Near-surface Disposal Research Articles

Development of a reference framework to assess stylized human intrusion scenarios using GENII Version 2 considering design features of planned near-surface disposal facility in Korea

Abstract A standard form of six stylized human intrusion scenarios for a near-surface disposal facility (e.g. the planned Korean repository, Gyeongju Phase II) is proposed through re-categorization of totally thirty-one past cases reported in public literature. A reference assessment framework for inadvertent human intrusion is newly developed using GENII Version 2 conforming to the ICRP Publication 60 and thereafter. Calculated dose from the assessment framework is verified by comparing with hand calculation results for simplified model equations independently derived. Results from GENII Version 2 and 1.485 show inevitable differences, which is mainly attributed to the difference in the external dose assessment model. If intake dose coefficients in GENII Version 1.485 are modified, the difference can be reduced but still exist to an extent. Through deterministic and probabilistic sensitivity analysis, most affecting four parameters are derived and uncertainties of the parameters are quantified. It is expected that the reference assessment framework together with representative stylized scenarios can be used to do a human intrusion impact assessment for a specific repository using site-specific information. Especially, the past practice of human intrusion impact assessment using GENII Version 1.485 with or without modification may be replaced with the new assessment framework developed in this study.

Safety study on HIC containing waste resin with respect to hydrogen release

To explore the behavior of radiolytically produced hydrogen release from the waste resin stored in a high integrated container (HIC), and the mechanism of hydrogen diffusion in a near-surface disposal facility, both experimental studies and numerical simulations were performed through an accelerated irradiation test and simulated disposal, respectively. Results indicated that, 100 years after disposal, the highest hydrogen concentration appeared in the cell where the HICs were placed. The volume fraction for different scenarios postulated in the numerical simulation was 2.64% for Scenario 1, 2.28% for Scenario 2, and 3.965% for Scenario 3, all of which are lower than the hydrogen explosion limit of 4.1%. The results indicated that the simulated HIC disposal scheme is safe.

Safety assessment of near surface disposal for radioactive waste in Serpong nuclear area using PRESTO software

The planning of the construction and operation of Near Surface Disposal (NSD) facilities for Radioactive Waste at Serpong Nuclear Area needs a safety assessment in order to provide assurance that the radioactive waste disposal system functions properly. The purpose of the safety assessment is to provide a reasonable scientific assurance that the disposal system will provide an adequate level of safety and meet the requirements to protect human health and the environment. The NSD facility is vault type with a size of 20.2 × 20.2 m with a depth of 2.6 m from the ground surface (4.43 m above the highest ground water level). The facility was designed to accommodate waste packaged in 200 l drums and 950 l concrete shells. The total capacity of the disposal will be 1350 drums and 144 concrete shells. The waste that was considered in the study contain radionuclides which are dominantly contained in the low level packaged waste and, as reference for the safety assessment, with short half-lifes (less than or equal to about 30 years), namely Co-60 and Cs-137. This study used the PRESTO software version 4.2 to calculate the migration/exposure of radionuclides from the facility into the environment as well as individual doses. The results indicate that the predicted individual doses of the population around the facility at the Serpong site still fulfill the dose limit value of the public based on BAPETEN Chairman Regulation No. 4/2013 on the Protection and Radiation Safety in the Utilization of Nuclear Energy.

A Systems Approach to Near Surface Disposal Facility Design

A fundamental near-surface disposal challenge is the development and use of a risk-informed and performance-based decision tool that incorporates the impacts of waste, waste form and waste container-specific, facility design-specific and facility location-specific factors into the overall assessment of anticipated facility performance. This decision-making tool must have the ability to assess the influence of each disposal facility component on overall performance for specific wastes and disposal site environments.

