Decommissioning Of Nuclear Power Plants Research Articles

Ultra-fast, selective and pseudo-quantitative analysis of 99Tc in nuclear waste for screening purposes

Controlling radioactivity is essential in various fields, such as the decommissioning of nuclear power plants and nuclear medicine. In some cases, a full characterization of samples is not required; instead, a screening analysis that provides an overall indication of the activity present can be sufficient to determine if a sample is radioactive. This article introduces a new system called PSkits designed specifically for ultra-fast and selective screening detection of 99Tc. PSkits consist of a plastic scintillation layer attached to the bottom of a scintillation vial, coated with aliquat·336® as a selective extractant. In this study, the preparation of PSkits was optimized by adjusting the proportions of crosslinker, porogen, and the type of vial used. The analysis method was developed, and the selectivity against common interferences was tested by optimizing the rinsing media. Finally, PSkits were validated by analysing simulated nuclear waste samples and urine spiked samples, achieving satisfactory results with quantification errors below 50 %, demonstrating their effectiveness for the intended purpose

Proposal of database of radiological characterization for decommissioning nuclear power plants

AbstractIn Korea, Kori Unit 1 was permanently closed in 2017 and scheduled for decommissioning. Kori Unit 1 is currently scheduled to undergo system decontamination and to be decommissioned by December 2038. Measures to properly control and remove radiological hazards must be established to go through decommissioning a nuclear power plant (NPP). Radiological characterization must precede NPP decommissioning. Radiological characterization generates a lot of data during this process. The consistency of radiological characterization data is important for systematically decommissioning NPPs. However, the data generated from radiological characterization may not be systematically managed because a database of radiological characterizations for Kori Unit 1 has not been properly established. Therefore, this study has proposed a database to be filled during radiological characterization in terms of occupational radiation exposure of personnel and decommissioning waste safety using key performance indicators presented by the International Atomic Energy Agency. From the perspective of occupational exposure, databases have been proposed for operational history investigations, decommissioning source term calculations, and investigation of the level of radiological contamination. From the perspective of the safety of decommissioning waste, databases were proposed for operational history investigation, decommissioning source term calculation, investigation of the level of radiological contamination, and investigation of radiological characteristics of decommissioning waste.

Mass dependence of the sensitivity factors in the efficiency calibration of clearance monitors

Context:The radioactivity level of waste materials from the decommissioning of nuclear power plants and particle accelerators must be measured to classify the materials by comparing the obtained activity value with the legal exemption levels set by the regulatory authorities. Clearance monitors (CMs), commonly used for this purpose, require a scenario-dependent efficiency calibration (e.g. efficiency as a function of mass of the material to be measured) using sensitivity factors related to a reference nuclide called the key (or leading) nuclide. In commercial CMs, however, these factors are typically fixed for a given nuclide, regardless of the different measurement scenarios. Objective:The objective of this study was to determine the dependence of the sensitivity factors on the mass of measured material for a specific scenario (copper parts placed inside a stainless steel drum), with the goal of defining a general recommendation on how to select the appropriate sensitivity factor for a given measurement scenario. Approach:Monte Carlo simulations of the CM were performed for this specific scenario to obtain the sensitivity factor for five nuclides (22Na, 133Ba, 137Cs, 152Eu and 232Th) and various geometries. Measurements were carried out with certified sources of those nuclides for four specific geometries to validate the Monte Carlo calculations. Additionally, clearance measurements using the manufacturer-provided sensitivity factors and the factors calculated in this work were compared to determine in which situations one single sensitivity factor could be used for various geometries. Results:The results show that, in case of measurements with the same material but mass varying from 10.0 kg to 40.0 kg, the use of a single sensitivity factor for monoenergetic nuclides (e.g. 137Cs), if not selected properly, leads to an over- or underestimation of the activity level of the measured materials. If overestimation is not an issue, the lowest sensitivity factor can be used for all possible measurement configurations. Otherwise, a situation-dependent sensitivity factor should be estimated and implemented. The use of a single fixed sensitivity factor can be easily justified in the case of nuclides emitting gammas in a wide energy range (e.g. 152Eu), where the dependence of these factors in the studied mass range is small. Significance:These findings facilitate more accurate decisions regarding the calibration of CMs with the aim of classifying waste from the dismantling of nuclear power plants and particle accelerators, with potential logistic and economic implications.

