• Journal of Radiation Industry
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• v.18 no.3
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• pp.255-266
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• 2024

The uranium nuclear fuel used in nuclear power generation needs to be replaced with new fuel after a certain period. In South Korea, the spent nuclear fuel generated during this process is temporarily stored within the nuclear power plant site, and there are ongoing issues with the saturation of storage capacity. To address these problems, the South Korea government has established a plan to manage high-level radioactive waste, including provisions for securing interim storage facilities. An interim storage facility is designed to safely store spent nuclear fuel for certain period before its permanent disposal. This study analyzed leading international cases of interim storage facilities that are technically feasible and can reduce the operating period of temporary storage facilities for spent nuclear fuel within nuclear power plant sites. It also presented the technical concepts required for the operation of interim storage facilities for spent fuel from PWR(Pressurized Water Reactor), reflecting the situation in South Korea.

• The Korean Journal of Nuclear Medicine Technology
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• v.17 no.1
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• pp.3-6
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• 2013

Purpose: The treatment of thyroid cancer patients was continuously increased. According to the increment of thyroid cancer patients, the establishment of iodine therapy site was also increased in each hospital. This treatment involves the administration of radioactive iodine, which will be given in the form of a capsule. Therefore, protections and managements for radioactive source pollution and radiation exposure should be necessary for radiation safety. Among the many problems, the problem of disposing the radioactive wastes was occurred. In this study, The date for self-disposal for radioactive wastes, which were contaminated in clothes, bedclothes and trash, were calculated. Materials and Methods: The number of iodine therapy ward was 15 in Korea Institute of Radiological Medical and Sciences. Recently, 8 therapy wards were operated for iodine therapy patients and others were on standby for emergency treatment ward of any radiation accidents. Radioactive wastes, which were occurred in therapy ward, were clothes, bedclothes, bath cover for patients washing water and food and drink which was leftover by patients. Each sample was hold into the marinelli beaker (clothes, bedclothes, bath covers) and 90 ml beaker (food, drink, and washing water). The activities of collected samples were measured by HpGe MCA device (Multi Channel Analysis, CANBERRA, USA) Results: The storage period for the each kind of radioactive wastes was calculated by equation of storage periods based on the measurement outcomes. The average storage period was 60 days for the case of clothes, and the maximum storage period was 93 days for patient bottoms. The average storage period and the maximum storage period for the trash were 69 days and 97 days, respectively. The leftover foods and drinks had short storage period (the average storage period was 25 days and maximum storage period was 39 days), compared with other wastes. Conclusion: The proper storage period for disposing the radioactive waste (clothes, bedclothes and bath cover) was 100 days by the regulation on self-disposal of radioactive waste. In addition, the storage period for disposing the liquid radioactive waste was 120 days. The current regulation for radioactive waste self-disposing was not suitable for the circumstances of each radioactive therapy facility. Therefore, it was necessary to reduce the leftover food and drinks by adequate table setting for patients, and improve the process and regulation for disposing the short-half life radioactive wastes.

• Journal of Korean society of Dental Hygiene
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• v.20 no.1
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• pp.73-83
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• 2020

Objectives: This study aimed to measure the knowledge of dental hygienists regarding dental waste, and identify the relationship between general characteristics and infection control characteristics. Methods: This study was conducted using a self-reported questionnaire in 250 dental hygienists. The questionnaire consisted of 21 items: storage container (5 items), storage locations (5 items), storage period (9 items), and storage method (3 items). The collected data were analyzed using the t-test and analysis of variance. Results: The education experience of infection management within the last year was reviewed for significant differences among dental hygienists regarding storage container and storage periods (p<0.05). Statistically significant differences were observed regarding the knowledge of storage container, storage locations, storage period, and storage method among the enrolled dental hygienists (p<0.05). Conclusions: This study examined the necessity of developing conservative education and job education programs to improve the knowledge level of dental hygienists regarding dental waste management.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.313-317
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• 2018

Licenses for the storage of spent nuclear fuel (SNF) and vitrified highly active waste in casks under dry conditions are limited to 40 years and have to be renewed for prolonged storage periods. If such a license renewal has to be expected since as in accordance with the new site selection procedure a final repository for spent fuel in Germany will not be available before the year 2050. For transport and possible unloading and loading in new casks for final storage, the integrity and the maintenance of the geometry of the cask's inventory is essential because the SNF rod cladding and the cladding of the vitrified highly active waste are stipulated as a barrier in the storage concept. For SNF, the cladding integrity is ensured currently by limiting the hoop stress and hoop strain as well as the maximum temperature to certain values for a 40-year storage period. For a prolonged storage period, other cladding degradation mechanisms such as inner and outer oxide layer formation, hydrogen pick up, irradiation damages in cladding material crystal structure, helium production from alpha decay, and long-term fission gas release may become leading effects driving degradation mechanisms that have to be discussed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.2
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• pp.205-214
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• 2021

Currently, the interim storage pools of spent fuels in South Korea are expected to become saturated from 2024. It is required to prepare an operation plan of a domestic dry storage facility during a long-term period, with the researches on safety evaluation methods. This study modified the FRAPCON code to predict the spent fuel integrity evaluation such as the axial cladding temperature, the hoop stress and hydrogen distribution in dry storage. The cladding temperature in dry storage was calculated using the COBRA-SFS code with the burnup information which was calculated using the FRAPCON code. The hoop stress was calculated using the ideal gas equation with spent fuel information such as rod internal pressure. Numerical analysis method was used to calculate the degree of hydrogen diffusion according to the hydrogen concentration and temperature distribution during a dry storage period. Before 50 years of dry storage, the cladding temperature and hoop stress decreased rapidly. However, after 50 years, they decreased gradually and the cladding temperature was below 400 K. The initial temperature distribution and hydrogen concentration showed a parabolic line, but hydrogen was transferred by the hydrogen concentration and temperature gradient over time.