DETERMINATION OF CONCRETE VAULT THICKNESS OF NEAR SURFACE DISPOSAL FOR RADIOACTIVE WASTE AT SERPONG NUCLEAR AREA

In order to support and complement the radioactive waste management facilities in Indonesia, BATAN will build a demonstration disposal facility in Serpong Nuclear Area (SNA). Demonstration disposal that will be built is Near Surface Disposal (NSD) type. Engineered vault for NSD is reinforced concrete. The calculations for determining the thickness of NSD concrete vault is based on the conceptual design as the result of the placement optimization of demonstration disposal that takes into account the inventory of radioactive waste and environmental geology conditions of the site at Serpong Nuclear Area. The thickness of the vault in this paper is focused on its ability to withstand radiation from stored waste so that workers or people who are around the disposal facility is safe with maximum radiation dose limit rate of 0.3 μSv / h. The calculation is performed with the aid of MicroShield 7:02 and Rad Pro Calculator Version 3:26 software. From the calculation so that the dose rate at the outer surface of the vault to be 0.3 μSv / h, required walls made of concrete with a density of 2:35 g / cm3 is 62.8 cm thickness.

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.

The Status and Experiences of LILW Disposal Facility Construction

This paper presents the status and experiences of Low and Intermediate-level radioactive Waste(LILW) disposal facility in Korea. The LILW disposal facility of 800,000 drums (200 L/drum) is being constructed in the site of KORAD's LILW Disposal Center(206 m2, Gyeongju-si). The first stage of the LILW facility construction was completed in June 2014, and approved for use by Nuclear Safety and Security Commission (NSSC) in December 2014. The first stage of the LILW facility is a type of underground silo with disposal capacity of 35,200 m3 (100,000 drums). The second stage of the LILW facility is under construction to be completed by December 2020. Site construction is currently underway. The second stage of the LILW facility is a type of near-surface disposal with disposal capacity of 32,875 m3 (125,000 drums)

Assessing Impact of Sorption in Geological Medium on Permissible Activity of Radioactive Waste in Near-Surface Disposal Facilities

The paper considers the conservative scenario of potential exposure that envisages simultaneous destruction of barriers with simultaneous release of radionuclides by the example of Lot 3 near-surface radioactive waste disposal facility at the Vektor Industrial Complex located in the Chornobyl Exclusion Zone. A conceptual model that considers migration of radionuclides through the aeration zone and aquifer to the potable water well, as well as mixing of infiltration water containing radionuclides with ground water in case of reaching the aquifer, was developed to analyze the mentioned scenario. Permissible specific activity of radioactive waste in the facility is calculated under the assumption that radioactive waste contains only 90Sr radionuclide. Normalysa software is used for calculations.

Improvement of Safety Approach for Accidents During Operation of LILW Disposal Facility : Application for Operational Safety Assessment of the Near-surface LILW Disposal Facility in Korea

Improvement of Safety Approach for Accidents During Operation of LILW Disposal Facility : Application for Operational Safety Assessment of the Near-surface LILW Disposal Facility in Korea

How Thick Should Cover Layer be for Radioactive Waste Disposal Facility? The Case of the Yamin Plain, Israel

Two main natural processes which control the migration of radioactive contaminants to the biosphere were studied in the Yamin Plain in order to evaluate the thickness of a cover layer needed for near-surface radioactive waste disposal facility. The first is the natural erosion of the cover layer, and the second is the infiltration during rain and runoff events. The erosion rate of the soil surface was studied by optical stimulation luminescence technique. It was found that during the last 14,000 years, the erosion rate was 0.3 mm/y which are 3 m for 10,000 years. The infiltration depth assessment was based on water content measurements and numerical modeling. It shows that under the most extreme rain event having an equivalent rain of 84 mm, infiltration depth was limited to 4.5 m. Taking into account the two processes, the effective cover layer thickness over 10,000 years should be at least 7.5 m thick.