Design Change and Operational Consideration of the HVAC System during Nuclear Power Plant Decommissioning

The heating, ventilation, and air conditioning (HVAC) system plays a crucial role in ensuring the safety of workers and preventing the release of gaseous radioactive materials into the environment during the decommissioning of a nuclear power plant (NPP). To establish an HVAC operation strategy, decommissioning phases were divided into four stages, and the HVAC systems were reclassified. In addition, assumptions have been made regarding design modifications and maintenance for the reactor containment building (RCB) HVAC, fuel handling building (FHB) HVAC, and main control room (MCR) HVAC. Based on these, for RCB HVAC operation, natural ventilation and RCB purge operation during the transition period are proposed. In the decommissioning stage, recirculation operation, entire ventilation operation consisting of continuous operation and purge operation, and finally partial ventilation operation to purify local space were proposed. Moreover, during the transition period, the FHB HVAC was proposed to operate as normal NPP, and the MCR HVAC was suggested to operate with safety-related equipment removed.

A Review of Recent Low-dose Research and Recommendations for Moving Forward.

The linear no-threshold (LNT) model has been the regulatory "law of the land" for decades. Despite the long-standing use of LNT, there is significant ongoing scientific disagreement on the applicability of LNT to low-dose radiation risk. A review of the low-dose risk literature of the last 10 y does not provide a clear answer, but rather the body of literature seems to be split between LNT, non-linear risk functions (e.g., supra- or sub-linear), and hormetic models. Furthermore, recent studies have started to explore whether radiation can play a role in the development of several non-cancer effects, such as heart disease, Parkinson's disease, and diabetes, the mechanisms of which are still being explored. Based on this review, there is insufficient evidence to replace LNT as the regulatory model despite the fact that it contributes to public radiophobia, unpreparedness in radiation emergency response, and extreme cleanup costs both following radiological or nuclear incidents and for routine decommissioning of nuclear power plants. Rather, additional research is needed to further understand the implications of low doses of radiation. The authors present an approach to meaningfully contribute to the science of low-dose research that incorporates machine learning and Edisonian approaches to data analysis.

Impurity sensitivity analysis for activated structure materials of OPR-1000

Various kinds of radioactive wastes are expected to be generated resulting from the decommissioning of nuclear power plants. For the safe management of decommissioning wastes, the preliminary classification of radioactive wastes such as very low level, low level, intermediate level, and regulatory clearance has to be performed. In this regard, we intend to conduct an activation analysis and radioactive waste level estimation using a computational simulation code for the structure materials of an OPR-1000, a standard nuclear power plant in South Korea. In addition, the structure material that constitutes the reactor vessel, reactor vessel internals, and bio-shield concrete of the nuclear power plant contains small amounts of unintended impurities. The results of the activation analysis may vary depending on the presence of those impurities. In this study, a detail modeling of nuclear reactor and the activation analysis are carried out combined the MCNP code with the ORIGEN-S code. Based on this methodology, the sensitivity test is done for impurities of structure materials. From the result of the specific activity of activated structure material, the cobalt impurity is found to be the most important contributor in the level determination of radioactive wastes. However, the bio-shield concrete has the highest specific activity when Li impurity is contained. From this study, some impurities affect radioactive waste level, thus it is recommended to consider structure material impurities for decommissioning the nuclear power plants.

