• 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.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2274-2281
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• 2024

RESRAD-ONSITE v7.2 code is used to assess the radiation effects on a farmer resident located in a hypothetical contaminated site planned for the first nuclear power plant project in Vietnam, namely Ninh Thuan 1, after decommissioning. Derived concentration guideline levels are preliminarily calculated for 17 radionuclides that are assumed to remain on a contaminated surface soil with an initial concentration of 1 pCi/g in the protected area of NPP site. For a reliable estimation, the site-specific conditions regarding the geological, hydrological, climate, and occupancy data gathered from the Feasibility Study Report (FSR) and relevant literatures for the Ninh Thuan 1 NPP site is employed as input parameters. The calculation results indicate that the peak of total exposure dose is estimated to be ca. 0.191 mSv/yr at the time of decommissioning, and then decrease over time. Furthermore, the protected site is assessed to be released at ca. 6.71 years after decommissioning under the regulation on radiation protection in Vietnam. Through this study, a radiation exposure model for residents living near the Ninh Thuan 1 NPP is preliminarily established by using the RESRAD-ONSITE code, which are expected to be useful for future implementation of the Ninh Thuan 1 NPP project in Vietnam.

• 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 $^{14}C$.

• 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.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.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1781-1788
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• 2022

The containment building Kori Unit 1 may require sequential steps for full decommissioning. This study assumes that the containment building is to be used as an auxiliary building that handles nuclear power systems and materials during decommissioning before conversion into a greenfield. Through the derivation of guidelines and dose evaluation, it was confirmed whether the radiation workers were satisfied with the ALARA decision. The specific modeling of the external radiation exposure was performed based on the facility investigation procedures. The external radiation specific derived concentration guideline levels (DCGLs) for radiation workers in containment building were obtained using the RESRAD-BUILD code and were applied to the VISIPLAN 3D ALARA Planning Tool code to calculate the working dose and check worker safety. The derivation of site-specific and realistic DCGLs and dose evaluation via 3D modeling can contribute to the scenario development for the decommission and remediation of containment building.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1951-1956
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• 2019

As a part of the preliminary decommissioning plan of KEPCO-NF fuel fabrication facility, DCGLs of three target radionuclides, ²³⁴U, ²³⁵U, and ²³⁸U, were derived using RESRAD-BUILD code and contaminated areas of the facility were classified based on contamination levels from the derived DCGLs. From code simulations, one-room modeling results showed that the grinding room in building #2 was the most restrictive (DCGL_gross = 10493.01 Bq/㎡). The DCGL_gross results in contaminated areas from one-room modeling were slightly more conservative than three-room modeling. Prior to the code simulation, field survey and measurements conducted by each survey unit. For a conservative approach, the most restrictive DCGL_gross in each survey unit was taken as a reference to classify the contaminated areas of the facility. Accordingly, seven rooms and 37 rooms in the nuclear-fuel buildings were classified as Class 1 and Class 2, respectively. As expected, fuel material handling and processing rooms such as the grinding room, sintering room, compressing room, and powder collecting room were included in the Class 1 area.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.3
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• pp.347-354
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• 2020

Site characterization for decommissioning Kori Unit 1 is ongoing in South Korea after 40 years of successful operation. Kori Unit 1's containment building is assumed to be mostly radioactively contaminated, and therefore radiation exposure management and detailed contamination investigation are required for decommissioning and dismantling it safely. In this study, site-specific Derived Concentration Guideline Levels (DCGLs) were derived using the residual radioactivity risk evaluation tool, RESRAD-BUILD code. A conceptual model of containment building for Kori Unit 1 was set up and limited occupational worker building inspection scenario was applied. Depending on the source location, the maximum contribution source and exposure pathway of each radionuclide were analyzed. The contribution of radionuclides to dose and exposure pathways, by source location, is expected to serve as basic data in the assessment criteria of survey areas and classification of impact areas during further decommissioning and decontamination of sites.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1289-1297
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• 2018

Preliminary surface soil Derived Concentration Guideline Levels (DCGLs) were derived conforming to the Multi-Agency Radiation Site Survey and Investigation Manual (MARSSIM) procedure for the site release and reuse of Kori Unit 1 in Korea. Based on the decommissioning experiences of the U.S. nuclear power plants, a suite of residual radionuclides was determined, and uncertainties contributed to the resultant dose by the input parameters were quantified via the sensitivity analysis of parameters. The peak of the mean dose was obtained via the probabilistic analysis of the RESRAD (RESidual RADioactivity)-ONSITE code. Consequently, $DCGL_w$ of Kori Unit 1 in accordance with two scenarios, industrial worker and residential farmer scenario, were derived and the results were compared respectively with other NPPs. It could be used as a basic guideline for establishing regulatory standards for reuse planning, designing the site characterization surveys and implementing final status survey (FSS).