Earticle

Radioactive waste generation tends to increase as the nuclear energy industry develops, and controlling the radioactivity’s entry into the warehouse is becoming increasingly challenging. The informational framework for managing radioactive waste increases data entry accuracy, lowers data management error rates, ensures data traceability about production, treatment, storage, transportation, disposal, and other processes, and lowers radiation exposure levels for staff. This literature review collected and studied the relative research reports and patents on radioactive waste management from China National Knowledge Infrastructure, Web of Science, Elsevier, SpringerLink, and other databases were collected and studied, especially the application cases of information management and knowledge management to promote the innovation and development of knowledge management in the whole radioactive waste management process. Managers intuitively ascertain the radioactivity degree of waste, the location of various waste management sections, and the state of radioactive waste management using information management systems. However, the conventional information management system is unsuitable and no longer meets the needs of managing radioactive waste because information on radioactive waste is readily lost and cannot be shared. The authors designed a radioactive waste knowledge management comprehensive system, which systematically integrates radioactive waste information recording and information sharing. The system increases the safety management of radioactive waste, helps operators of nuclear power facilities apply existing knowledge, handles outdated and invalid knowledge, and reduces the risk of operation mistakes. The idea of a comprehensive system for knowledge management of radioactive waste is advanced, and theoretical support is provided to rapidly create a national information management system for radioactive wastes using a combination of domestic and international application cases and cutting- edge knowledge management technologies.

Background: South Korea’s extensive nuclear energy capabilities, driven by 25 operational nuclear power plants (NPPs), have been instrumental in supporting industrial development. However, concerns among antinuclear activists regarding the potential risks associated with nuclear accidents have emerged. To address these concerns, this study evaluated the long-term radiological effects of a nuclear accident on individuals living near an NPP site. In particular, this study focused on the most likely type of accident, a loss of coolant accident (LOCA), assuming that it occurs in a single unit because of the rarity of simultaneous accidents across multiple units. Materials and Methods: The source terms from the LOCA comprising 131I, 134Cs, 137Cs, and 90Sr were computed using the Radiological Assessment System for Consequence AnaLysis code, and these were fed into the GENII code, which computed the annual individual effective doses and cancer incidences at various distances and directions from the NPP. Results and Discussion: The effective dose contributions from food ingestion, ground exposure, and soil particle ingestion were 86.4%, 13.5%, and 0.01%, respectively, in children and adults. Food, particularly vegetables and meat, had the greatest impact on effective doses. Regarding individuals exposed at distances of 3, 16, and 30 km from the NPP, adults received annual effective doses of 98, 19, and 12 mSv, respectively, whereas children received doses of 56, 10, and 6.6 mSv, respectively. Comparison of the computed cancer incidence with South Korea’s past 8-year data resulted in a radiation exposure contribution to cancer of 0.012%, which was significantly below the NPP safety goal of 0.1%. Conclusion: This study suggests that radiation exposure from nuclear accidents has minimal impact on cancer incidence. Therefore, restrictions on vegetable and meat consumption should be implemented after accidents to mitigate the long-term health effects of radiation exposure.

Background: Experiments to measure the neutron activation cross-section using the 30 MeV cyclotron at Advanced Radiation Technology Institute are planned. Neutrons, generated during experiments, will activate the devices and affect the surrounding area of the facility. Therefore, analyses were conducted to confirm the radiation shielding of the facility and investigate the effects of neutron radiation. Materials and Methods: Two devices were installed for neutron activation cross-section measurements: a beam energy measurement device and a neutron generator. Monte Carlo simulation was used to investigate the effect of neutrons produced by the neutron generator on the facility and its surroundings. Results and Discussion: Dose rate distribution inside and around the facility during beam irradiation (BI) was investigated for the radiation shielding analysis. Furthermore, the dose rates at key locations around the facility were calculated and compared to the public or occupational dose limits. The activation analysis involves calculating major nuclide activities. Additionally, the dose rates that workers would receive after 30-minute and 1-hour BI were computed. Conclusion: The dose rate in the surrounding area of the facility is <0.5 μSv·hr−1, which is lower than the public dose limit. Accessing the facility takes 5 hours after the beam off for 1-hour BI and 3 hours after the beam off for 30-minute BI.