Development of Remote Construction System of Modularized Rail Structure for Inexperienced Operators

Remotely controlled robots are required to operate in human-inaccessible environments, such as the decommissioning of nuclear power plants. In our previous studies, we proposed a modularized rail structure for such environments. In long-term decommissioning work at a nuclear power plant, robots are controlled by inexperienced operators who have limited knowledge of the design details of the robots. In this study, we developed a remote construction system for supporting inexperienced operators by transmitting the status information of the constructor robot to the operator during the remote construction of a modularized rail structure. The proposed remote construction system was designed based on a human information-processing model. Experiments were conducted using the developed remote construction system. The proposed method achieved a success count of inexperienced operators equivalent to that of skilled operators. Furthermore, a questionnaire administered to participants showed that their cognitive, psychological, and physical loads during construction were sufficiently low during the experiment.

Dismantling Quantity Estimation for Nuclear Power Plant: Scan-to-BIM versus Conventional Method

This paper presents a case study of using Scan-to-BIM technology for evaluating the quantity and cost of a nuclear power plant during the decommissioning process. The study proposes a process for applying Scan-to-BIM in the dismantling stage and develops a BIM-based VPL (Visual Programming Language) to acquire accurate as-is databases that are crucial for nuclear power plants handling radioactive materials. The proposed process and system were applied to K-1 NPP (Nuclear Power Plant), the first nuclear power plant in Korea, and the results were compared to those obtained using the conventional method. The study found that errors in quantity evaluation using the conventional method ranged from around 10% to over 10 times, and were attributed to various causes such as outdated as-is drawings and limitations in expressing information in drawings. By using the Scan-to-BIM process and system, errors in quantity evaluation were significantly reduced, and the conventional method was found to have underestimated the dismantling cost by 20%. The findings of this study provide important insights for the efficient and accurate decommissioning of nuclear power plants, and demonstrate the effectiveness of the proposed method and system for evaluating the quantity and cost of a nuclear power plant during the decommissioning process.

Effects of EDTA on the leaching behaviors of Cs(I) and Co(II) from cement waste forms

Cementation is extensively employed for immobilizing radionuclides in low- and intermediate-level radioactive wastes generated during the decommissioning of nuclear power plants. Ethylenediaminetetraacetic acid (EDTA), used as a decontamination agent during the decommissioning process, can be introduced into the cement waste form containing radionuclides. This study investigated the effects of the EDTA present in simulated radioactive decommissioning wastes on the leaching behaviors of Co and Cs immobilized in the cement waste form. Co leaching was facilitated by the formation of highly mobile Co–EDTA complexes. However, Cs leaching was impeded by the competition for leaching with other metal–EDTA complexes. Moreover, the EDTA-induced carbonated layer with a dense pore structure played a crucial role as a retardation barrier for the Cs leaching. The calcite contents of the samples with 8 wt% EDTA were approximately three times higher than those of the samples without EDTA. The introduced EDTA affected the leaching behaviors of both Co and Cs, as well as the microstructure evolution of the cement waste form. Nevertheless, the addition of EDTA had a relatively low positive effect on the efficiency of Cs immobilization, but also an obvious negative effect on the efficiency of Co immobilization, regardless of the concentration of EDTA. Finally, an EDTA dosage of 1 wt% in the cement waste forms containing Co or Cs is suggested as a potential waste acceptance criterion for solidified low- and intermediate-level radioactive waste.

Development of metal radioactive liquid reference material for proficiency test

This study developed liquid reference materials containing various metals, to be used for quality assurance of radiation measurements of the most common metallic wastes generated during the operation or decommissioning of nuclear power plants. The liquid reference materials were prepared by assuming the dissolution of stainless-steel using acid and melting of the major metals present in the stainless steel, namely Fe, Ni, Cr, and Mn, along with the standard sources (134Cs, 137Cs, 60Co, 90Sr). The theoretical massic activity of the standard sources added to the samples and the measured reference values of the gamma and beta emitters in the samples were compared, and they showed good agreement within a one-sigma confidence interval (k = 1). Using the developed reference materials, a proficiency test was conducted on three domestic labs, and the results were evaluated using Z-score. While the evaluation results showed good agreement between the reference values and the reported values for 137Cs and 60Co, all participating labs reported lower values than the reference value for 134Cs. For 90Sr, two out of the three labs reported significantly higher values than the reference value. Based on the results of this study, the developed metal radioactive liquid reference material is expected to be registered as certified reference material (CRM) in the future. They will be used as the CRM for measuring and ensuring the quality of radioactive metal waste.

