• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.397-410
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• 2023

This study examined the efficacy of various chlorinating agents in partitioning light water reactor spent fuel, with the aim of optimizing the chlorination process. Through thermodynamic equilibrium calculations, we assessed the outcomes of employing MgCl₂, NH₄Cl, and Cl₂ as chlorinating agents. A comparison was drawn between using a single agent and a sequential approach involving all three agents (MgCl₂, NH₄Cl, and Cl₂). Following heat treatment, the utilization of MgCl₂ as the sole chlorinating agent resulted in a moderate separation. Specifically, this method yielded a solid separation with 96.9% mass retention, 31.7% radioactivity, and 44.2% decay heat, relative to the initial spent fuel. In contrast, the sequential application of the chlorinating agents following heat treatment led to a final solid separation characterized by 93.1% mass retention, 5.1% radioactivity, and 15.4% decay heat, relative to the original spent fuel. The findings underscore the potential effectiveness of a sequential chlorination strategy for partitioning spent fuel. This approach holds promise as a standalone technique or as a complementary process alongside other partitioning processes such as pyroprocessing. Overall, our findings contribute to the advancement of spent fuel management strategies.

• Nuclear Engineering and Technology
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• v.9 no.4
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• pp.211-222
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• 1977

The decomposition of all the individual chemicals used in the Harwell inactive vitrification pilot plant has been studied by means of a thermal balance. Weight loss curves to 110$0^{\circ}C$ have been obtained. The four materials (sodium nitrate, cesium nitrate, lithium nitrate and ruthenium nitroso-nitrate solution) showed a greater weight loss than that based on an oxide yield, and hence these compounds of their products of decomposition are volatile below 110$0^{\circ}C$. The remaining materials suffered a weight loss no more than that corresponding to a full yield of the oxide, and hence they were not volatile below 110$0^{\circ}C$. Most of chemicals begin to decompose at less than 75$^{\circ}C$ but the nitrates of cesium, strontium, barium and sodium not until 295$^{\circ}C$ to 59$0^{\circ}C$. The results obtained can be used in the analysis of process conditions in the vitrification and calcination of highly radioactive wastes and also of the thermal decomposition behaviour of mixtures containing those materials.

• Journal of Radiation Industry
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• v.18 no.1
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• pp.43-52
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• 2024

Currently, there are no interim storage facilities and permanent disposal facilities in Korea, so all spent nuclear fuels are temporarily stored. However, the temporary storage facility is approaching saturation, and as a measure to this, the 2nd Basic Plan for the Management of High-Level Radioactive Waste presented an operation plan for dry interim storage facilities and dry temporary storage facilities on the NPP on-site. The dry storage can be operated in various ways, and to select the optimal dry storage method, the reduction of exposure for workers must be considered. Accordingly, it is necessary to develop a worker exposure scenario according to the dry storage method and evaluate and compare the radiological impact for each method. The purpose of this study is to develop an exposure scenario for workers transporting spent nuclear fuel by dry storage method. To this end, first, the operation procedure of the foreign commercial spent nuclear fuel dry storage system was analyzed based on the Final Safety Analysis Report (FSAR). 1) the concrete overpack-based system, 2) the metal overpack-based system, and 3) the vertical storage module-based system were selected for analysis. Factors were assumed that could affect the type of work (working distance, working hours, number of workers, etc.) during transportation work. Finally, the work type of the processes involved in transporting spent nuclear fuel by dry storage method was set, and an exposure scenario was developed accordingly. The concrete overpack method, the metal overpack method, and the vertical storage module method were classified into a total of 31, 9, and 23 processes, respectively. The work distance, work time, and number of workers for each process were set. The product of working hours and number of workers (Man-hour) was set high in the order of concrete overpack method, vertical storage module method, and metal overpack method, and short-range work (10 cm) was most often applied to the concrete overpack method. The results of this study are expected to be used as basic data for performing radiological comparisons of transport workers by dry storage method of spent nuclear fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.161-172
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• 2017

To evaluate radiological impact from the operation of a low- and intermediate-level radioactive waste disposal facility, a logical presentation and explanation of expected accidental scenarios is essential to the stakeholders of the disposal facility. The logical assessment platform and procedure, including analysis of the safety function of disposal components, operational hazard analysis, operational risk analysis, and preparedness of remedial measures for operational safety, are improved in this study. In the operational risk analysis, both design measures and management measures are suggested to make it possible to connect among design, operation, and safety assessment within the same assessment platform. For the preparedness of logical assessment procedure, classification logic of an operational accident is suggested based on the probability of occurrence and consequences of assessment results. The improved assessment platform and procedure are applied to an operational accident analysis of the Korean low- and intermediate-level radioactive waste disposal facility and partly presented in this paper.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.369-380
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• 2017

In order to manage the uncertainty about the evaluation and analysis of the human intrusion scenario of the Gyeongju Low and Intermediate Level Radioactive Waste(LILW) disposal facility, the calculation result by the GENII code was assessed using the RESRAD code, which was developed to evaluate the radiation effects of contaminated soil. The post-drilling scenario was selected as a human intrusion scenario into the near-surface disposal facility to analyze the uncertainty of the modeling by identifying any limitations in the simulation of each code and comparing the evaluation results under the same input data conditions. The results revealed a difference in the migration of some nuclides between the codes, but confirmed that the dose trends at the end of the post-closure control period were similar for all exposure pathways. Based on the results of the dose evaluation predicted by RESRAD, sensitivity analysis on the input factors was performed and major input factors were derived. The uncertainty of the modeling results and the input factors were analyzed and the reliability of the safety evaluation results was confirmed. The results of this study can be applied to the implementation 'Safety Case Program' for the Gyeongju LILW disposal facility.

