This paper investigates the potential of visible light communication (VLC) for autonomous railway vehicles (ARVs) engaged in nuclear waste disposal. The research focuses on designing an ARV system equipped with VLC technology to ensure safe, reliable communication in hazardous environments, such as deep geological repositories. Initial testing demonstrated that VLC can effectively maintain communication between ARV components over distances of up to 30 m, even when operating at limiting angles. In this study, angles of 60° and 75° were tested, corresponding to the angles of the curves of the tested route. With 4-QAM modulation it was possible to measure at a distance of approximately 5.5 m with an angle of 75°. The system has shown promise in addressing key challenges, such as high radiation levels and confined spaces, where traditional RF-based communication systems may fail. The results indicate that VLC could significantly improve the safety and efficiency of ARV operations in nuclear waste disposal, offering advantages in terms of robustness and reliability. Future work will focus on integrating VLC more deeply into ARV control systems, further testing in real-world environments, and exploring its application nuclear waste management.
Recently, 3D radiation field reconstruction using data collected by robot has been examined to reduce radiation exposure to workers inside and outside nuclear facilities. Many studies have proposed radiation field reconstruction methods that consider radiation diffusion characteristics; however, they often face challenges in accounting for the effects of the surrounding environment. We proposed a novel kernel function of Gaussian process regression that considers both radiation characteristics and attenuation effects. Our kernel function incorporates diffusion properties and attenuation effects caused by obstacles. We extract geometry information from point cloud data and applied it to model the attenuation effects when building radiation map. To validate the novel kernel function, radiation data are collected from simulation and real-world datasets, then compared with radiation field estimation methods that do not consider attenuation effects. The experimental results show that in the attenuation regions, our proposed method achieved root mean square error reductions of 78.24 %, 43.65 %, and 69.26 % for the barrel environment, reactor environment, and real-world dataset, respectively, compared to the previous method. The proposed strategy demonstrated higher accuracy in predicting radiation maps by better reflecting the physical characteristics of real world. This capability enables rapid responses during incidents and helps minimize potential damage.
NI Weixuan, JIA Lixia, DOU Yankun, HE Xinfu, CAO Jinli, WANG Dongjie, YANG Wen
Molybdenum (Mo) alloys are widely employed in high-temperature environments, such as advanced nuclear systems, due to their excellent properties, including high melting point and thermal conductivity. However, Mo alloys suffer from poor ductility at room and intermediate temperatures, which can be improved by adding rhenium (Re) or dispersing second-phase particles. These particles refine the grain size by promoting nucleation and inhibiting grain growth, improving strength while reducing the concentration of harmful solutes. Nevertheless, creep behavior is a critical performance aspect for Mo-Re alloys during service, as it can limit their high-temperature applications. Creep refers to the time-dependent plastic deformation that occurs at high-temperatures under stress levels below the yield strength of a material. In polycrystalline Mo alloys, grain boundary sliding exacerbates creep at elevated temperatures, restricting their use. Previous studies on Mo creep behavior have indicated that subgrain formation and grain growth occur during high-temperature service, and nanocrystalline materials exhibit distinct creep mechanisms compared to polycrystalline counterparts. Recent findings suggest that nanocrystalline materials can experience significant creep even at lower temperatures, indicating the importance of investigating the effects of high grain boundary density on the creep behavior of Mo alloys. Given that experimental creep tests require long durations, molecular dynamics (MD) simulation offers an efficient alternative for studying atomic-scale processes at grain boundaries. In this study, MD simulation was employed to investigate the tensile creep behavior of nanocrystalline Mo with varying grain sizes under different temperature and stress conditions. The Voronoi method was used to generate nanocrystalline Mo structures with random grain orientations, and tensile creep simulations were conducted using the LAMMPS software. The results reveal that increasing temperature and applied stress accelerates the creep process, with smaller grain sizes exhibiting more pronounced creep behavior. Atomic-level visualization shows that dislocation density and grain structures remain largely unchanged, while local atomic environments change due to vacancy diffusion along grain boundaries. These changes are responsible for the observed deformation mechanisms, particularly Coble creep, which dominates under the simulated conditions. This study provides valuable insights into the mechanisms of creep in nanocrystalline Mo at 800-1 400 K, which is critical for its potential application in nuclear industry designs. The findings highlight the importance of understanding grain boundary diffusion and its role in controlling the overall creep behavior in nanocrystalline materials.
