Derived Concentration Guideline Levels Research Articles

Evaluation of derived concentration guideline levels reflecting the site-specific data for the soil of the Korea research reactor unit 1 and 2

The purpose of this study is to evaluate the derived concentration guideline levels for unrestricted site reuse the Korea research reactor unit 1 and 2. Distribution coefficients for Co-60 and Sr-90 were derived, and site-specific values of the KRR soil were applied for the DCGLs for the seven target nuclide. The distribution coefficients of Co-60 and Sr-90 were 6,128 and 86.0 mL/g. The DCGLs derived from the dose by age group were 0.053 Bq/g for Co-60 and 45.0 Bq/g for H-3.

Preliminary assessment of derived concentration guideline level (DCGL) for a hypothetical contaminated site planned for Ninh Thuan 1 nuclear power plant project in Vietnam by using RESRAD-ONSITE code

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.

The effect of sensitive and non-sensitive parameters on DCGL in probability analysis for decommissioning of nuclear facilities

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 non-sensitive 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.

Derivation of site-specific derived concentration guideline levels at Korea Research Reactor 1 and 2 sites using probabilistic analysis

The purpose of this study is to apply a probabilistic method to derive the derived concentration guideline levels for decommissioning of Korea Research Reactor 1 and 2. A total of seven parameters were found to be the sensitive parameters of the target nuclides. The DCGLs of Co-60 and H-3 were 0.063 Bq/g and 85.470 Bq/g, respectively. The concentrations of the gamma ray-emitting nuclides in the actual reactor sites were 7.7–215 times lower than the derived DCGLs for gamma ray-emitting nuclides.

Preliminary Evaluation of Derived Concentration Guideline Level for Surface Soil at Wolsong NPP Site Using RESRAD-ONSITE Code

A preliminary derived concentration guideline level (DCGL) was calculated for the soil of the Wolsong nuclear power plant (NPP) (the first commercial pressurized heavy water reactor in South Korea) site using the RESRAD-ONSITE computational code. In total, fourteen selected radionuclides were analyzed after considering the preliminary evaluation information of radionuclides observed in the pressure tube specimen in the Wolsong Unit 1 heavy water reactor and previous NPP decommissioning cases. Furthermore, a geological structure model of the Wolsong NPP site was established according to the safety analysis report of the Wolsong NPP. In addition, the distribution coefficients (Kd) of various radionuclides were derived using the JAEA-SDB and the pH information of the groundwater around the Wolsong NPP site. The DCGL for surface soil of the Wolsong NPP site was derived via the application of criteria for the site release to facilitate unrestricted reuse. Moreover, preliminary dose evaluation and relevant analysis were performed according to the Wolsong NPP site resident scenario. The novelty of this study lies in the first calculation of the preliminary DCGL values for the case of the pressurized heavy water reactor (Wolsong NPP) site. It is expected that further reliable DCGL results might be achievable if more precise radionuclide information and site-specific parameters with respect to the Wolsong NPP site are secured and applied in the future.

External exposure specific analysis for radiation worker in reuse of containment building for Kori Unit 1

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.

Derivation of site-specific derived concentration guideline levels at Korea Research Reactor-1&2 sites

The objective of this study was to derive derived concentration guideline levels (DCGLs) reflecting the site-specific characteristics of KRR-1&2. A total of 7 nuclides (H-3, C-14, Co-60, Sr-90, Cs-137, Eu-152, and Eu-154) were selected for DCGLs derivation. Radiation dose at the sites was evaluated with RESRAD-ONSITE program. The dose contribution due to direct external exposure was the highest during the entire evaluation period. Ingestion had the second effect. The DCGLs of Co-60 was derived to be 0.051 Bq/g, and DCGLs of Cs-137 was 0.193 Bq/g. The DCGLs of H-3 showed the highest value of 129 Bq/g. The ratio of DCGLs derived by applying site-specific values and default values ranged from 0.27 to 19.6. For six nuclides excluding H-3, KRR-1&2 sites and the overseas NPP sites showed similar DCGLs. H-3 showed large differences in DCGLs from this study and overseas NPPs. The large difference resulted from input parameter values applied to the sites. In conclusion, it is critical to apply site-specific parameter values reflecting the site characteristics to derive DCGLs for decommissioned site clearance. The result of this study can be used as a reference for nuclide selection and DCGLs derivation reflecting the site characteristics when decommissioning nuclear facilities, including nuclear power plants in Korea.