System reliability analysis for near-surface radioactive waste disposal facilities

ABSTRACTThe low- and intermediate-level radioactive wastes are generally disposed in near-surface disposal facility (NSDF). The NSDF is composed of engineered barriers. The probabilistic safety assessment model has been developed to analyse the performance of NSDF. The endpoints of the assessment are the concentration of the radionuclide in the groundwater and the corresponding dose rate of the radionuclide. The barrier system can have multiple failure modes but practically the possible failure modes are failure of top cover, failure of waste container, degradation of waste form and failure of bottom cover, which are usually considered as independent failure events. Through a sensitivity analysis, the most critical parameters affecting the design reliability for failure criteria are identified as the groundwater velocity and distribution coefficient. The study shows that for the NSDF considered, there is a high degree of dependence between the failure modes, and demonstrates the probability of the simultaneous occurrence of failures. Thus, the need to consider the system reliability in the NSDF is highlighted. The study also advocates the use of optimisation techniques to evaluate the probability of failure, which provides a better estimate of the probability of failure, as validated from the results obtained from Monte Carlo simulations.

Occupational and Public Exposure During Normal Operation of Radioactive Waste Disposal Facilities

This paper focuses on occupational and public exposure during operation of disposal facilities receiving liquid and solid radioactive waste of various classes and provides a comparative analysis of the relevant doses: actual and calculated at the design stage. Occupational and public exposure study presented in this paper covers normal operations of a radioactive waste disposal facility receiving waste. Results: Analysis of individual and collective occupational doses was performed based on data collected during operation of near-surface disposal facilities for short-lived intermediate-, lowand very low-level waste in France, as well as nearsurface disposal facilities for long-lived waste in Russia. Further analysis of occupational and public doses calculated at the design stage was completed covering a near-surface disposal facility in Belgium and deep disposal facilities in the United Kingdom and the Nizhne-Kansk rock massive (Russia). The results show that engineering and technical solutions enable almost complete elimination of internal occupational exposure, whereas external exposure doses would fall within the range of values typical for a basic nuclear facility. Conclusion: radioactive waste disposal facilities being developed, constructed and operated meet the safety requirements effective in the Russian Federation and consistent with relevant international recommendations. It has been found that individual occupational exposure doses commensurate with those received by personnel of similar facilities abroad. Furthermore, according to the forecasts, mean individual doses for personnel during radioactive waste disposal would be an order of magnitude lower than the dose limit of 20 mSv/year. As for the public exposure, during normal operation, potential impact is virtually impossible by delaminating boundaries of a nuclear facility sanitary protection zone inside which the disposal facility is located and can be solely attributed to the use of public roads during radioactive waste transportation to the disposal facility site.

Conceptual Model to Determine Maximum Activity of Radioactive Waste in Near-Surface Disposal Facilities

For development of the management strategy for radioactive waste to be placed in near-surface disposal facilities (NSDF), it is necessary to justify long-term safety of such facilities. Use of mathematical modelling methods for long-term forecasts of radwaste radiation impacts and assessment of radiation risks from radionuclides migration can help to resolve this issue. The purpose of the research was to develop the conceptual model for determining the maximum activity of radwaste to be safely disposed in the NSDF and to test it in the case of Lot 3 Vector NSDF (Chornobyl exclusion zone). This paper describes an approach to the development of such a model. The conceptual model of 90Sr migration from Lot 3 through aeration zone and aquifer soils was developed. The results of modelling are shown. The proposals on further steps for the model improvement were developed.

Radionuclide Release Prediction in Water and Soil at Demonstration Plant of Near Surface Disposal for Radioactive Waste