In nuclear reactors, the fuel cladding is exposed to high temperature and high pressure for a long period of time. Chalk River Unidentified Deposits (CRUD) will form on the surface of the fuel cladding during the conventional operation of pressurized water reactors, and the formation of the CRUD will affect the flow heat transfer on the surface of the fuel rods. In order to investigate the effect of surface fouling on the flow characteristics of the CRUD, as well as to explore the influence of cladding surface deposition layer on the active nucleation site density (NSD), the present study was based on the flow-boiling under atmospheric pressure. The flow boiling visualization experiment was conducted to simulate the actual fuel rod cladding with CRUD by using layer-by-layer deposition of SiO2, under two mass flow rates (0.12 m/s and 0.17 m/s) and three degrees of subcooling conditions (0, 3, 5 K), investigating the flow boiling heat transfer characteristics of fuel cladding Zr-4 non-deposition with two Zr-4 SiO2 depositions (1 μm and 3 μm). Focus on the relationship between the active nucleation site density Na with wall superheat, and analyze the main reasons for the differences under different operating conditions, and contrast differences in the active nucleation site density on the different of SiO2 deposited thicknesses, then compare it with existing models for the active nucleation site density. The results show that the deposited Zr-4 has a higher flow heat transfer capacity than the undeposited Zr-4, and this difference is mainly caused to the difference in surface porosity. The active nucleation site density increases on SiO2 deposited surfaces compared to undeposited surfaces, with a maximum in the 3 μm SiO2 deposited experimental group. Enhancing wall superheat increases the active nucleation site density, and the increase is more pronounced on surfaces with SiO2 deposits. For the same sample, under the condition that one of the degree of subcooling and flow rate is the same and the other is different, the difference in flow rate has a greater effect on the active nucleation site density than the degree of subcooling, and the increase in both flow rate and the degree of subcooling decreases the active nucleation site density. Based on the experimental data, nine prediction models were used for calculation and analysis, and the Končar model can better predict the active nucleation site density under the working conditions of this experiment.
Highly radioactive materials classified as high-level nuclear waste (HLW) of atomic power engineering should be disposed of deeply underground in special geological disposal facilities (GDFs), which can be of either shaft or borehole type. The advantages of borehole-type GDFs result from smaller volumes of mining operations, a simpler construction technology, shorter construction time and cost. This allows us to consider them as an alternative to shaft-type GDFs. The parts of the boreholes in which waste containers should be placed can be both vertical and horizontal. Computer simulation of the migration of radionuclides from a group of parallel horizontal boreholes into the biosphere made it possible to conclude that horizontal GDF boreholes have significant advantages over vertical ones. We determined a forecast of 241Am migration by a method of mathematical modelling of 241Am release from vitrified HLW disposed of in several horizontal drillholes. The maximum concentrations of americium in the near-surface groundwater above the repository are calculated depending on the number of boreholes, the depth of their location and the distance between them, the permeability of rocks and the time of waste storage prior to disposal. Influence of different conditions on the safety of a GDF of borehole type is estimated. Calculations show that the heat generated by HLW causes a weaker groundwater convection near horizontal boreholes compared to vertical boreholes of the same capacity. In addition to that, at an equal thickness of the rock layer separating the HLW from the surface, the geothermal temperature of the host rocks in the near field of a horizontal borehole will be lower than the average geothermal temperature near a vertical borehole. As a result, the rate of radionuclides leaching from the waste forms by groundwaters will also be lower in the case of horizontal boreholes.
CAI Boshuai;YU Tianjun;LIN Xuan;ZHANG Jilong;WANG Zhixuan;YUAN Cenxi
The random forest algorithm was applied to study the nuclear binding energy and charge radius. The regularized root-mean-square of error (RMSE) was proposed to avoid overfitting during the training of random forest. RMSE for nuclides with Z, N>7 is reduced to 0.816 MeV and 0.020 0 fm compared with the six-term liquid drop model and a three-term nuclear charge radius formula, respectively. Specific interest is in the possible (sub) shells among the superheavy region, which is important for searching for new elements and the island of stability. The significance of shell features estimated by the so-called shapely additive explanation method suggests (Z, N)=(92, 142) and (98, 156) as possible subshells indicated by the binding energy. Because the present observed data is far from the N=184 shell, which is suggested by mean-field investigations, its shell effect is not predicted based on present training. The significance analysis of the nuclear charge radius suggests Z=92 and N=136 as possible subshells. The effect is verified by the shell-corrected nuclear charge radius model.