Preliminary DCGLW for surface soil of Kori Unit-1 decommissioning site according to habit data and age of human receptors

The preliminary derived concentration guideline level (DCGLW) for the surface soil of Kori Unit-1 was calculated considering the habit data of residents living near the nuclear power plant in South Korea (a 1-y-old infant, a 10-y-old child, and an adult), following the ICRP 101a recommendation for radiation protection of the public. The radionuclides of interest were 3H, 14C, 60Co, 63Ni, 90Sr, 134Cs, 137Cs, and 241Pu. 14C, 3H. 14C, 3H, and 241Pu had the maximum total effective dose equivalent (TEDE) after a certain duration, whereas 60Co, 63Ni, 134Cs, 137Cs, and 90Sr had the maximum TEDE in the first year. In the initial period, 60Co, 134Cs, 137Cs, and 90Sr had the most dominant effect on TEDE in all the age groups, but their influence disappeared after 203.09 y and only doses of 241Pu and 90Sr were observed thereafter. The main exposure pathways of 60Co, 134Cs, and 137Cs were external pathways, but the main exposure pathways of 90Sr were plant (water-independent), meat (water-independent), and milk (water-independent). The maximum TEDEs caused by 60Co, 134Cs, and 137Cs were similar in all the age groups, but that caused by 90Sr considerably varied depending on the age group. The nuclides for which the DCGLW was notably different from the conservative DCGLW were as follows: 3H (increment of 109.21% and 183.08% for the child and adult, respectively), 63Ni (increment of 286.37% and 702.86% for the child and adult, respectively). The main exposure pathways for these radionuclides were ingestion pathways (water-dependent or independent), which led to internal exposure. The findings can provide insight for the considerations of the human receptor to calculate the DCGLW for the surface soil of Kori Unit-1 using RESRAD-ONSITE.

Calculation approach to building DCGLs of waste treatment facilility for Kori unit 1 decommissioning and comparative analysis of the results

The Derived Concentration Guideline Levels (DCGLs) of remaining buildings in decommissioning site realistically need to be considered because all nuclear power plants in Korea are operated with the adjacent plant in multi-unit site. In addition, a few buildings are expected to be utilized for the decommissioning such as waste treatment facility to reduce the final disposal volume of radioactive waste. Therefore, DCGLs of buildings postulated for the decommissioning of Kori site have been assessed by using RESRAD-BUILD code. As a result, DCGLs were derived, reflecting the site reuse criteria and input prameters based on site conditions and assumptions, and compared and analyzed how they differ from DCGLs of other nuclear plants depending on radionuclides. The occupancy and renovation scenarios were applied and more than half of the target radionuclides were conservatilvely evaluated in the occupancy scenarion, while some were evaluated in the renovation scenario. The approach and methodology presented in this paper, which can be considered at the level of the transition phase, can be utilized for deriving results that reflect reality at the time of license termination.

Radionuclide-Specific Exposure Pathway Analysis of Kori Unit 1 Containment Building Surface

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.

Comparative Analysis of Probabilistic Analysis and Deterministic Analysis by RESRAD Code

The decommissioning of nuclear facilities indicates that the site is finally released according to a limited or unlimited site reuse purpose. In this process, the assessment of exposure dose to decommissioning workers and nearby residents is essential. Based on MARSSIM, a widely used decommissioning guideline in the United States, derivation of the exposure dose and derived concentration guideline level (DCGL) is mandatory using the probabilistic analysis of the RESRAD code. Here, DCGL is the radionuclide-specific concentration that satisfies the site release criteria. By applying the priority 1 parameter, which has the greatest effect on the dose, the dose is derived through deterministic and probabilistic analyses. The results were compared and analyzed. The purpose of this study was to provide a basic database that can be applied to the development of parameter lists and distributions suitable for the characteristics of nuclear facilities in South Korea. In addition, the process of deriving the dose by applying the deterministic and probabilistic analyses of RESRAD was assessed.

Study on the soil evaluation methodology of on and offsite Kori Unit 1

To optimize site remediation and reuse, we evaluated contamination of the soil surface and sub-surface in situ, inside and outside of Kori Unit 1. Realistic dose modeling was performed using RESRAD-OFFSITE code, accounting for air dispersion, groundwater transport, and the positional movement of secondary contamination. The Derived Concentration Guideline Level (DCGL) was derived; this is a useful guideline value and exposure factor for estimating maximum groundwater concentration. Site-specific parameters were used to analyze the maximum contributed exposure pathways in and out of the site, and preliminary dose assessment was performed for offsite ‘Area 1’. This study provides a general guideline for soil evaluation of nuclear power plants decommissioning site, from surface and subsurface soil contamination to the establishment of a basement fill model. It may be used as basic site evaluation methodology and guideline values based on this characterization of Kori Unit 1.

