of Tritiated Stainless Steel and Cement Particles in Human Lung Cell Models.

During the decommissioning of nuclear facilities, the tritiated materials must be removed. These operations generate tritiated steel and cement particles that could be accidentally inhaled by workers. Thus, the consequences of human exposure by inhalation to these particles in terms of radiotoxicology were investigated. Their cyto-genotoxicity was studied using two human lung models: the BEAS-2B cell line and the 3D MucilAirTM model. Exposures of the BEAS-2B cell line to particles (2 and 24 h) did not induce significant cytotoxicity. Nevertheless, DNA damage occurred upon exposure to tritiated and non-tritiated particles, as observed by alkaline comet assay. Tritiated particles only induced cytostasis; however, both induced a significant increase in centromere negative micronuclei. Particles were also assessed for their effects on epithelial integrity and metabolic activity using the MucilAirTM model in a 14-day kinetic mode. No effect was noted. Tritium transfer through the epithelium was observed without intracellular accumulation. Overall, tritiated and non-tritiated stainless steel and cement particles were associated with moderate toxicity. However, these particles induce DNA lesions and chromosome breakage to which tritium seems to contribute. These data should help in a better management of the risk related to the inhalation of these types of particles.

Clearance measurement for concrete waste generated by the decommissioning of uranium processing facilities

Decommissioning is planned at nuclear facilities that have been discontinued. We examined the evaluation method of uranium radioactivity for concrete waste generated by the decommissioning of nuclear facilities. Since the peaks of Ac-228, Tl-208, and K- 40 are derived from concrete waste, it is difficult to distinguish the 1001 keV peak emitted from the uranium source. We have derived a formula to correct gamma rays from concrete and the environment, and the amount of uranium was quantified. When the weight of concrete waste is about 300 kg, if the weight of uranium is 3 g or more, it can be quantified within a relative error of about 30%. Measurement tests were performed using homogeneous simulated concrete waste. Since uranium contamination is on the concrete surface at the uranium processing facility and small chunks generated by scraping the concrete surface will be stored in a drum and measured, it seems that the test of homogeneous concrete reflects the actual waste.

Перспективы использования ксенонового гаммаспектрометра в качестве регистрирующего устройства на борту автоматизированного и роботизированного гамма-спектрометрического комплекса для вывода из эксплуатации ядерно-физических установок

The paper considers the structure of robotic gamma-spectrometric complexes for the decommissioning of nuclear facilities. It presents physical and technical characteristics of xenon gamma-ray spectrometer considered a key instrument for gamma-ray spectra measurements of analyzed objects and their activity. The study demonstrated the feasibility of xenon gamma spectrometers application in the decommissioning of nuclear facilities. Remotely received information from a gamma spectrometric complex is transmitted via Internet to a remote computer enabling realtime processing of the experimental data.

Investigation of the mechanism for simulated graphite waste treatment via microwave sintering technology

With the decommissioning of nuclear facilities, an increasing amount of radioactive graphite needs to be treated safely and effectively. In this work, microwave sintering technology was initially applied to treat simulated graphite waste. The effects of microwaves on the phase composition and microstructure of graphite were investigated. Combined with the mass loss results and fitting results of the phase, the mechanism for simulated graphite waste treatment via microwave sintering was obtained. According to the results, there are three stages regarding the effect of microwave sintering on the microstructure of graphite waste. This work reveals the potential application of microwave sintering technology in the treatment of graphite waste and its treatment mechanism.

Characteristic and remediation of radioactive soil in nuclear facility sites: a critical review.

A huge amount of radioactive soil has been generated through decommissioning of nuclear facilities around the world. This review focuses on the difficulties and complexities associated with the remediation of radioactive soils at the site level; therefore, laboratory studies were excluded from this review. The problems faced while remediating radioactive soils using techniques based on strategies such as dry separation, soil washing, flotation separation, thermal desorption, electrokinetic remediation, and phytoremediation are discussed, along with appropriate examples. Various factors such as soil type, particle size, the fraction of fine particles, and radionuclide characteristics that strongly influence radioactive soil decontamination processes are highlighted. In this review, we also survey and compare the pool of available technologies currently being used for the remediation of radionuclide-contaminated soils, as well as the economic aspects of soil remediation using different techniques. This review demonstrates the importance of the integrated role of various factors in determining the effectiveness of the radioactive soil decontamination process.