Plasma–facing components in future fusion reactors must endure high temperatures as well as high fluxes and fluences of high energy particles. Currently tungsten has been chosen as the primary plasma-facing material due to its good thermal conductivity, low erosion rate and low fuel retention. Materials with even better properties are still being investigated to be used in reactor regions with demanding plasma conditions. High entropy alloys (HEA) are a new class of metallic alloys and their exploitation in fusion applications has not been widely studied. In this work, the hydrogen isotope exchange effect in an equiatomic HEA containing W, Mo, Ta, Nb, and V was studied. Deuterium was implanted into HEA samples with 30 keV/D energy and the HEA and reference samples were annealed in H2 atmosphere and in vacuum at various temperatures up to 400 °C, respectively. The near-surface D concentration profiles were measured with ERDA and the isotope exchange was observed to remove over 90 % of the trapped deuterium from the implantation region at temperatures above 200 °C. TDS was used to measure retention deeper in the bulk in which the reduction of trapped deuterium was significantly lower. High total retention of H was found in the bulk after H2 atmosphere annealing which indicates permeation and deep trapping of H in the material.
The multi-physics coupled methodologies that have been widely used to analyze the complex process occurring in nuclear reactors have also been used to the R&D of numerical reactors. The advancement in the field of computer technology has helped in the development of these methodologies. Herein, we report the integration of ADPRES code and RELAP5 code into the SALOME-ICoCo framework to form a multi-physics coupling platform. The platform exploits the supervisor architecture, serial mode, mesh one-to-one correspondence and explicit coupling methods during analysis, and the uncertainty analysis tool URANIE was used. The correctness of the platform was verified through the NEACRP-L-335 benchmark. The results obtained were in accordance with the reference values. The platform could be used to accurately determine the power peak. In addition, design margins could be gained post uncertainty analysis. The initial power, inlet coolant temperature and the mass flow of assembly property significantly influence reactor safety during the rod ejections accident (REA).
The paper presents the results of modeling of changes in the isotopic composition of the liquid-salt fuel circulating in the experimental channel of the MBIR reactor facility. The authors tested the ISTAR software environment adapted for solving burnup equations in problems with variable power levels. The loop channel parameters, including two heat exchanger options, were estimated to obtain the appropriate salt transit time through the loop channel zones. Two problems of a circulating fuel system (loop) modeling are considered, namely: (1) modeling the equilibrium salt isotope composition in such a system; and (2) developing a technique for modeling nonstationary isotope kinetics in the MBIR reactor loop. Non-stationary isotope kinetics can be modeled as sequential burnup of nuclides in the neutron field and decay during movement in the external circuit. The authors also developed an algorithm for modeling changes in the isotopic composition of fuel salt during its circulation, taking into account the sequential transfer of a given salt volume from the burnup zone to the zone outside the reactor core. Based on this algorithm, a software package was created using the Python 3.9 programming language and ISTAR modules. In addition, a description of the calculation methodology was given and some calculation results obtained using the software were presented. In the process of working with the program, it was found that, for the given times of the fuel being in each of the zones (2 and 200 seconds, respectively), modeling the change in the isotopic composition during the fuel campaign (500 days) will require the calculation of more than 500 thousand steps. In order to save time, it is necessary to find out whether it will be possible to reduce the number of calls to the neutronic calculation code due to a slight change in the isotopic composition of the fuel in the loop per one burnup step. Work is currently underway to optimize this process.
A.A. Pshenov, A.S. Kukushkin, A.V. Gorbunov
et al.