Study on probabilistic safety assessment in decommissioning of nuclear power plant

The decommissioning of Nuclear power plant involves many activities different from the normal operation stage of Nuclear power plant. In the process of decommissioning, there are many new safety problems, which usually require new safety evaluation. The safety management of decommissioning runs through the whole decommissioning stage until the decommissioning of the plant site is completed. The management level depends on the risk level of each stage of decommissioning, especially the decommissioning process of high radioactivity in nuclear power plant, so it is necessary to pay attention to the risk of different stages of decommissioning and improve the decommissioning safety management. At present, probabilistic safety assessment method has become an effective tool to support and supplement deterministic safety assessment, and is used in regulatory activities. The safety risk analysis of nuclear power plant in the decommissioning stage in China is in the initial stage, with little experience. Therefore, it is necessary to study the corresponding technical methods and risk management applications. This paper study on the probabilistic safety assessment in the decommissioning of nuclear power plant, including its scope of application, risk-informed assessment method and safety management application. The probabilistic safety assessment of decommissioning of nuclear power plant is an effective means to evaluate the potential risks related to decommissioning activities and their possible consequences. This method can be used for safety review and safety management in the decommissioning stage of nuclear power plant.

Evaluation of Radiological Effects on the Aptamers to Remove Ionic Radionuclides in the Liquid Radioactive Waste

Background: Aptamers are currently being used in various fields including medical treatments due to their characteristics of selectively binding to specific molecules. Due to their special characteristics, the aptamers are expected to be used to remove radionuclides from a large amount of liquid radioactive waste generated during the decommissioning of nuclear power plants. The radiological effects on the aptamers should be evaluated to ensure their integrity for the application of a radionuclide removal technique.Materials and Methods: In this study, Monte Carlo N-Particle transport code version 6 (MCNP6) and Monte Carlo damage simulation (MCDS) codes were employed to evaluate the radiological effects on the aptamers. MCNP6 was used to evaluate the secondary electron spectrum and the absorbed dose in a medium. MCDS was used to calculate the DNA damage by using the secondary electron spectrum and the absorbed dose. Binding experiments were conducted to indirectly verify the results derived by MCNP6 and MCDS calculations.Results and Discussion: Damage yields of about 5.00×10<sup>-4</sup> were calculated for 100 bp aptamer due to the radiation dose of 1 Gy. In experiments with radioactive materials, the results that the removal rate of the radioactive <sup>60</sup>Co by the aptamer is the same with the non-radioactive <sup>59</sup>Co prove the accuracy of the previous DNA damage calculation.Conclusion: The evaluation results suggest that only very small fraction of significant number of the aptamers will be damaged by the radioactive materials in the liquid radioactive waste. Keywords: Aptamer, Liquid Radioactive Waste, Decommissioning, Decontamination, Radiological Damage

Development of a methodology to select suitable cutting techniques for the decommissioning of nuclear power plants considering site-specific requirements

Development of a methodology to select suitable cutting techniques for the decommissioning of nuclear power plants considering site-specific requirements

Assessment of chlorination technique for decontamination of radioactive concrete waste using simulated concrete waste

