RESRAD Code Research Articles

Measurement of Natural Radioactivity in Fly Ash from a Lignite-Fired Thermal Power Plant and Radiological Risk Assessment Using RESRAD Code

In this study, the activity concentrations of natural primordial radionuclides in fly ash (FA) samples collected from the landfill area of Afşin-Elbistan A lignite-fired thermal power plant (AE-A LFTPP) located in Kahramanmaraş province of Türkiye were measured using a gamma-ray spectrometer. Also, the radiological risk was assessed using the Residual Radioactivity Onsite 7.2 code. The average activity concentrations of 226Ra, 232Th and 40K in FA samples varied from 291 ± 10 to 853 ± 25 Bq/kg, 23 ± 1 to 125 ± 4 Bq/kg and 112 ± 10 to 315 ± 18 Bq/kg, respectively. The predicted maximum total annual effective dose rate due to the natural radionuclides and their progenies is 4.7, 4.4, and 4.0 mSv/y at t = 1000 y for 0.5-m, 1.25-m, and 2.0-m cover scenarios, respectively. While the predicted maximum total cancer risk is 4.56 × 10−3 (3.95 × 10−3 for 226Ra, 5.18 × 10−4 for 232Th and 6.14 × 10−15 for 40K), 4.00 × 10−3(3.80 × 10−3 for 226Ra, 1.90 × 10−4 for 232Th and 7.69 × 10−15 for 40K) and 3.71 × 10−3 (3.71 × 10−3 for 226Ra, 1.820 × 10−8 for 232Th and 7.66 × 10−15 for 40K) at t = 1000 y for 0.5-m, 1.25-m and 2.0-m cover scenario, respectively. Results also revealed that the total doses and cancer risks are less in thicker cover scenarios throughout the 1000 y time span and that 226Ra is the major contributor to the risk in all the scenarios owing to its higher activity concentration than 232Th and 40K.

A study on the dose evaluation of critical groups for unrestricted reuse of the site after the decommissioning of Kori Unit 1

In the calculation of DCGLs using RESRAD code, the selection of input parameter values is paramount, which is basically possible when specific measures for selecting reuse scenarios and probabilistic evaluation methods are established. When referring to the impact of the age group considered in the environmental impact assessment among the existing operations of the plant, it may also need to be applied in the decommissioning project, and there will be considerations along with site reuse scenarios and probabilistic evaluations. Therefore, the main purpose of this study is to distinguish the types of exposure affected by residual radioactivity contamination during the process of deriving DCGL to be used in the final site release stage of NPPs and to evaluate them by reasonably assigning age group parameter values. In addition, the case of entering a single value for parameter values by age group and the case of entering probabilistic distributions were compared and analyzed.

Evaluation of Cs-137 and natural radionuclides on different marine biota (crustacean and fishes) along Beheira governorate Coast-Egypt: RESRAD biota