Background: As Egypt ventures into nuclear energy, ensuring the safety and security of its nuclear facilities is paramount. The ongoing construction of the El Dabaa Nuclear Power Plant (NPP) directs the necessity for Egypt to meticulously evaluate the potential outcomes of any conceivable accidents. A focused analysis on design-based accidents such as steam generator tube ruptures (SGTR) is imperative for assessing their impact on the local populace and the environment. Through comprehensive assessments, Egypt aims to enhance the safety protocols of its NPP. This preemptive strategy facilitates the identification and minimization of risk, along with the formulation of efficient emergency response frameworks. Such diligent preparations are essential for cultivating public confidence while ensuring compliance with international safety norms and securing the long-term sustainability of Egypt’s nuclear energy program. Materials and Methods: The hypothetical scenario of a SGTR at the El Dabaa NPP was modeled using the Radiological Assessment System for Consequence Analysis (RASCAL) code. This model examined the potential consequences of such an accident by incorporating meteorological data from 2013 to 2023 across all seasons to compute the source term, total effective dose equivalent (TEDE), and thyroid dose. Then, these metrics were evaluated against established safety thresholds to determine compliance. The outcomes of this analysis provide critical insights and aid in formulating strategies for effective response to an accident, especially if the calculated doses exceed the permissible limits. Results and Discussion: The analysis of the worst-case scenario for the SGTR accident at the El Dabaa NPP involves a U-tube breakage above the water level, exacerbated by a concurrent station blackout. This condition potentially leads to a considerable dispersal of radioactive materials, especially within a 0.4 km radius, with a TEDE reaching 4.80×103 mSv during autumn. However, the severity of this worst-case scenario fluctuates with seasonal weather conditions; notably in spring, the highest TEDE was 15 mSv at a 40 km distance. The source term distribution indicates that noble gases account for 31.6%, iodine group for 32.0%, and other sources constitute 36.4% of the total radioactive release. These findings confirmed that TEDE and thyroid dose exceeded the permissible thresholds, thereby highlighting the importance for protective measures to mitigate the potential risks of such accidents. Conclusion: A comprehensive assessment of SGTR accidents at the El Dabaa NPP emphasizes the critical necessity for stringent safety protocols and protective interventions in worst-case scenarios. By preemptively addressing these risks, Egypt can fortify its nuclear safety framework, emergency response capabilities, and adherence to global safety standards. This proactive stance not only assures the long-term sustainability of Egypt’s nuclear energy program but also solidifies public confidence.

Background: Boron neutron capture therapy (BNCT) is a potential radiation therapy for treating refractory cancers. Measuring the neutron energy spectrum is important to evaluate the biological effect of BNCT. We have developed a novel low-energy neutron spectrometer with a 3He position-sensitive proportional counter. The effectiveness of this spectrometer remained unvalidated due to the inadequate response function. Materials and Methods: The response function of this spectrometer was evaluated by Monte Carlo simulation to perform the validation experiment of this spectrometer. The 3He(n,p)3H reaction rate for each monochromatic neutron (energy range, 0.5 eV–10 keV) was calculated. Numerical experiments were conducted to investigate the performance of this spectrometer. Several estimated values of the energy spectrum were obtained and compared with a calculated energy spectrum at the exit of the epithermal column by the Monte Carlo simulation by preparing several candidates for signals. The Bayesian estimation algorithm was utilized for the estimation. Results and Discussion: The results indicate that the spectrometer and the evaluated response functions can be utilized to measure the energy spectrum of neutrons in the low-energy range. However, the energy spectrum could not be well represented >4 keV due to the poor response function, including uncertainties in response function calculation as well as the poor energy resolution of this detector in the high-energy region. Conclusion: The response function of the low-energy neutron spectrometer for the epithermal energy region can be assessed by the Monte Carlo simulation. The numerical experiments reveal that the effectiveness of this spectrometer can be shown in the range of 0.5 eV.

Background: This study aimed to investigate the activation characteristics of concrete in synchrotron- type proton therapy facilities for future decommissioning. The larger synchrotrontype proton therapy facilities have a greater potential impact on decontamination than the cyclotron proton therapy facilities investigated in our previous study. Specific activity levels in the concrete after 30 years of operation in synchrotron-type proton therapy facilities were predicted from the measured thermal neutron fluence rates on the concrete during the operation to compare them with the clearance level. Materials and Methods: The investigations were conducted in the synchrotron-type proton therapy facilities at Medipolis Proton Therapy Research Center and Proton Medical Research Center, University of Tsukuba Hospital. The thermal neutron fluence rates on the concrete during the operation were measured by three different methods: using 24Na radioactivity produced in concrete, thermoluminescence dosimeters, and Au foils. Results and Discussion: The specific activity levels in the concrete throughout the synchrotron proton therapy facilities were negligible compared with the clearance level. The specific activity level of concrete in the accelerator room in synchrotron-type proton therapy facilities where an accelerator controls the proton energy was much lower than that in cyclotron-type proton therapy facilities where a degrader controls the proton energy. Conclusion: Concrete does not need to be treated as radioactive waste when decommissioning synchrotron-type proton therapy facilities.