A chlorination technique was investigated as an option for the decontamination of radioactive concrete waste, which is one of the most common types of wastes generated during the decommissioning of nuclear power plants. Simulated concrete waste that includes target nuclides of Co, Sr, and Cs was utilized to evaluate the removal efficiency of each nuclide after the chlorination process. The effects of the reaction temperature, the Cl2 flow rate, and the reaction time were explored in this study. Among these parameters, the reaction temperature was identified as the key parameter that most strongly affects the removal efficiency. The removal efficiency of Cs surpassed 90% when the temperature exceeded 800 °C, while that of Co was over 95% at 1000 °C. The removal efficiency of Sr was relatively lower at 70–80% at temperatures of 700 and 800 °C. It was estimated that the chlorination technique can reduce the radioactivity of Cs and Co in low-level concrete waste to meet the clearance level. In consideration of its simple process, the chlorination technique can be a promising option for the decontamination of radioactive concrete waste.

Development of concrete reference material for quality assurance/quality control of gamma radioactivity measurement for nuclear power plant decommissioning waste

The Korea Research Institute of Standards and Science has developed a new concrete reference material (RM) for measuring gamma-emitting radionuclides, such as 134Cs, 137Cs, 65Zn, 241Am, and 60Co, to improve and maintain the quality assurance and quality control of the radioactivity measurement in radioactive waste generated during the decommissioning of nuclear power plants. In this study, cement, SiO2, and bentonite, which are the main components of concrete, were mixed in an appropriate ratio with radionuclides. For certification and homogeneity assessment, 10 bottles were randomly selected, two sub-samples were collected from each bottle, and radionuclides were measured via HPGe gamma spectrometry. The results of the homogeneity tests using a one-way analysis of variance on 241Am, 134Cs, 137Cs, 65Zn, and 60Co in the concrete RM fulfilled the requirements of ISO Guide 35. Coincidence summing and self-absorption correction were performed on measurement results by introducing the Monte Carlo efficiency transfer code and Monte Carlo N-Particle transport code. The reference values for five radionuclides (60Co, 65Zn, 241Am, 134Cs, and 137Cs) in the RM were in the range of 15–40 Bq/kg, and the expanded uncertainty was within 10% (k = 2). To the best of our knowledge, this was the first study to develop concrete RM for measuring gamma-emitting radionuclides.

Prioritization of radiological accident scenarios during decommissioning of nuclear power plants by risk matrix and AHP method

As part of the safety evaluation during the decommissioning process of Kori Unit 1, candidate scenarios were selected using the risk matrix method. Since Korea has no experience in dismantling commercial nuclear power plants, systematic research on how to select accident scenarios in risk assessment is performed. In this paper, the priority of 11 candidate scenarios selected by the risk matrix method is determined using the AHP method to facilitate the selection of additional scenarios.

Demolition of Reinforced Concrete by Steam Pressure Cracking System

The authors developed an environment-friendly demolition mechanical system for a large reinforced concrete structure for an actual site. The steam pressure cracking agent (SPC, non-explosive) is a method that can safely and quickly separate concrete because it produces lesser vibration and sound than the blasting method, which uses explosives. The authors showed that the direction of cracking can be controlled by an induction hole. The principle of control is that the elastic wave of the compression stress generated from the SPC reaction changes to a tensile elastic wave at the induction hole, which initiates a crack. Furthermore, in the SPC method, a large amount of concrete powder generated by the explosion method was not produced, and there was no risk of secondary contamination by fine concrete powder. The area over which the crack propagated depends on the energy generated from the SPC. The relationship between the two is linear. For reinforced concrete, the energy of the SPC is used for both the destructive energy of the concrete and the energy of the cutting of the reinforcing steel bar, which quickly breaks with low energy. By applying an SPC to dismantle large reinforced concrete structures, controlled cracking can be achieved safely and quickly without any environmental pollution. A fracturing method using a SPC is an effective method for the decommissioning of nuclear power plants and the dismantling of concrete structures. In this report, we report a remote drilling system that can be used to remotely install loading holes and guiding holes for the SPC and perform effective controlled fracturing.