Purpose This work will focused on the environmental and radio-ecological impacts occurred on an Egyptian coastal region, based on the radiochemical measurement of 238U, 137Cs, 232Th, 40K. The novelty of the study was cleared by the using of new technique showing the integration of two biological RESRAD models, lead to a probabilistic estimation of the radionuclides bioaccumulation in different consumed marine organisms and determination the probability of human cancer risk at different ages. Material and methods The 20 water samples were collected and mounted into clean containers, and their decay products, were measured in Bq.L−1, using different quality assurance tools. The study will used different statistical analysis and different RESRAD modeling codes were used in the study to predict the degree of environmental and radio-ecological impacts at the studied area, this will be helpful in order to define the impacts resulted from the transferring and accumulation of different radionuclides. Results showed that the highest human dose conversion factor values of (5, 10, 15, with adult ages) were measured in Th-232 (mrem/pCi) while the lowest ingestion conversion factors values were observed in artificial 137Cs. On the other hand the highest values of external risk factor in case of (5, 10, 15 with adult age) were observed in artificial Cs-137(mrem/pCi), while the lowest value of external risk factor with the same ages were observed in U-238. This will lead to continual monitoring of artificial Cs-137 in different marine coastal regions. The arrangement of the bioaccumulation value (BIV) in Bq.kg−1 which being calculated by using RESRAD-Biota in case of crustacean animals will be as follow: U-238 > Th-232 > Cs-137, while the arrangement of BIV in case of fish animals will be: Cs-137 > U-232 > Th-232. On the other hand the arrangement of Internal Dose Conversion factors in case of crustacean animals will be: U-238 > Th-232 > Cs-137. While the arrangement of Internal Dose Conversion factors in case of fish animals will be: Cs-137> Th-232 > U-238. Conclusion RESRAD code’s results showed that the arrangement of the bioaccumulation and Dose Conversion factors were depend on the type of marine living organism. RESRAD code also showed that there are increments of the calculated external risk factor values which resulted from the adult than all the infants (5,10 and 15 ages) ages this may be related to the continuous replacement of new human’s body cells during the growth stages. The study results showed that, environmental bioaccumulation impacts of the artificial Cs-137 were very effective in both marine living organisms and human as this will support the relation between the ingestion Cs-137 in the body(inside the soft tissues), and the probability of the human cancer risk. On the other hand, study results showed the importance of using RESRAD BIOA code for the EIA's ecological and radiological studies which should be done for any future industrial coastal .projects.

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.

Studying the Impact of a Nuclear-Powered Naval Ship Severe Accident on Aquatic Biota using Resrad Code

Studying the Impact of a Nuclear-Powered Naval Ship Severe Accident on Aquatic Biota using Resrad Code

Radiological safety evaluation of dismantled radioactive concrete from Kori Unit 1 in the disposal and recycling process

For evaluating the radiological safety of dismantled concrete, the process of disposal and recycling of the radioactive concrete generated during the dismantling of Kori Unit 1 is analyzed. Four scenarios are derived based on the analysis of the concrete recycling and disposal process, and the potential exposure to the workers and public during this process are calculated. VISIPLAN and RESRAD code are used for evaluating the dosages received by the workers and public in the following four scenarios: concrete inspection, transport of concrete by the truck driver, driving on a recycled concrete road, and public living near the landfilled concrete waste. Two worker exposure scenarios in the processing of concrete and two public exposure scenarios in recycling and disposal are considered; in all the scenarios, the exposure dose does not exceed the annual dose limit for each representative.

Derivation of preliminary DCGL for Kori Unit 1 using RESRAD-OFFSITE and comparison with RESRAD-ONSITE

The successful decommissioning of nuclear facilities requires the derivation of DCGL in accordance with national site release criteria. MARSSIM, a widely used decommissioning guideline, recommends that a probabilistic analysis should be performed to derive the DCGL, which can be done by using the RESRAD code. Among the RESRAD codes, the RESRAD-OFFSITE code is the latest software edition and can derive exposure dose of receptor located both on and off-site. In this study, the preliminary exposure dose and DCGL were obtained by using the RESRAD-OFFSITE code and compared with the results obtained by the RESRAD-ONSITE code. As a result of the analysis, it was found that some radionuclides exhibited similar exposure doses and DCGL values, whereas a deviation is observed in a few cases. In this study, the applicability of the RESRAD-OFFSITE code to the decommissioning was investigated, and the similarities and differences with the RESRAD-ONSITE code were analyzed.

Proposal for the list of potential radionuclides of interest during NPP site characterization or final status surveys