• The Korean Journal of Nuclear Medicine Technology
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• v.26 no.1
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• pp.15-26
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• 2022

Purpose Recently, in the process of examining the self-disposal of radioactive waste by the Korea Institute of Nuclear Safety, it is difficult to reach the final approval process for self-disposal. In connection with this, we intend to increase the processing efficiency of self-disposal and strengthen safety by analyzing cases of recent supplementary matters. Materials and Methods From 2018 to 2021, we compare and review a supplementary requests that preparing the procedures and plans for the self-disposal of radioactive waste by 20 institutions. In this regard, based on the provisions of the Atomic Energy Safety Act, we derive a detailed proposals for the self-disposal of radioactive waste by arranging the review processing period calculation and supplementary requests that occurred during the review process. Results The representative supplementary requests of the Korea Institute of Nuclear Safety are the calculation of the storage period by type and nuclide of radioactive waste, the contents of the packaging container, the RASIS reporting method, the planned storage method for self-disposal, confirmation of the final disposal company, and the storage period of the waste filter Calculation, radioactive labeling, etc. And it is emphasized as important. Conclusion The expected effects of the guidelines reflecting the latest supplements include reduction of the time required for document preparation and increase of work processing efficiency, improvement of storage efficiency in the radioactive waste storage room, and economic cost reduction. If the radioactive waste self-disposal guideline presented in this study is applied to the field, it is thought that it will be helpful in improving the work efficiency of those who are experiencing difficulties.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.411-422
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• 2016

Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R&D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0~4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

• Journal of Radiation Protection and Research
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• v.40 no.2
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• pp.92-100
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• 2015

For the purpose of improving the efficiency of the radiation impact assessment of dry interim storage facilities for the spent nuclear fuel of pressurized water reactors (PWRs), radiation impact assessment was performed after the application of sensitivity assessment according to the radiation source term designation method, development of a 2-step calculation technique, and cooling time credit. The present study successively designated radiation source terms in accordance with the cask arrangement order in the shielding building, assessed sensitivity, which affects direct dose, and confirmed that the radiation dosage of the external walls of the shielding building was dominantly affected by the two columns closest to the internal walls. In addition, in the case in which shielding buildings were introduced into storage facilities, the present study established and assessed the 2-step calculation technique, which can reduce the immense computational analysis time. Consequently, results similar to those from existing calculations were derived in approximately half the analysis time. Finally, when radiation source terms were established by adding the storage period of the storage casks successively stored in the storage facilities and the cooling period of the spent nuclear fuel, the radiation dose of the external walls of the buildings was confirmed to be approximately 40% lower than the calculated values; the cooling period was established as being identical. The present study was conducted to improve the efficiency of the Monte Carlo shielding analysis method for radiation impact assessment of interim storage facilities. If reliability is improved through the assessment of more diverse cases, the results of the present study can be used for the design of storage facilities and the establishment of site boundary standards.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.22 no.1
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• pp.1-7
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• 2024

In this research, a detailed analysis of the decay heat contributions of both actinides and non-actinides (fission fragments) from spent nuclear fuel (SNF) was made after 50 GWd·tHM^-1 burnup of fresh uranium fuel with 4.5% enrichment lasted for 1,350 days. The calculations were made for a long storage period of 300 years divided into four sections 1, 10, 100, and 300 years so that we could study the decay heat and physical disposal ratios of radioactive waste in medium- and long-term storage periods. Fresh fuel burnup calculations were made using the code MCNP, while isotopic content and then decay heat were calculated using the built-in stiff equation solver in the MATLAB code. It is noted that only around 12 isotopes contribute more than 90% of the decay heat at all times. It is also noted that the contribution of actinides persists and is the dominant ether despite decreasing decay heat, while the effect of fission products decreases at a very rapid rate after about 40 years of storage.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.113-119
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• 2011

Safe operation and maintenance of engineered dry storage systems for spent fuel from nuclear power plants basically depends on adequately adopted design requirements. The most important design target of the system are those which provide the necessary assurances that spent fuel can be received, handled, stored and retrieved without undue risk to health and safety of workers and the public. To achieve these objectives, the design of the system incorporates features to remove spent fuel residual heat, to provide for radiation protection, and to maintain containment over the lifespan of the system as specified in the design specifications. The features also provide for all possible anticipated operational occurrences and design basis events in accordance with the design basis as guided by the designated regulations. The general performance requirements of a projected storage system are introduced in this paper. The storage system is designed to store fuel assemblies in associated with designated regulatory requirements. Small increases/decreases in maximum burnup can be adjusted with cooling time. These variations are compensated for by a corresponding small site-specific increase/decrease in the design basis-cooling period, as long as the maximum heat load and radioactivity of loaded fuel assemblies are met. Generic design basis events considered for the storage system are summarized. Shielding and radiological requirements along with mechanical and structural are derived in this study.