Surface removal of stainless steel using a single-mode continuous wave fiber laser to decontaminate primary circuits

Removing radioactive contaminated metal materials is a vital task during the decommissioning of nuclear power plants to reduce the cost of the post-dismantling process. The laser decontamination technique has been recognized as a key tool for a successful dismantling process as it enables a remote operation in radioactive facilities. It also minimizes exposure of workers to hazardous materials and reduces secondary waste, increasing the environmental friendless of the post-dismantling processing. In this work, we present a thorough and efficient laser decontamination approach using a single-mode continuous-wave (CW) laser. We subjected stainless steels to a surface-removal process that repetitively exposes the laser to a confined region of ∼75 μm at a high scanning rate of 10 m/s. We evaluate the decontamination performance by measuring the removal depth with a 3D scanning microscope and further investigate optimal removal conditions given practical parameters such as the laser power and scan properties. We successfully removed the metal surface to a depth of more than 40 μm with laser power of 300 W and ten scans, showing the potential to achieve an extremely high DF more than 1000 by simply increasing the number of scans and the laser power for the decontamination of primary circuits.

Simplified Assessment Methodology of Radionuclide Contamination for Subsurface Soil

Determining whether to release a site after decommissioning a nuclear facility should be preceded by an environmental impact assessment of the exposure radiation dose according to the radionuclides in the soil. Currently, in Korea, various evaluation methodologies and decommissioning technologies are being studied for the first decommissioning of nuclear power plants, starting with Kori Nuclear Power Plant Unit (Kori-1), which is based on the “Multi-Agency Radiation Survey and Site Investigation Manual MARSSIM” developed in the United States. The scope and evaluation targets of deep soil may differ depending on the purpose, but it has been confirmed that the International Atomic Energy Agency and the U.S. Nuclear Regulatory Commission are targeting subsurface soil. MARSSIM outlines the need for an evaluation of this subsurface soil but does not suggest specific methods. In NUREG-1757, which complements MARSSIM, it is confirmed that subsurface soil specifically means a soil layer that is 15 to 30 cm deep in the surface layer. In the current study, using the previously verified computational code RESidual RADioactivity (RESRAD)-ONSITE, a methodology for summation is proposed to evaluate the impact of subsurface soil more flexibly and realistically while minimizing the exposure dose evaluation procedure. When using RESRAD-ONSITE according to this evaluation methodology, it was confirmed that it is possible to respond to changes in the depths of various soil layers. In addition, it was also confirmed that this methodology is adaptable to the contamination of nuclides, such as 60Co, 137Cs, 152Eu, and 154Eu, which are expected to be major nuclides when decommissioning nuclear facilities.

Improved TDV algorithm for three-dimensional space path planning in a complex radioactive environment with obstacles

Decommissioning of nuclear power plants is a complex process that exposes workers to radiation risks. Recently, many shortest path algorithms have been proposed to find a minimum dose path and the applications are generally based on three concepts: environmental modeling, radiation dose computation, and shortest path algorithm. However, the conventional minimum dose path search methods are sub-optimal because they suffer from semi-empirical approximation, complex geometries restrictions, and lack integrated environmental modeling capability. This study presents an improved three-degree of vertex (iTDV) algorithm, an extension of the novel three-degree of vertex (TDV) shortest path algorithm previously developed. In this work, we enhance its search mechanism to find the shortest path with the least amount of radiation dose in a complex three-dimensional (3D) space. The work also outlines the effective coupling of TDV with FLUKA Monte Carlo environmental modeling, which prevents interpolation uncertainty associated with the conventional path planning methods when dealing with large geometries. To demonstrate its efficiency, the iTDV is used to find the minimum dose walking path in environments with and without obstacles. The performance of the iTDV algorithm is also compared with the original TDV, and the Dijkstra's algorithm. The evaluation results and the improvements of the proposed method are discussed in this paper.