• Journal of Radiation Protection and Research
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• v.9 no.2
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• pp.118-128
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• 1984

The analysis of radiological protection at Kori Nuclear Unit #1, which started commercial operation on April 29, 1978, has been carried out in three areas, namely, radiation exposure, radioactive waste disposal and environmental management. The analysis results for radiation exposure and radioactive effluent releases appear to increase gradually with plant operating years. On the other hand, the apex of radiation exposure and radioactive effluent releases was in 1983 and 1982 respectively and then decreased or stabilized. Also, the operation environmental radioactivity seems to be no higher than preoperation environmental radioactivity.

• Journal of the Korean Society of Radiology
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• v.12 no.5
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• pp.565-573
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• 2018

Cyclotron for medical purposes generates nuclear reaction by accelerating protons in high speed, in order to produce radiopharmaceuticals, and unnecessary neutrons are generated through such nuclear reaction. Neutrons cause activation in the parts of cyclotron which then cause exposure to radiation for people working in the field. This study, in that regard, aims to analyze exposure level by finding out the degree of activation of aluminum body, silver body, and havar foil which are the parts of Targetry where the nuclear reaction takes place. The results of the experiment showed that aluminum body and silver body had no problems re-using them as the energy and half-life of activated nuclides were small and short, making the affect on the people working in the field extremely low. However for havar foil, its activated nuclides had a high level of energy which resulted in high level of affect to the people working in the field. The activation factors of the cyclotron were analyzed, and the results showed that the Havar foil was activated the most among the targetry parts, and greatly exposed workers due to regular replacement, and needed special management as radioactive waste.

• Journal of Radiation Protection and Research
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• v.22 no.3
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• pp.183-193
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• 1997

The amount and kinds of radionuclide contained in waste volume should be known to prepare for occupational exposure management, perform safety assessment and finally to license a repository. Although the volume of filters and resins are small, activities of them comprise most of the radioactivity that made during power generation. This study aims at developing a method of estimating the radionuclide accumulation at the filters and resins of coolant systems. In this study, accumulated amount of radionuclides is estimated by a computer program which makes use of instantaneous decontamination factor, DF, instead of average DF. A FORTRAN program was developed for the estimation. Data from in-plant source-term measurements at Rancho-Seco nuclear power plant in the United States are employed for verification of the estimating method. And experimental data are employed, too. The instantaneous-DF-method showed smaller error than the average-DF-method. Accumulated amount of radionuclides can be calculated with only the DF and the radionuclide concentration, which are measured periodically according to the operating guide. However, especially, when the operating condition of nuclear power plant changes rapidly, the measuring term of DF and radionuclide should be shortened to ensure the accurate estimation.

• Journal of Radiation Protection and Research
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• v.25 no.2
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• pp.89-96
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• 2000

Nuclear fuel cycle choices and costs are important in considering energy policies, fuel diversity, security of supply and associated social and environmental impacts. Particularly, the nuclear spent fuel is very important in view of high activity and the need of long term management. This study focuses on the comparison of reprocessing and direct disposal of nuclear spent fuel in terms of cost, safety and public acceptability. The results of the study show that the direct disposal is about 7% more economical than the reprocessing. In terms of safety, the results show that the risk of vitrified HLW (high-level radioactive waste) is less than directly disposed spent fuel. For the public acceptability, both of the methods are not well understood and therefore they are not accepted. In conclusion, it is necessary to guarantee the safety of the both spent fuel processing methods through continuous development of associated technology and to have a fuel cycle policy which should consider not only the economics but also social and environmental impacts.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.157-167
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• 2020

Removing radioactive concrete is crucial in the decommissioning of nuclear power plants. However, this process generates radioactive aerosols, exposing workers to radiation. Although large amounts of radioactive concrete are generated during decommissioning, studies on the internal exposure of workers to radioactive aerosols generated from the cutting of radioactive concrete are very limited. In this study, therefore, we calculate the internal radiation doses of workers exposed to radioactive aerosols during activities such as drilling and cutting of radioactive concrete, using previous research data. The electrical-mobility-equivalent diameter measured in a previous study was converted to aerodynamic diameter using the Newton-Raphson method. Furthermore, the specific activity of each nuclide in radioactive concrete 10 years after nuclear power plants are shut down was calculated using the ORIGEN code. Eventually, we calculated the committed effective dose for each nuclide using the IMBA software. The maximum effective dose of ¹⁵²Eu constituted 83.09% of the total dose; moreover, the five highest-ranked elements (¹⁵²Eu, ¹⁵⁴Eu, ⁶⁰Co, ²³⁹Pu, ⁵⁵Fe) constituted 99.63%. Therefore, we postulate that these major elements could be measured first for rapid radiation exposure management of workers involved in decommissioning of nuclear power plants, even if all radioactive elements in concrete are not considered.