Near Surface Disposal (NSD) for Radioactive Waste that should be developed due to increment of the low level radioactive waste, need to be analyzed and evaluated related to the radiological impact of the environment. A research method applied is done by modeling the distribution of radionuclide releases process. Analysis related with the releases of radionuclide in water and soil is using PRESTO (Prediction of Radiological Effects Due to Shallow Trench Operations). The application scenarios selected in this safety assessment is the migrations of Co-60 and Cs-137 scenario through the shallow groundwater flow pattern in the NSD site. The SigmaPlot software is also used to determine the concentration equation in well water and river water. The final results showed the concentration of radionuclide in wells and streams below the provision. Radionuclide activity concentrations in well ranged from 10<sup>-10</sup>Bq/m<sup>3</sup> to 10<sup>0</sup>Bq/m<sup>3</sup> and in the river ranged from 10<sup>-15</sup>Bq / m<sup>3</sup> to 10<sup>-1</sup>Bq / m3. The impact of radioactive waste of radionuclide Co-60 and Cs-137 will decrease to the background radiation level at a distance less than 10m and penetrate into the saturated layer up to 4m. In this study, an equation have been obtained that can predict radionuclide concentration patterns based on the distance and the depth of the ground surface against to the facility operation time.

Evaluation of distribution coefficient of vadose zone soil from proposed near-surface disposal facility at Kalpakkam, India

Near-surface disposal facility for storing low and intermediate levels of radioactive waste is proposed to be built in the premises of the Madras Atomic Power Station, Kalpakkam, India. The waste inventory of proposed near-surface disposal facility (NSDF) at Kalpakkam would predominantly contain radioactive strontium and cesium ions, and major portion of the NSDF would be located in the 3.2 m thick unsaturated soil zone. As distribution coefficient (K d) has an important bearing on contaminant transport, the paper examines how strontium retention varies with surface coverage in column experiments, and compares cumulative strontium retention using column and batch techniques. Results of column experiments show that strontium ion retention by the vadose zone soil for individual pore volumes ( $$K_{{{\text{d}}_{{{\text{pv}}_{i} }} }}$$ ) is maximum at immediate surface coverage (between 0.002 and 0.007 C/C 0). The maximum $$K_{{{\text{d}}_{{{\text{pv}}_{i} }} }}$$ values at immediate surface coverage are strongly influenced by variations in cation composition of the permeants. The maximum K d values steeply reduce (by 99 %) to near equilibrium values at 0.1–0.29 C/C 0 for all permeant compositions, suggesting that preferred adsorption site for strontium retention is exhausted at low surface coverage. The K d values from batch experiments are smaller than cumulative K d values of column experiments owing to much larger solution volume to soil solid ratio of batch experiments. Retardation factors based on cumulative K d values of column experiments and those based on batch K d values are comparable.

Health physics operating experience in transporting cleanup spent resin and verification of methodology for estimating the shielding thickness of cask

Cleanup (CU) demineralizer of TAPS 1 and 2, containing mixed bed resins, is used to maintain reactor water chemistry. Once the bed is saturated, spent resin needs to be transferred to resin fixation facility for fixation and subsequent disposal in near-surface disposal facility. This is done by transporting the spent resin in a shielded resin transportation cask (SRTC) from radwaste building. The SRTC was manufactured with shielding thickness, based on methodology proposed by Rautela et al., “Shielding adequacy of proposed cask for transporting TAPS 1 and 2 CU spent resins.” Extensive radiation monitoring was carried out during the entire transfer and transportation process. Observed radiation levels on cask surface were within the accepted and calculated radiation levels. This paper describes verification methodology proposed by Rautela et al. and health physics operating experience during first CU spent resin transfer process at TAPS 1 and 2.

Application of vadose zone approach for prediction of radionuclide transfer from near-surface disposal facility

About 100 000 m3 of solid conditioned Low and Intermediate Level Waste, generated during operation and decommissioning of the Ignalina nuclear power plant, are to be disposed of in a near-surface repository.In this study a computer code FEFLOW 5.0 was applied for simulating the transport of the most mobile radionuclides (3H, 14C, 59Ni and 94Nb) within moisture through an unsaturated vault of the near-surface repository in Stabatiškė Site. HYDRUS-1D analysis was used to assess the radionuclide transport in the repository and to estimate initial activity concentrations of radionuclides transported from the cemented waste. Radionuclide release from the vaults in the unsaturated conditions after closure of the repository and consequent contaminant plume transport has been assessed taking into account site-specific natural and engineering conditions and based on a normal evolution scenario.The highest value of 14C activity concentration (about 3.3·108 Bq/m3) in the moisture of soils below engineering barriers is predicted after about 7300 years after repository closure. The predicted maximum activity concentrations of the others modeled mobile radionuclide 3H (after 100 years after repository closure), 94Nb (after 300 000 years) and 59Ni (after 173 000 years) at the groundwater table are 1.1·105, 3.7·104 and 2.4·106 Bq/m3, respectively.

