• Journal of Korean Society of Industrial and Systems Engineering
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• v.27 no.4
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• pp.1-6
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• 2004

For a given plant design, improved decisions on when to shutdown an existing plant may be obtained by making better predictions of failure rates, by exerting efforts to collect more relevant information or by improving decision making models which put that information to best use. It is important that the models include the value of possible loss of lives and fear along with cleanup, decommissioning, relocation if the decisions derived from the model are to be useful. The decision model we have described enables us to investigate a class of optimal decisions on whether to shutdown or continue operating one period of time. The analysis and decision process is repeated at the end of each period with additional information about new costs and risks.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.556-561
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• 2017

Serving as an excellent adsorbent and inorganic ion exchanger in the water purification field, zeolite 4A has in this work presented a strong capability for purifying radioactive waste, such as $Sr^{2+}$, $Cs^+$, and $Co^{2+}$ in water. During the processes of decontamination and decommissioning of suspended solids and organics in low-level radioactive wastewater, the purification performance of zeolite 4A has been studied. Under ambient temperature and neutral condition, zeolite 4A absorbed simulated radionuclides such as $Sr^{2+}$, $Cs^+$, and $Co^{2+}$ with an absorption rate of almost 90%. Additionally, in alkaline condition, the adsorption percentage even approached 98.7%. After conducting research on suspended solids and organics of zeolite 4A for the treatment of radionuclides, it was found that the suspended clay was conducive to absorption, whereas the absorption of organics in solution was determined by the species of radionuclides and organics. Therefore, zeolite 4A has considerable potential in the treatment of radioactive wastewater.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.1010-1018
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• 2017

The European Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project sought to develop best practices in the retrieval, treatment, and disposal of irradiated graphite including other irradiated carbonaceous waste such as structural material made of graphite, nongraphitized carbon bricks, and fuel coatings. Emphasis was given on legacy irradiated graphite, as this represents a significant inventory in respective national waste management programs. This paper provides an overview of the characteristics of graphite irradiated during its use, primarily as a moderator material, within nuclear reactors. It describes the potential techniques applicable to the retrieval, treatment, recycling/reuse, and disposal of these graphite wastes. Considering the lifecycle of nuclear graphite, from manufacture to final disposal, a number of waste management options have been developed. These options consider the techniques and technologies required to address each stage of the lifecycle, such as segregation, treatment, recycle, and ultimate disposal in a radioactive waste repository, providing a toolbox to aid operators and regulators to determine the most appropriate management strategy. It is noted that national waste management programs currently have, or are in the process of developing, respective approaches to irradiated graphite management. The output of the Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project is intended to aid these considerations, rather than dictate them.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1224-1235
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• 2021

An integrated waste management approach for irradiated graphite was developed during the European Commission project 'Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste'. This included the identification of potential options for the management of irradiated graphite, taking account of storage, retrieval, treatment and disposal methods. This paper describes how these options can be assessed using multi-criteria decision analysis (MCDA) for a case study relating to a generic power reactor. Criteria have been defined to account for safety, environmental, economic and socio-political factors, including radiological impact, resource usage, economic costs and risks. The impact of each option against each criterion has been assessed using data from the project and the wider literature. A linear additive approach has been used to convert the calculated impacts to scores. To account for the relative importance of the criteria, example weightings were allocated. This application has shown that MCDA approaches can be used to support complex decisions regarding irradiated graphite management, accounting for a wide range of criteria. Use of this approach by individual countries or organisations will need to account for the specific options, scores, weightings and constraints that apply, based on their national strategies, regulatory requirements and public acceptability.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2277-2288
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• 2021

To regulate the safety protocols in nuclear facilities, radioactive aerosols have been extensively researched to understand their health impacts. However, most measured particle-size distributions remain at low resolutions, with the particle sizes ranging from nanometer to micrometer. This study combines the high-resolution detection of 500 size classes, ranging from 6 nm to 10 ㎛, for aerodynamic diameter distributions, with a regional lung deposition calculation. We applied the new approach to characterize particle-size distributions of aerosols generated during the plasma arc cutting of simulated non-radioactive steel alloy wastes. The high-resolution measured data were used to calculate the deposition ratios of the aerosols in different lung regions. The deposition ratios in the alveolar sacs contained the dominant particle sizes ranging from 0.01 to 0.1 ㎛. We determined the distribution of various metals using different vapor pressures of the alloying components and analyzed the uncertainties of lung deposition calculations using the low-resolution aerodynamic diameter data simultaneously. In high-resolution data, the changes in aerosols that can penetrate the blood system were better captured, correcting their potential risks by a maximum of 42%. The combined calculations can aid the enhancement of high-resolution measuring equipment to effectively manage radiation safety in nuclear facilities.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.2011-2018
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• 2021