In the research or project planning for the decommissioning of a nuclear power plant, one of several preparations will be the establishment of a list of potential radionuclides to be considered at the time of characterization or final status surveys. Reliable data for selection of potential radionuclides during the transition period to prepare for decommissioning will depend heavily on historical data at the site or, where possible, sampling analysis. However, during the transition period, direct sampling can be challenging, depending on the circumstances of the site or national regulation. A methodology of selecting potential radionuclides for nuclear facility sites which largely consists of three major processes: production of initial list of radionuclides, selection of the insignificant radionuclide that will be eliminated, and consideration of site characterization or sampling. For developing a preliminary list of potential radionuclides for Kori Unit 1 decommissioning, the list of initial radionuclides was made referring to the technical documents applied at decommissioned NPPs in the U.S and additional reference materials applied until the operation of NPPs in Korea. For the screening of insignificant radionuclides, we applied criterion of less than 0.1% of the amount of radioactivity inventory and confirmed the dose fraction using the RESRAD code. The final suit of radionuclides was established, which should be supplemented by reflecting site characterization and sampling process in the future. Thus, the methodology and results for the selection of potential radionuclides suggested in this paper can give an insight as a future reference to deriving DCGLs in relation to site remediation of decommissioning nuclear plants.

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.

Safety Assessment on the Incineration Disposal of Regulation Exempt Waste by RESRAD Code

Safety Assessment on the Incineration Disposal of Regulation Exempt Waste by RESRAD Code

Uncertainty Management on Human Intrusion Scenario Assessment of the Near Surface Disposal Facility for Low and Intermediate-Level Radioactive Waste: Comparative Analysis of RESRAD and GENII

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.

Evaluation of distribution coefficients and concentration ratios of 90Sr and 137Cs in the Techa River and the Miass River

Empirical data on the behavior of radionuclides in aquatic ecosystems are needed for radioecological modeling, which is commonly used for predicting transfer of radionuclides, estimating doses, and assessing possible adverse effects on species and communities. Preliminary studies of radioecological parameters including distribution coefficients and concentration ratios, for 90Sr and 137Cs were not in full agreement with the default values used in the ERICA Tool and the RESRAD BIOTA codes. The unique radiation situation in the Techa River, which was contaminated by long-lived radionuclides (90Sr and 137Cs) in the middle of the last century allows improved knowledge about these parameters for river systems. Therefore, the study was focused on the evaluation of radioecological parameters (distribution coefficients and concentration ratios for 90Sr and 137Cs) for the Techa River and the Miass River, which is assumed as a comparison waterbody. To achieve the aim the current contamination of biotic and abiotic components of the river ecosystems was studied; distribution coefficients for 90Sr and 137Cs were calculated; concentration ratios of 90Sr and 137Cs for three fish species (roach, perch and pike), gastropods and filamentous algae were evaluated. Study results were then compared with default values available for use in the well-known computer codes ERICA Tool and RESRAD BIOTA (when site-specific data are not available). We show that the concentration ratios of 137Cs in whole fish bodies depend on the predominant type of nutrition (carnivores and phytophagous). The results presented here are useful in the context of improving of tools for assessing concentrations of radionuclides in biota, which could rely on a wider range of ecosystem information compared with the process limited the current versions of ERICA and RESRAD codes. Further, the concentration ratios of 90Sr are species-specific and strongly dependent on Ca2+ concentration in water. The universal characteristic allows us to combine the data of fish caught in the water with different mineralization by multiplying the concentration of Ca2+. The concentration ratios for fishes were well-fitted by Generalized Logistic Distribution function (GLD). In conclusion, the GLD can be used for probabilistic modeling of the concentration ratios in freshwater fishes to improve the confidence in the modeling results. This is important in the context of risk assessment and regulatory.

An evaluation of health risk to the public as a consequence of in situ uranium mining in Wyoming, USA

In the United States there is considerable public concern regarding the health effects of in situ recovery uranium mining. These concerns focus principally on exposure to contaminants mobilized in groundwater by the mining process. However, the risk arising as a result of mining must be viewed in light of the presence of naturally occurring uranium ore and other constituents which comprise a latent hazard. The United States Environmental Protection Agency recently proposed new guidelines for successful restoration of an in situ uranium mine by limiting concentrations of thirteen groundwater constituents: arsenic, barium, cadmium, chromium, lead, mercury, selenium, silver, nitrate (as nitrogen), molybdenum, radium, total uranium, and gross α activity. We investigated the changes occurring to these constituents at an ISR uranium mine in Wyoming, USA by comparing groundwater quality at baseline measurement to that at stability (post-restoration) testing. Of the groundwater constituents considered, only uranium and radium-226 showed significant (p < 0.05) deviation from site-wide baseline conditions in matched-wells. Uranium concentrations increased by a factor of 5.6 (95% CI 3.6–8.9 times greater) while radium-226 decreased by a factor of about one half (95% CI 0.42–0.75 times less). Change in risk was calculated using the RESRAD (onsite) code for an individual exposed as a resident-farmer; total radiation dose to a resident farmer decreased from pre-to post-mining by about 5.2 mSv y−1. Higher concentrations of uranium correspond to increased biomarkers of nephrotoxicity, however the clinical significance of this increase is unclear.