Background: The Fukushima Daiichi nuclear accident gave us a lesson on the importance of effective and timely implementation of public protective actions. National protection strategies, including decision-making process of public protective actions, should be developed at the preparedness stage of nuclear accidents. In particular, dose criteria for the decision-making of public protective actions should be predetermined. Materials and Methods: The dose criteria adopted in 24 countries and those provided by international standards were investigated for urgent protective actions (sheltering, evacuation, and iodine thyroid blocking [ITB]) and early protective actions (temporary relocation and restriction on food consumption). Other important factors such as exposure duration, dose quantities, and dose concept were also reviewed in this study. Results and Discussion: Most countries have applied different dose criteria ranging from 10 mSv to 100 mSv and exposure duration for dose estimation. An effective dose has been used for decision- making of all protective actions except for ITB, whereas thyroid dose (equivalent dose or absorbed dose) has been applied for ITB. Projected dose as the concept of radiation dose was found to be more appropriate than avertable dose due to conservative and practical reasons. The general range of dose criteria, which is in accordance with reference levels of 20–100 mSv, and exposure duration were obtained from the statistical data analysis. Conclusion: The dose criteria as a single value suggested in this study will be utilized for updating the dose criteria for decision-making of public protective actions in Republic of Korea. The revised dose criteria will be reflected in the emergency response plans of government authorities and nuclear licensees in the future.

Background: Measuring the energy spectrum of γ rays to assess the exposure dose and identifying the leaked radioisotopes at the accident site and the surrounding environment is necessary in the case of an accident involving the leakage of radioactive material at a nuclear power plant or a radiation-related facility. High-purity germanium semiconductor detectors are utilized for γ-ray spectrometry due to their high energy resolution, but they exhibit a high initial and operational cost due to the need for a cooling mechanism. Materials and Methods: We improved the unfolding method for scintillators with a refined response function that involves the energy dependence of scintillation efficiency for secondary electrons produced by incident γ-ray interaction. Monte Carlo simulation code EGS5 with a mesh width of 5 keV in the energy range of 0–3 MeV was used to calculate response functions, assuming γ rays irradiated parallel to the detector sides. The unfolding algorithm involves an iterative approximation method, which is independent of initial guesses. Results and Discussion: The measurement accuracy of the γ-ray fluence rate was almost constant at <10% at ≥0.3 MeV caused by repeated measurements using 137Cs, 133Ba, 88Y, and 60Co radiation sources. Additionally, we confirmed linearity concerning the γ-ray intensity by changing the distance from the source to the detector. Conclusion: We verified that the unfolding method separated peaks for each γ-ray energy, although the difference in γ-ray energies was several tens of keV. Moreover, the accuracy of the unfolding method was almost constant and had linearity concerning γ-ray intensity.

Background: Soft errors in semiconductor devices caused by cosmic rays have been recognized as a significant threat to the reliability of electronic devices on the ground. Recently, concerns about soft errors induced by cosmic-ray muons have increased. Some previous studies have indicated that low-energy negative muons have a more significant contribution to the occurrence of soft errors than positive muons. Thus, charge-identified low-energy muon flux data on the ground are required for accurate evaluation of the soft error rate. However, there are no such experimental data in the low-energy region. Materials and Methods: We designed a new muon detector system to measure low-energy muon flux data with charge identification. The major components consist of two drift chambers and a permanent magnet. The charge and momentum of detected muon can be identified from the deflection of the muon trajectory in the magnetic field. An algorithm to estimate the muon momentum is developed using numerical optimization by combining the classical Runge-Kutta and quasi-Newton methods. The momentum search algorithm is applied to event-by-event data of positive and negative muons obtained by Monte Carlo simulations with Particle and Heavy Ion Transport code System, and its performance is examined. Results and Discussion: The momentum search algorithm is fully applicable even in the case of an inhomogeneous magnetic field. The precision of the momentum determination is evaluated by considering the stochastic fluctuation caused by multiple scattering and the position resolution of the drift chambers. It was found that multiple scattering has a significant contribution to the precision in the momentum region below 50 MeV/c, while the detector position resolution considerably affects the precision above that. Conclusion: It was confirmed that the momentum search algorithm works well with a sufficient precision of 15% in the low-momentum region below 100 MeV/c, where no muon flux data exist.