In high recycling and detached plasma regimes the recycling region largely defines the properties of the whole scrape-off layer plasma by setting up the density and temperature profiles. As detachment approaches, the plasma temperature in the vicinity of the divertor targets drops to Te ∼ 1 eV and density increases to ne ∼ 1020–21 m−3, and the recycling region becomes increasingly opaque to the hydrogen radiation. Hydrogen radiation trapping influences both power and particle balance, both directly – reducing hydrogen radiation loss and changing plasma ionization/recombination balance inside the recycling region, and indirectly – affecting impurity radiation, charge-exchange and other important processes through the changes in the edge plasma density and temperature profiles.This paper reviews three-decade-long experimental and simulation efforts of quantifying the effects of hydrogen radiation trapping on the tokamak divertor plasma. It is demonstrated that opacity has a strong impact on the local divertor plasma parameters, changing both the electron and atomic hydrogen densities by a factor of several compared to the transparent plasma limit. Moreover, it is shown that allowing for opacity shifts the operational window of tokamak divertor towards higher separatrix densities and higher impurity content in order to maintain the desired level of power dissipation within the divertor. However, the most important is the effect of the opacity on the spectroscopic diagnostics. It is shown that despite the divertor plasma is virtually always transparent to the Balmer series emission, the resulting signals are affected by the changes in the populations of the excited states of atomic hydrogen, associated with the Lyman series absorption. As a result, the intensity of individual Balmer series signals can increase by a factor of 4 – 8 and the ratios of the Balmer line signals change by a factor of 2. This result implies that reliable quantitative measurements of plasma parameters with spectroscopic diagnostics at high power, high density tokamaks require opacity strength measurements and corresponding corrections to the rate coefficients used for the analysis of the spectroscopic data.
Abstract One of the main priorities in the modern thermal power engineering development is the problem of energy saving due to the economical use of fuel and energy reserves. Increasing energy consumption with a simultaneous increase in energy prices and widespread environmental degradation necessitates the development and implementation of energy efficient technologies to save fuel, materials and labour costs. The object of study is tubular heat exchangers of variable cross section, which are widely used in steam generators of nuclear power plants, gas turbines and transport plants. Scale deposit properties and the composition of the heat coolant were studied, and their influence on the energy efficiency of heat exchangers was analysed. To study the scale deposits and coolant influence on the energy efficiency of heat exchangers, their properties were examined using an atomic emission analysis with the help of a TESCAN electron microscope. The principles of implementing technologies aimed at intensifying heat transfer, reducing hydraulic and heat losses in heat exchangers were formulated.
Anthropogenic radionuclides of plutonium and 137Cs in the environment were usually applied for tracing the environmental impact of nuclear activities as well as establishing the chronology of recent lake sediment. In this study, the vertical distribution and spatial differentiation of plutonium and 137Cs in lake sediments of China were systematically contrastive analyzed to elucidate the regional influence by the local fallout from Chinese nuclear tests. The results demonstrated that the inventory of 137Cs and 239,240Pu, and their isotopic composition in lake sediments in northwestern China were obviously different from other regions, indicating that a portion of radionuclides in lakes in northwest China were originated from the Chinese Lop Nor Nuclear Test site (CNTs), and radionuclides in lakes of other regions may be completely originated from global fallout. Usually, the anthropogenic radionuclides recorded in lake sediments showed a single peak corresponding to the maximum deposition of 1964. However, two or three accumulation peaks of radionuclides were also observed in a few lakes. For the application of radionuclides on the dating of recent sediment, the accumulative peak of 239,240Pu and 137Cs signed to the period of 1963–1964 is the only practicable time marker for lake sediments in other regions of China. Isotopic information, such as 240Pu/239Pu ratio and 239,240Pu/137Cs activity ratio, may be useful for accurate sediment chronology, as well as used as a fingerprint to identify the source of the associated radioactive substance in the environment.
An approximate method for calculating the cadmium ratio of a cyclotron-based thermal neutron radiography system was developed. In this method, the Monte-Carlo code, MCNP6.2, was employed to calculate the neutron capture rates of Au-197, and the cadmium ratio was obtained by computing the ratio of neutron capture rates. From the simulation results, the computed cadmium ratio is reasonably acceptable, and the assumption of ignoring the fast neutron contribution to the cadmium ratio is valid.
Abstract As a part of probabilistic risk (or safety) assessment (PRA or PSA) of nuclear power plants (NPPs), the primary role of human reliability analysis (HRA) is to provide credible estimations of the human error probabilities (HEPs) of safety-critical tasks. In this regard, it is vital to provide credible HEPs based on firm technical underpinnings including (but not limited to): (1) how to collect HRA data from available sources of information, and (2) how to inform HRA practitioners with the collected HRA data. Because of these necessities, the U.S. Nuclear Regulatory Commission and the Korea Atomic Energy Research Institute independently developed two dedicated HRA data collection systems, SACADA (Scenario Authoring, Characterization, And Debriefing Application) and HuREX (Human Reliability data EXtraction), respectively. These systems provide unique frameworks that can be used to secure HRA data from full-scope training simulators of NPPs (i.e., simulator data). In order to investigate the applicability of these two systems, two papers have been prepared with distinct purposes. The first paper, entitled “SACADA and HuREX: Part 1. The Use of SACADA and HuREX Systems to Collect Human Reliability Data”, deals with technical issues pertaining to the collection of HRA data. This second paper explains how the two systems are able to inform HRA practitioners. To this end, the process of estimating HEPs is demonstrated based on feed-and-bleed operations using HRA data from the two systems.