Practical application of the MARSSIM process to the site release of a Uranium Conversion Plant following decommissioning

The final stage in the decommissioning process consists of releasing a site and a building from regulatory control. In this study, we describe the practical application of MARSSIM process in performing a site remediation and conducting a final status survey and discuss the lessons learned. The release criterion for the site and the building were set up using site-specific parameters that were calculated using RESRAD and RESRAD-Build codes. The planning stage of the final status survey consisted of classifying the site, identifying the survey units and selecting the background reference area using the DQO process. The planning, implementation and assessment of the final status survey for the site and the building of the Uranium Conversion Plant were carried out to demonstrate that residual radioactivity levels meet the release criterion. The null hypothesis must reject in order to demonstrate the objective, null hypothesis is that residual radioactivity in the survey unit exceed the release criterion. The survey results containing the hot spots in the building, a nonparametric statistical test (Wilcoxon Rank Sum) was chosen for assessment due to the presence of the uranium contamination in background soil. The final status survey results showed that the release criteria were satisfied.

Development of DCGLs By Using Both Probabilistic and Deterministic Analyses in RESRAD (Onsite) and RESRAD-OFFSITE Codes

The Derived Concentration Guideline Levels for two building areas previously used in waste processing and storage at Argonne National Laboratory were developed using both probabilistic and deterministic radiological environmental pathway analysis. Four scenarios were considered. The two current uses considered were on-site industrial use and off-site residential use with farming. The two future uses (i.e., after an institutional control period of 100 y) were on-site recreational use and on-site residential use with farming. The RESRAD-OFFSITE code was used for the current-use off-site residential/farming scenario and RESRAD (onsite) was used for the other three scenarios. Contaminants of concern were identified from the past operations conducted in the buildings and the actual characterization done at the site. Derived Concentration Guideline Levels were developed for all four scenarios using deterministic and probabilistic approaches, which include both "peak-of-the-means" and "mean-of-the-peaks" analyses. The future-use on-site residential/farming scenario resulted in the most restrictive Derived Concentration Guideline Levels for most radionuclides.

해체사업의 최종현황조사를 위한 MARSSIM 적용

해체의 최종단계는 대상 부지 및 건물을 규제로부터 제외하는 것이다. MARSSIM은 부지를 개방하기 위한 최종현황조사를 수행함에 있어 요구되는 자료의 수집 등의 다양한 요건을 만족키기 위한 총괄적인 지침을 제공한다. 연구로 해체 후 부지 및 건물의 최종현황조사를 위해 MARSSIM에서 제시하고 있는 방법을 적용하였다. 연구로 부지의 특성을 반영한 개방기준을 도출하기 위해 RESRAD 및 RESRAD-Build 전산코드를 이용하여 부지 및 건물에 대해서 계산하였다. 부지 및 건물의 조사설계(Survey Design)를 위해서 잠재적 오염도 및 측정 결과를 활용하여 조사구역을 구분하였고, 개략조사 및 특성조사를 통해 수집된 다양한 결과에 기초하여 통계학적 검사를 통해 조사구역 별로 요구되는 시료의 수를 산정하게 된다. 측정된 결과에 기초하여 연구로 최종 개방기준에 만족하는 것으로 평가되었다. 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.

