• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.305-320
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• 2016

The metal waste generated during nuclear power plant decommissioning constitutes a large proportion of the total radioactive waste. This study investigates the current status of domestic and international regulatory requirements for clearance and the clearance experience of domestic institutions. The RESRAD-RECYCLE code was used for analyzing the clearance of the metal wastes generated during actual nuclear power plant decommissioning, and assessment of the exposure dose of twenty-six scenarios was carried out. The evaluation results will be useful in preliminary analysis of clearance and recycling during nuclear power plant decommissioning. As a next step, the effects of reducing disposal costs by clearance can be studied.

• 대한방사선방어학회:학술대회논문집
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• 2011.04a
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• pp.162-163
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• 2011

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.107-111
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• 2011

The release of a site and building from regulatory control is the final stage of the decommissioning process. The MARSSIM (Multi-Agency Radiation Survey and Site Investigation Manual) provides overall framework for conducting data collection for a final status survey to demonstrate compliance with site closure requirements. The KAERI carried out establishing a final status survey by using the guidance provided in the MARSSIM for of a site and building of the Korea Research Reactor. The release criteria for a site and building were set up based on these results of the site specific release levels which were calculated by using RESRAD and RESRAD-Build codes. The survey design for a site and building was classified by using the survey dataset and potential contamination. The number of samples in each survey unit was calculated by through a statistical test using the collected data from a scoping and characterization survey. The results of the final status survey were satisfied the release criteria based on an evaluation of the measured data.

• 대한방사선방어학회:학술대회논문집
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• 2011.11a
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• pp.242-243
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• 2011

• 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 Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.1
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• pp.123-132
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• 2021

Radiological impact analyses were carried out for a near-surface radioactive waste repository at Gyeongju in South Korea. The RESRAD-ONSITE code was applied for the estimation of maximum exposure doses by considering various exposure pathways based on a land area of 2,500 ㎡ with a 0.15 m thick contamination zone. Typical influencing input parameters such as shield depth, shield materials' density, and shield erosion rate were examined for a sensitivity analysis. Then both residential farmer and industrial worker scenarios were used for the estimation of maximum exposure doses depending on exposure duration. The radiation dose evaluation results showed that 60Co, 137Cs, and 63Ni were major contributors to the total exposure dose compared with other radionuclides. Furthermore, the total exposure dose from ingestion (plant, meat, and milk) of the contaminated plants was more significant than those assessed for inhalation, with maximum values of 5.5×10-4 mSv·yr-1 for the plant ingestion. Thus the results of this study can be applied for determining near-surface radioactive waste repository conditions and providing quantitative analysis methods using RESRAD-ONSITE code for the safety assessment of disposing radioactive materials including decommissioning wastes to protect human health and the environment.

• Journal of radiological science and technology
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• v.41 no.1
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• pp.67-73
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• 2018

In this paper, risk assessment was conducted to verify self - disposal requirements by landfill for exempted incineration ash by using Resrad Ver.6.5 computer code. The result of risk assessment by landfill for the incineration by-product is that individual dose is $6.91{\times}10^{-2}{\mu}Sv\;y-1$ and collective dose is $3.475{\times}10^{-7}man-Sv\;y-1$. It proved that the result meets reference dose of individual dose $10{\mu}Sv\;y-1$ and collective dose 1 man-Sv y-1 for general public. According to the current 'Nuclear Safety Commission Notice [No. 2014-3]', it states that the exempted wastes can be disposed of by incineration, landfill and recycling. However, most of recently documents and papers related to exempted wastes are disposed of by landfill and recyling and it could not confirm the case of exempt by incineration. If the national consensus is derived and treating the waste by using process of incineration is activated, it could be considered to treat low level of radiation wastewater and activated carbon excluded from exempted waste because of nuclide $^3H$ and $^{14}C$.

• International Journal of Contents
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• v.16 no.2
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• pp.102-108
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• 2020

The purpose of this study was to estimate and analyze the potential radiation dose that the future visitors and the cleaning staff will be exposed to when the KRR-1 reactor is converted into a memorial hall. The radiation doses were estimated using the RESRAD-BUILD software, where case, building, receptor, shielding, and source parameters were applied as the input data. Also, the basic data for the assessment of the radiation doses were determined in an indirect manner using the data on the waste generated during the decommissioning process of the reactor. The assessment results indicate that the potential radiation dose to the visitors and the cleaning staff will be less than 1 mSv, the annual dose limit for the general public. However, if anyone for a significant period of time is close to the reactor, the overall dose will increase. The radiation dose for the future visitors and the cleaning staff was determined to be lower than the annual dose limit for the general public. Given such a risk, systematic measures, such as periodic monitoring or limiting hours, are imperative.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1231-1242
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• 2020

The Kori Unit 1 will be decommissioned after a permanent shutdown in June 2017. South Korea has a 0.1 mSv/yr exposure limit standard for limited or unlimited site release. This is South Korea's first commercial NPP; therefore, if the containment building is reused as a memorial hall, it will contribute to the improvement of public understanding and enhance the public's acceptance of NPPs. Also, existing Kori Unit 1 nuclear power plant manpower resources can be reused after decommissioning and resident staff and memorial hall visitors can activate nearby commercial areas. Therefore, such a reuse scenario may also prevent an economic recession. The exposure dose was calculated using the following scenarios: worker in the containment building, visitor in the containment building, and worker in buildings other than the containment building. The exposure dose in the buildings was calculated by the RESRAD-BUILD developed by the Argonne National Laboratory (ANL). The preliminary exposure dose and derived concentration guideline level (DCGL) were derived.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3559-3570
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• 2023

In the decommissioning of nuclear facilities, Derived Concentration Guideline Level (DCGL) derivation is necessary for the release of the facility after the site remediation, which also needs to be implemented in the stage of establishing a decommissioning planning. In order to derive DCGL, the dose assessment for the receptors can be conducted from residual radioactivity by using RESRAD code. When performing sensitivity analysis on probabilistic parameters, secondary evaluation is performed by assigning a single value for parameters classified as sensitive. However, several options may arise in the handling of nonsensitive parameters. Therefore, we compared the results of the first execution of RESRAD applying probabilistic parameters for each scenario with the results of the second execution applying a single value to sensitive parameters among the probabilistic parameters. In addition, we analyzed the effect of setting options for non-sensitive parameters. As a result, the effect on DCGL were different depending on the application scenario, the target radionuclides, and the input parameter selections. In terms of the overall evaluation period, the DCGL graph of the default option was generally shown as the most conservative except for some radionuclides. However, it will not necessarily be given priority in the aspect of the need to reflect site characteristics. The reason for selecting a probabilistic parameter is the availability of the parameter and the uncertainty of applying a single value. Therefore, as an alternative, it can be consistently applied to distribution as an option for non-sensitive parameters after sensitivity analysis.