Michael Hamel-Green, Peter Hayes, Yongsoo Hwang
et al.
In October–November 2020, RECNA-Nagasaki University and Asia-Pacific Leadership Network (APLN) with the support of Nautilus Institute convened the Nagasaki 75th Anniversary scenario workshop to address the pandemic-nuclear nexus. About fifty participants of diverse background explored the future of nuclear war and nuclear disarmament in light of the uncertainty created by the global coronavirus pandemic. Their specific task was to develop four scenarios to identify the opportunities driven by global pandemics for Northeast Asian governments, civil society and market actors to reduce nuclear risk and resume nuclear disarmament. Based on these scenarios, the workshop recommended sixteen urgent steps that could be implemented globally and in particular, in Northeast Asia. With three more policy measures added after the workshop, this article elaborates nineteen policy recommendations.
Nuclear engineering. Atomic power, International relations
Michael Duerrschnabel, Steffen Antusch, Birger Holtermann
et al.
A detailed microstructural analysis is one key factor for establishing structure–property relationships, which themselves are essential for manufacturing any device or part thereof. In particular, this paper focuses on the microstructural analysis of tungsten composite materials produced by powder injection molding (PIM). Our combined scanning electron microscopy (SEM) and transmission electron microscopy (TEM) approach revealed that W/TiC and W/Y2O3 composites are promising candidates for e.g. plasma facing components in future fusion reactors. The grains size distribution of all present phases was a log-normal one. TiC and Y2O3 precipitates in contrast to HfC ones limited the grain growth of the tungsten matrix during sintering about three times more efficient. The precipitate grain size was for all samples in the range of 1.7 µm–3.5 µm. Chemical interaction was only observed for TiC-based composites in the form of W diffusion into the TiC precipitate forming a mixed (Ti, W) carbide retaining the face-centered cubic (fcc) based crystal structure of pure TiC. The tungsten content in Y2O3 and HfC precipitates was found to be negligible. La2O3 was only observed in TEM attached to (Ti, W)C particles in the form of about 100 nm sized precipitates. As result, the Y2O3 and TiC containing samples are considered as promising materials for further detailed mechanical and microstructural investigations.
Hans-Josef Allelein, Ernst-Arndt Reinecke, Stephan Kelm
et al.
In Germany, the Helmholtz Association’s program NUSAFE (Nuclear Waste Management, Safety and Radiation Research) tackles the scientific challenges for the safety of nuclear reactors in the key areas of design basis (DBA) and beyond design basis accidents (BDBA). As one of the three research centers in NUSAFE, the R&D activities at Forschungszentrum Jülich (FZJ) focus on containment phenomena and processes during severe accidents. FZJ supports international efforts to further develop computational methods and tools by providing worldwide unique facilities in the field of wall condensation, aerosol behavior and hydrogen mitigation for advanced code development and validation. Development of computational fluid dynamics (CFD) tools for containment application aims at the integration of existing knowledge and the transfer of research results into application. All research activities are integrated into international collaborations and projects.
Chenchong Wang, Chunguang Shen, Xiaojie Huo
et al.
In order to make reasonable design for the improvement of comprehensive mechanical properties of RAFM steels, the design system with both machine learning and high-throughput optimization algorithm was established. As the basis of the design system, a dataset of RAFM steels was compiled from previous literatures. Then, feature engineering guided random forests regressors were trained by the dataset and NSGA II algorithm were used for the selection of the optimal solutions from the large-scale solution set with nine composition features and two treatment processing features. The selected optimal solutions by this design system showed prospective mechanical properties, which was also consistent with the physical metallurgy theory. This efficiency design mode could give the enlightenment for the design of other metal structural materials with the requirement of multi-properties. Keywords: Machine learning, High-throughput optimization, Mechanical property, RAFM steel