RELATIVE RADIOLOGICAL IMPACT FROM A REACTOR ACCIDENT IN THE CASE OF EMERGING NUCLEAR FUELS

An assessment has been carried out on the radiological impact on an area contaminated from an accident of a nuclear reactor loaded with different actinide fuels considered in transmutation and recycling schemes. The impact of these schemes is compared to reference cases of commercial UO2 and MOX fuels. The effective dose equivalent delivered to permanent residents has been calculated using the RESRAD code and used as an index for the assessment purposes. The highest and lowest doses would be delivered from the self-generating recycling of actinides in fast and thermal reactors, respectively. External irradiation is the main contributor to the dose delivered to the target population in comparison to ingestion and inhalation. The external dose delivered would be attributed for the first few years to 134Cs and for the following several tens of years to 137Cs.

RESRAD-OFFSITE – A new member of the RESRAD family of codes

The RESRAD-OFFSITE code provides expanded state-of-the art, cost-effective, and user-friendly methods for evaluating the radiological consequences to a receptor located onsite or outside the area of primary contamination. It calculates radiological dose and excess lifetime cancer risk with the predicted radionuclide concentrations in the environment, and derives soil cleanup guidelines corresponding to a specified dose limit. The RESRAD-OFFSITE code is an extension of the RESRAD (onsite) code, which has been widely used for calculating doses and risks from exposure to radioactively contaminated soils. The development of RESRAD-OFFSITE started more than 10 years ago, but new models and methodologies have been developed, tested, and incorporated since then. Some of the new models have been benchmarked against other independently developed (international) models. The databases used have also expanded to include all the radionuclides (more than 830) contained in the International Commission on Radiological Protection (ICRP) 38 database. Detailed information on the design and application of the RESRAD-OFFSITE code is presented in its recently published User's Manual.

Radiological Dose Assessment for Clearance of Biological Shield Concrete from KRR-2

A large amount of radioactive concrete waste was generated during the decommissioning work of the Korean Research Reactor #2 (KRR-2). More than 85% of the dismantled concrete waste will be disposed of by a recycling or landfill by the application of a clearance principle. Currently, the Korean Atomic Energy Act and relevant regulation comply with the international recommendation of the sources/practices meeting an individual dose criterion 10μsv/y and a collective dose criterion of less than 1 man.Sv/y. A variety of radionuclides was generated in the biological shield concrete by a neutron reaction during the operation of a reactor. The specific activities of these radionuclides were measured by using the developed methods and procedures. For the evaluation of a radiological dose, seven scenarios were selected which encompassed a realistic situation of a recycling such as a road construction and landfill disposal route. The external and internal exposure doses of each scenario were calculated by using the MCNP4C code and a mathematical model respectively. The dose assessment regarding the residents and industrial workers was evaluated by using the RESRAD code for the landfill disposal scenario. The maximum individual dose rate was 7.21 μSv/y and the maximum collective dose was 0.05 man. Sv/yr for the landfill disposal scenario.

Radiological Dose Assessment for a Regulatory Clearance of the Soil and Concrete Wastes at KAERI

At the KAERI, about 4,800 drums of radioactive contaminated soils and concretes wastes have been stored since their generation in 1988. As these wastes have been stored for more than 18 years, some of them can be regulatory cleared. Before such a regulatory clearance, the radiological characteristics of these wastes should be analyzed first. And based on the analysis result, radiological dose should be estimated to confirm the safety of a disposal of these cleared wastes. In this study, a working procedure for a representative sampling and analysis of the wastes was developed. According to the analysis result, Co-60 and Cs-137 were major radionuclides while some wastes were contaminated with Mn-54, Fe-59, I-131, Cs-134 or Eu-152 of an extremely low concentration. The analysis results show that about 75% of wastes have a radioactivity concentration below 0.1Bq/g and were considered for a clearance. It was assumed that radiation workers are exposed by the inhalation dose and the external dose. The radiological dose was estimated using the RESRAD code for the disposal workers and the management workers as 5.32µSv/yr and 3.13µSv/yr for an individual dose and 1.60E-4 man · Sv/yr and 3.13E-5 man · Sv/yr for a public dose, respectively.