Derivation of preliminary derived concentration guideline level (DCGL) by reuse scenario for Kori Unit 1 using RESRAD-BUILD

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.

Preliminary surface soil area factor for elevated residual radioactivity of Kori Unit 1 considering adjacent Unit 2

Based on the U.S. standard decommissioning procedure guidance manual, Multi-Agency Radiation Site Survey and Investigation Manual (MARSSIM), DCGL (Derived Concentration Guideline Level) can be used to classify contaminated sites and suggests a methodology for assessing site deregulation safety as an essential criterion for determining decontamination activities. In this study, area factors for DCGLEMC, which is DCGL applicable in hot spot, were derived by adjusting the contaminated area at the Kori Unit 1, the first commercial Nuclear Power Plant (NPP) in South Korea. DCGLW and DCGLEMC were investigated for Kori Unit 1 according to changes in preliminary contaminated baseline area including the area of Unit 2. Area factors for surface soil were divided into scenarios, gamma and beta emitters. Sensitive parameters and exposure pathways were identified for each specific contaminated area. Additionally, the trend of DCGLEMC was analyzed in the industrial worker scenario due to baseline area and contaminated area changes.

Scenario options to calculation of Derived Concentration Guideline Levels for a multi-unit decommissioning site

To release a site after nuclear power plant decommissioning, it is necessary to demonstrate compliance with a radiological requirement that the residual radioactivity in a decommissioned site does not result in dose in excess of the site release criteria applicable. This is generally accomplished by ensuring that the residual concentrations of radionuclides are below Derived Concentration Guideline Levels (DCGLs) which are derived, using RESRAD dose modeling code, from the site release criteria under a site reuse scenario assumed for the site. In this paper, we propose a methodology to calculate the preliminary DCGLs for a multi-unit decommissioning site. For this, five site reuse scenario options were applied taking into account the most foreseeable and practical future site reuse conditions at Kori site, especially the operation and decommissioning schedules of other units: Option 1 of a single residential farmer scenario with unrestricted condition, Option 2 of a single industrial worker scenario with unrestricted condition, Option 3 of a hybrid industrial worker plus residential farmer scenario with unrestricted condition, Option 4 of a single nuclear facility scenario with restricted condition, and Option 5 of a hybrid nuclear facility plus residential farmer scenario with restricted condition. For restricted cases (Options 4 and 5), a case where institutional controls of the original scenario fail is also considered by applying residential farmer scenario instead. We hope that the approach to applying multiple site reuse scenario options to calculate DCGLs, proposed in this paper, will give a valuable insight to other related researcher who need to calculate DCGLs, especially for a multi-units site.

Calculation of preliminary site-specific DCGLs for nuclear power plant decommissioning using hybrid scenarios

Abstract Korea's first commercial nuclear power plant at Kori site was permanently shut down in 2017 and is currently in transition stage. Preparatory activities for decommissioning such as historical site assessment, characterization, and dismantling design are being actively carried out for successful D&D (Dismantling and Decontamination) at Kori site. The ultimate goal of decommissioning will be to ensure the safety of workers and residents that may arise during the decommissioning of nuclear facilities and, thereby finally returning the site to its original status in accordance with the release criteria. Upon completion of decommissioning, the resident's safety at a site released will be assessed from the evaluation of dose caused by radionuclides expected to be present or detected at the site. Although the U.S. commercial nuclear power plants with decommissioning experience use different site release criteria, most of them are 0.25 mSv/y. In Korea, both the unrestricted and restricted release criteria have been set to 0.1 mSv/y by the Nuclear Safety and Security Commission. However, since the dose is difficult to measure, measurable concentration guideline levels for residual radionuclides that result in dose equivalent to the site release criteria should be derived. For this derivation, site reuse scenario, selection of potential radionuclides, and systematic methodology should be developed in planning stage of Kori site decommissioning. In this paper, for calculation of a preliminary site-specific Derived Concentration Guideline Levels (DCGLs) for the Nuclear Power Plant site, a novel approach has been developed which can fully reflect practical reuse plans of the Kori site by taking into account multiple site reuse scenarios sequentially, thereby striking a remarkable distinction with conventional approaches which considers only a single site scenario.