Overview of ICRP Committee 4: application of the Commission's recommendations.

Committee 4 develops principles and recommendations on radiological protection of people in all exposure situations. The committee meeting in 2014 was hosted by GE Healthcare in Arlington Heights, IL, USA on 27 July-1 August 2014. The programme of work of Committee 4 encompasses several broad areas, including a series of reports covering various aspects of existing exposure situations, leading the efforts of the International Commission on Radiological Protection (ICRP) to update and elaborate recommendations in light of the accident at Fukushima Daiichi nuclear power plant for emergencies and living in contaminated areas, elaborating the underpinnings of the system of radiological protection, and developing focussed reports on specific topic areas in consultation with ICRP's special liaison organisations. Committee 4 has six active Task Groups working on naturally occurring radioactive material; cosmic radiation in aviation; updates of ICRP Publications 109 and 111; ethics of radiological protection; surface and near-surface disposal of solid radioactive waste; and exposures resulting from contaminated sites from past industrial, military, and nuclear activities. In addition, there is a Working Party on tolerability of risk, and ongoing work with the various special liaison organisations of ICRP.

Role of the vadose zone in mitigating strontium transport at the near-surface disposal facility (NSDF) in Kalpakkam, India

A proposal has been advanced to construct a near-surface disposal facility for storing low- and intermediate-level radioactive waste on the premises of the Madras Atomic Power Station, Kalpakkam, located 80 km south of Chennai, India. The NSDF comprises a series of reinforced concrete trenches (RCTs), with major portions located in the unsaturated (vadose) soil zone. This zone is an important barrier against contaminant transport, as discontinuity in water-filled voids hinders solute transport to the underlying groundwater table. The hydraulic properties of the vadose zone, including soil-water characteristic curve (SWCC) and unsaturated permeability coefficient (k (unsat)), are important parameters in developing transport models for moisture and contaminants. As geotechnical characterization of the vadose zone soil at the proposed NSDF location has not been performed, the present study establishes hydraulic properties from field and laboratory measurements of soil samples. A one-dimensional model has been created for predicting strontium migration through the vadose zone for a range of volumetric content (theta; 0.16-0.44), based on the computed groundwater velocities and measured K (d) value. The results indicate that the presence of a 3-m vadose zone below the NSDF at Kalpakkam could result in a period of greater than 1000 years to attain strontium breakthrough concentration in groundwater, while the absence of the vadose zone would lead to strontium breakthrough concentration in 1-10 years.

Risk and Reliability Analysis of Multibarrier System for Near-Surface Disposal Facilities

AbstractThe performance of near-surface disposal facilities (NSDFs) for low-level radioactive waste depends on the type of materials used in the construction of barrier system, the properties of geological formations around the facility, and the type of waste containment system. By taking into account the effect of the factors mentioned here, a mathematical model can be developed to evaluate the migration of contaminants and radionuclides (via the contaminant transport model) from NSDFs to the nearest geosphere and biosphere, and assess the performance of the whole system. The model for migration of contaminants in a multibarrier system is considered in this paper for the safety assessment of NSDFs. The model considers scenarios during dumping period, after the dumping period, and after the closure of the NSDF. The annual radiation dose due to a radionuclide through drinking water pathway (for the single dump mode and multiple dump mode) is evaluated with the model. In the present work, a comprehensive an...