Background: At nuclear facilities and decommissioning sites, monitoring radioactivity concentrations in airborne particulates is crucial to prevent worker exposure. To avoid internal exposure, alpha-decay radionuclides must be detected. When monitoring alpha-decay radionuclides in airborne dust, we want to measure the concentration of only artificial radionuclides (e.g. , 238Pu, 239Pu, 240Pu, 235U, 238U, 241Am, and 244Cm). The radioactivity concentration must be measured separately for artificial (4.3–5.8 MeV) and natural radionuclides (212Bi, 214Po, and 212Po at 6.0, 7.7, and 8.8 MeV, respectively). Materials and Methods: We created response functions for various alpha-ray energies using a radiation simulation toolkit. Utilizing these response functions, we deconvolved the alpha-ray energy spectra measured while collecting dust on filter paper. To ensure the precision of the response function, we prepared a model including the distance between the filter and the detector and the structure of the light shield in detail. Results and Discussion: The deconvolved spectra had three clear peaks at 6.0, 7.7, and 8.8 MeV. These energies were consistent with those of 212Bi, 214Po, and 212Po. The deconvolved energy spectra showed that only a few measurements (4.0–5.8 MeV) were included in the energy range due to artificial radionuclides. From these measurements, the decision threshold for artificial radionuclides was determined to be about 1.5×10−7 Bq/cm3. Conclusion: Our findings demonstrated that we could measure artificial and natural radionuclides separately with the deconvolved alpha-ray energy spectra. Specifically, we were able to monitor artificial radionuclides down to low radioactivity concentrations in 10-minute measurements.

Background: Proton therapy facilities have improved in recent years, not only providing benefits to patients but also bringing potential radiation hazards and risks. Induced radioactivity in patients is a crucial issue that remains unresolved for a long time and deserves further investigation. Therefore, this study aimed to investigate the induced radioactivity in proton therapy rooms under various treatment conditions. Materials and Methods: The Monte Carlo code FLUktuierende KAskade (FLUKA) and advanced interface Flair are employed to simulate residual dose rate distribution and different material activation levels in the proton therapy room. The simulation involved irradiating the phantom with energies of 220, 150, and 70 MeV for 2 minutes, respectively. Results and Discussion: The total specific activity of the patient exceeds that of other materials but rapidly attenuates thereafter after 2 minutes of irradiation. The residual dose rate at shutdown exceeds 2.50 μSv/hr at 30 cm from the patient’s surface in high energy conditions, and the time required to reduce it to 2.50 μSv/hr differs with different energy scenarios (10 minutes, 8 minutes, and none for scenarios A, B, and C, respectively). Conclusion: The patient plays a crucial role in identifying the radiation dose, although their importance significantly diminishes over time due to the presence of radionuclides with short half-lives. Diverse irradiation scenarios cause varying activation levels. Thus, we manage the timing of staff entry for positioning or escorting in the treatment room and introduce customized protective measures based on the specific requirements of the tumor.

Background: This research focuses on the simulation techniques for the migration and diffusion of aircraft radioactive material, evaluating the applicability and simulation performance of various modules. Materials and Methods: A multiscale meteorological forecast simulation was developed to provide grid wind fields for dispersion modeling. In collaboration with the National Nuclear Emergency Response Technical Support Center, the China Institute for Radiation Protection has successfully developed the Nuclear Accident Consequence Assessment and DecisiOn Support System (NACADOS). This system, which has independent intellectual property rights, is specifically tailored to China’s national conditions and incorporates multiscale response capabilities for nuclear accident consequence assessment. This system also includes diffusion modules applicable to various evaluation scopes and environmental conditions. Results and Discussion: The effectiveness of the module’s assessment was validated through comparisons with on-site monitoring data, high-quality tracer experiment results (e.g. , European Tracer Experiment, Urban2000), and simulation outcomes from other consequence assessment systems (e.g. , Java-based Real-time On-line Decision Support System [JRODOS] and Radiological Assessment System for Consequence Analysis for radiological emergencies [RASCAL]). Furthermore, the functionality, performance, and effectiveness of the NACADOS system were rigorously tested through physical module validation, system testing, and engineering software testing. Conclusion: The NACADOS system meet the demands of radiological consequence assessment and response for nuclear and radiological emergencies worldwide, significantly enhancing nuclear emergency response capabilities.