Nowadays, evaluation of amounts and distributions of radioactive waste is an important preparatory step in the process of nuclear reactor decommissioning. For tentative estimation of radioactive waste, a cell-based rigorous 2 step (R2S) method usually is used; however, a poor resolution caused by the averaged flux and spectrum in a cell is still a great challenge because of leading to underestimated or overestimated results. To overcome the poor resolution, several systems were introduced. Neither system, however, provides any function for evaluation of radioactive waste amount and distribution. Thus, it is additionally required to classify radioactive waste based on the results of activation calculation. In this study, the advanced R2S (AR2S) system was developed. To verify the performance of the system, its results for a verification problem were compared with those of the cell-based R2S method. The results showed good agreement, which is to say, within 2.0% relative error. Also, several characteristics of fine/coarse mesh were analyzed. To demonstrate the performance of the AR2S system, the radioactive waste from the Japan Power Demonstration Reactor (JPDR) was estimated, and the result indicated a high-resolution distribution. Therefore, it is expected that the AR2S system will prove useful for precise evaluation of radioactive waste.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.4
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• pp.517-536
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• 2020

The radionuclide inventory in radioactive waste from nuclear power plants should be determined to secure the safety of final repositories. As an alternative to time-consuming, labor-intensive, and destructive radiochemical analysis, the indirect scaling factor (SF) method has been used to determine the concentrations of difficult-to-measure radionuclides. Despite its long history, the original SF methodology remains almost unchanged and now needs to be improved for advanced SF implementation. Intense public attention and interest have been strongly directed to the reliability of the procedures and data regarding repository safety since the first operation of the low- and intermediate-level radioactive waste disposal facility in Gyeongju, Korea. In this review, statistical methodologies for SF implementation are described and evaluated to achieve reasonable and advanced decision-making. The first part of this review begins with an overview of the current status of the scaling factor method and global experiences, including some specific statistical issues associated with SF implementation. In addition, this review aims to extend the applicability of SF to the characterization of large quantities of waste from the decommissioning of nuclear facilities.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.18-24
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• 2020

When a damage occurs in the nuclear fuel burning in the reactor, fission products that should be in the nuclear fuel rod are released into the reactor coolant. In this case, sipping test, a series of non-destructive inspection methods, are used to find leakage in nuclear fuel assemblies during the power plant overhaul period. In addition, the sipping test is also used to check the integrity of the spent fuel for moving to an intermediate dry storage, which is carried out as the first step of nuclear decommissioning, . In this paper, the principle and characteristics of the sipping test are described. The structure of the sipping inspection equipment is largely divided into a suction device that collects fissile material emitted from a damaged assembly and an analysis device that analyzes their nuclides. In order to make good use of the sipping technology, the radioactive level behavior of the primary system coolant and major damage mechanisms in the event of nuclear fuel damage are also introduced. This will be a reference for selecting an appropriate sipping method when dismantling a nuclear power plant in the future.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2736-2747
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• 2022

The operations in the design lifecycle of a nuclear power plant targeted to be decommissioned lead to neutron activation. Operations in the decommissioning process include cutting, decontamination, disposal, and processing. Among these, cutting is done close to the target material, and thus workers are exposed to radiation. As there are only a few studies on pressurizers, there arises the need for further research to assess the radiation exposure dose. This study obtained the specifications of the AP1000 pressurizer of Westinghouse and the distribution of radionuclide inventory of a pressurizer in a pressurised water reactor for evaluation based on literature studies. A cutting scenario was created to develop an optimal method so that the cut pieces fill a radioactive solid waste drum with dimensions 0.571 m × 0.834 m. The estimated exposure dose, estimated using the tool VISIPLAN SW, in terms of the decontamination factor (DF) ranged from DF-0 to DF-100, indicating that DF-90 and DF-100 meet the ICRP recommendation on exposure dose 0.0057 mSv/h. At the end of the study, although flame cutting was considered the most efficient method in terms of cutting speed, laser cutting was the most reasonable one in terms of the financial aspects and secondary waste.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2329-2333
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• 2022

Base metal corrosion forms a significant issue during the chemical decontamination of the primary coolant loop in nuclear power plants as it is directly related to the economic and safety viability of decommissioning. In this technical note, potentiodynamic evaluations of several base metals (304 stainless steel, SA106 Grade B carbon steel, and alloy 600) were performed to determine their corrosion behavior during the hydrazine (N₂H₄)-based reductive ion decontamination (HyBRID) process. The results suggested that N₂H₄ protected the surface of the base metals in the HyBRID solution, which is primarily composed of H₂SO₄. The corrosion resistance of the carbon steel was further improved through the addition of CuSO₄ to the solution. The corrosion rate of carbon steel in the H₂SO₄-N₂H₄-CuSO₄ solution was lower than that exhibited in an oxalic acid solution, a commonly used reaction medium during commercial decontamination processes. These results indicate the superiority of the HyBRID process with respect to the base metal stability.