Derivation of preliminary derived concentration guideline levels for surface soil at Kori Unit 1 by RESRAD probabilistic analysis

Abstract 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, DCGLw 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).

국내 원전 부지 해제 기준 준수 입증을 위한 미국의 유도농도기준(DCGL) 설정 방법에 대한 분석

원전 해체 이후 원전 부지의 제한적 또는 무제한적 이용에 대해서 미국 NRC는 NUREG-1757 문서를 통해 제한적 또는 무제한적 부지 이용에 관한 방사선학적 기준을 제시하였고 사업자가 제염 및 복원 후 이 선량 기준이 충족됨을 증명할 수 있어야만 부지가 제한적 또는 무제한적으로 해제될 수 있다고 하였다. 이와 관련하여 NRC는 운영허가종료계획서(LTP; License Termination Plan)에 방사선학적 부지 해제 기준 준수를 입증하기 위하여 부지 해제 기준, 부지 특성 평가, 최종 방사선 조사 계획에서 주요 방사선원항, 유도농도기준(DCGL) 등을 기재하도록 하고 있다. 이 논문은 국내 원전 해체에 있어서 참조사례가 될 수 있는 Rancho Seco 원전 해체 사례를 참고 및 절차를 분석함으로써 2017년 영구정지가 예정된 고리 1호기뿐만 아니라 향후 해체 원전 부지의 해제 기준 마련에 있어 사용될 수 있는 방법을 검토하였다. The U.S. NRC establishes a radiological criteria with regard to restricted or unrestricted use of nuclear plant site after decommissioning in NUREG-1757. According to this, a nuclear plant site can be released in a restricted way or unrestricted way only if a licensee demonstrates that the dose criteria is fulfilled after the site decontamination and remediation. In order to prove compliance with the radiological criteria of site release, LTP(License Termination Plan) must include the site release criteria, site characterization, final survey plan with major radionuclides and DCGL(Derived Concentration Guideline Levels), etc. Based on the decommissioning case of Rancho Seco nuclear power plant in the United States, this paper analyzed a method of setting the DCGL that can be applied to Kori NPP Unit 1 which will be permanently disabled in 2017.

Triad case study: Rattlesnake Creek

AbstractThe U.S. Army Corps of Engineers (US ACE) is responsible for conducting the cleanup of radiological contaminated properties as part of the Formerly Utilized Sites Remedial Action Program. One property is the Rattlesnake Creek (RSC) portion of the Ashland sites. The RSC stream sediments are contaminated with thorium‐230, radium‐226, and uranium. The US ACE is closing RSC using protocols contained within the Multi‐Agency Radiation Survey and Site Investigation Manual (MARSSIM). At RSC, the US ACE developed site‐specific derived concentration guideline level (DCGL) cleanup requirements consistent with the MARSSIM guidance. Because of uncertainty about the distribution of contamination within the creek, the US ACE used the Triad approach to collect data and design remedial actions. Systematic planning helped target the areas of concern, develop a conceptual site model, and identify data gaps to be addressed before remediation plans were finalized. Preremediation sampling and analysis plans were designed to be explicitly consistent with final status survey requirements, allowing data sets to support both excavation planning needs and closure requirements in areas where contamination was not encountered above DCGL standards. Judicious use of real‐ time technologies such as X‐ray fluorescence and gamma walkover surveys minimized expensive off‐ site alpha spectrometry analyses, and at the same time provided the ability to respond to unexpected field conditions. © 2004 Wiley Periodicals, Inc.

Scan MDCs for multiple radionuclides in Class 1 areas.

The MultiAgency Radiation Survey and Site Investigation Manual (MARSSIM), published in December 1997, has been used for designing final status surveys at a number of D&D sites to date. One of the more challenging aspects of the MARSSIM survey design is the strategy employed when multiple radionuclides are present, particularly in Class 1 survey units. The MARSSIM recommends that potential hot spots in a Class 1 survey unit--that could exceed the derived concentration guideline levels for small elevated areas of radioactivity (i.e., DCGL(EMC))--have a reasonably good probability of being detected. Soil sampling on a specified grid size, in conjunction with gamma radiation surface scanning, is used to obtain an adequate assurance level that these hot spots are not missed during the final status survey. Therefore, it is sometimes necessary to supplement the sample size required by the statistical test to provide additional assurance that potential hot spots of concern are detected.