Background: This study validated the applicability of employing artificial neural networks (ANNs) to convert aerial radiation survey data, constructed from experiences in Fukushima, into ambient dose rates at 1 m above ground level. Materials and Methods: The ANN utilized in this investigation was constructed based on the data collected after the Fukushima Daiichi Nuclear Power Plant accident. To validate the applicability of ANN, we applied it to the data used as input to ANN obtained from a detector that was different from the original detector. Results and Discussion: The measurement data from the different detectors were able to output the conversion results by ANN with equivalent accuracy as the original detectors using coefficients to convert the counting levels of the detectors. Conclusion: The results of this study suggest the potential for applying ANN generated from the accumulation of valuable monitoring data in Fukushima to other contexts.

Background: The construction and operation of nuclear facilities require that retrospective environmental impact assessments (REIAs) be conducted to identify the realistic radiologic impact of those facilities, thereby providing useful information for minimizing effects on the public and the environment. Materials and Methods: A retrospective assessment of nuclear facilities has two main aspects: spatial distribution of radionuclide concentrations and assessment of the dose to the representative individual. In the present study, a comprehensive REIA method was established in combination with effluent discharge and environmental monitoring data obtained for the nuclear facility. Using historical effluent discharge data and local parameters, the spatial distribution of radionuclide concentrations was simulated, and the representative individual was identified. In combination with the monitoring data, the actual radiation doses that representative individuals received were further assessed, ensuring more accurate assessment results. Results and Discussion: Effluent discharge and environmental monitoring data series collected between 2002 and 2019 for uranium processing facility in China were analyzed. Results indicated a significant declining trend in the radioactive effluent per 100 tons of raw material during that period. Based on the actual data for the discharge of airborne effluent, the maximum effective dose to the representative individual was 0.015 μSv/yr. In comparison, the dose to the representative individual based on environmental monitoring data was close to zero. Those results suggest that the uranium processing facility has had little radiologic impact on the representative individual, despite many years of operation. Conclusion: The REIA method proposed in this study can use effluent discharge data in conjunction with environmental monitoring data to effectively identify the representative individual and assess the realistic radiologic impact to that individual.

Background: Field atmospheric diffusion tracer experiments are commonly used methods for verifying the effectiveness of atmospheric diffusion models in nuclear accident consequence assessment systems. China Institute for Radiation Protection (CIRP) is conducting a study to verify the effectiveness of the nuclear accident consequence assessment system. This study proposes to conduct field atmospheric diffusion tracer experiments in different environmental conditions in China and use the experimental data to evaluate the effectiveness of model simulation in the China Nuclear Accident Consequence Assessment System (NACADOS). Materials and Methods: Some atmospheric diffusion tracer experiments have been conducted at a relatively flat coastal plant site. To evaluate the applicability of diffusion models in the Realtime On-line DecisiOn Support of Java version (JRODOS) system at China’s nuclear power plant sites, and the effectiveness of the model in the China NACADOS, the diffusion model Risø Mesoscale PUFF model (RIMPUFF) from JRODOS and a three-dimensional particle diffusion model from NACADOS were selected to simulate the sulfur hexafluoride (SF6) tracer experiment in the CIRP effectiveness verification research project. Similar to the currently commonly used methods for evaluating the effectiveness of atmospheric models, the simulation results were compared with the experimental results using various indicators such as fit factor (FACT), mean error (BIAS), and other statistical measures. Results and Discussion: The BIAS, mean absolute error, root mean square error, and normalised mean square error show that the model in JRODOS is better. The FACT2 and FACT3.5 also indicate that the simulation performance of the JRODOS models is better, but the FACT5 and FACT10 indicate that the NACADOS model is slightly better. Overall, the difference in model simulation results between the two systems is not particularly significant. Conclusion: For the currently conducted atmospheric diffusion tracer experiments in flat terrain conditions, the performance of the diffusion model in JRODOS is slightly better, and the difference in simulation results between the two systems is not particularly significant.