• 방사성폐기물학회지
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• 제19권1호
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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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• 제56권6호
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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.

• 대한방사선기술학회지:방사선기술과학
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• 제41권1호
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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 $^{14}C$.

• 방사성폐기물학회지
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• 제15권4호
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• pp.369-380
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• 2017

중 저준위방사성폐기물 표층처분시설 인간침입시나리오의 '평가/해석에 대한 불확실성'의 관리를 위해 GENII를 이용한 평가결과를 오염토양에 대한 방사선영향평가를 위해 개발된 RESRAD를 이용하여 검증하였다. 중저준위방사성폐기물 표층처분시설의 인간침입시나리오로 시추후거주시나리오를 선정하여 각 코드의 현상 모사에서 발생하는 한계점을 파악하고 동일한 입력데이터 조건에서 두 코드의 평가결과를 비교분석함으로써 모델링의 불확실성을 분석하였다. 평가결과 각 코드에서 일부 핵종의 거동모사에 대한 차이는 있었으나 폐쇄후관리기간 이후 선량평가 결과 모든 피폭경로에 대한 경향이 유사함을 확인하였다. 또한 RESRAD에서 확인한 선량평가 결과를 바탕으로 입력인자에 대한 민감도 분석을 수행하고 주요입력인자를 도출하였다. 이를 통해 모델링 결과 및 입력인자에 대한 불확실성을 분석하고 안전성평가 결과에 대한 신뢰성을 확인하였다. 본 연구의 결과는 중저준위방사성폐기물 처분시설의 Safety Case 구축에 활용될 수 있다.

• 방사성폐기물학회지
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• 제14권4호
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• pp.305-320
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• 2016

원자력발전소 해체 시 발생하는 금속폐기물은 폐기물 중에서 많은 비중을 차지하고 있다. 본 연구에서는 국내 자체처분 규제 요건 및 국내 기관별 자체처분현황을 조사하였다. 실제 원자력발전소 해체 시 발생되는 금속폐기물의 자체처분을 위하여 RESRAD-RECYCLE 코드를 이용하였으며 26가지 시나리오에 대한 선량평가를 수행하였다. 평가결과는 원자력발전소 해체 시 자체처분 및 재활용에 관한 사전자료로서 활용가치가 있을 것으로 사료된다. 추후 자체처분을 통한 처분비용 저감효과 연구가 추가로 가능할 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제54권5호
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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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• 제51권8호
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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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• 제55권10호
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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.

• 방사성폐기물학회지
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• 제18권3호
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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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• 제50권8호
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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).