Solid Radioactive Waste Disposal Research Articles

Site-specific characterisation of hydrogeological parameters for low-level solid radioactive waste disposal facility at Tarapur, India

Site-specific characterisation of hydrogeological parameters for low-level solid radioactive waste disposal facility at Tarapur, India

Верифікація коефіцієнтів масштабування, призначених для радіологічної характеризації твердих радіоактивних відходів АЕС за методологією МАГАТЕ. Огляд сучасних підходів

To ensure the safety of repositories used for final disposal of solid radioactive waste (SRW) of nuclear power plants and other nuclear installations, the compliance of their radionuclide contamination levels with valid acceptance criteria has to be monitored. Scaling factors (SF) are used in many world countries to quickly solve the problems with radiological characterization of SRW subject to burial. To ensure the reliability of SF application in different countries, episodic or periodic verification and update of SF values is conducted. The purpose of work is to substantiate the global experience of SF verification, which, according to the IAEA methodology, is used for certification of specific and total activity of difficult-to-measure radionuclides (DTM) in SRW batches destined for disposal. The work addresses the current state of SF methodology application, and analyzes the main steps of its implementation. It was stated that current IAEA recommends declare the need for their verification and update of previously established SFs only, without offering any clear algorithms for implementation of this procedure and terms of its realization. However, in practice, such a procedure is used for radiological characterization of SRW in order to ensure quality guarantees in obtaining information regarding their radiation and other characteristics. This work also examines in detail the experience in applying the SF verification on the examples of several world countries. It was concluded that the verification should be a needed element in ensuring the reliability of SF use for radiological characterization of SRW. The experience of verification by means of SF validation using quantitative criteria is summarized. An algorithm is defined, according to which, and based on obtained results, it is possible to come to a statistically justified decision regarding the further use of existing SFs or recognition of a need in their updating for radiological characterization of new batch of waste

Repurposing a Spent Nuclear Fuel Cask for Disposal of Solid Intermediate Level Radioactive Waste From Decommissioning of a Nuclear Power Plant in Korea

Repurposing a Spent Nuclear Fuel Cask for Disposal of Solid Intermediate Level Radioactive Waste From Decommissioning of a Nuclear Power Plant in Korea

Approaches to the definition of potentially exposed groups and potentially exposed populations of biota in the context of solid radioactive waste

A methodology for addressing the biosphere in safety assessments for solid radioactive waste disposal was developed through theme 1 of the IAEA coordinated research project on BIOsphere Modelling and ASSessment (BIOMASS) that ran from 1996 to 2001. This methodology provided guidance on how the biosphere can be addressed in safety assessments for disposal of solid radioactive waste. Since the methodology was developed, it has proven useful and has been widely referenced in assessments in a diversity of contexts encompassing both near-surface and deep geological disposal of solid radioactive waste. The principles that could be adopted for defining potentially exposed groups (PEGs) were an important aspect in the original BIOMASS methodology as the endpoint of an assessment usually includes the evaluation of individual dose or risk to human health. Identification of PEGs and definition of their characteristics are usually made to be consistent with the biosphere system description being developed, acknowledging that due to inherent uncertainties in projecting future human behaviour, the biosphere models adopted for assessing safety of a disposal system can only be illustrative. Since the publication of the original BIOMASS methodology, consideration has been extended to include potentially exposed populations of biota (PEPs), in the context of dose assessment and protection of the environment. Considering the need for the development of transfer pathways from a source term to an end point (for either PEGs or PEPs), the exposure modes that may occur and those to be assessed quantitatively should be identified. Within an expert working group (WG6) of the second phase of the IAEA coordinated project Modelling and Data for Radiological Impact Assessments (MODARIA II), the experience of participating organisations has been collected on topics associated with the definition of PEGs and PEPs using a questionnaire. The objective of the questionnaire was to review the current status and on-going discussions on the handling of issues related to definitions of PEGs and PEPs as an input to the development of biosphere models for assessing radiological impacts on human health and the environment. The answers received to the questionnaire provided a clear overview of the progress that has been made since the original BIOMASS methodology was published, together with the lessons learned from the application of that methodology in the development of safety cases. This paper summarises the questionnaire responses in five subject areas: (1) environment of the PEGs and its evolution; (2) linking the choice of PEGs to these environments; (3) food habits and consumption rates; (4) populations of non-human biota (PEPs) and (5) national and international regulations and guidance. We illustrate how the results of the questionnaire have been used to enhance the original BIOMASS methodology (IAEA Enhanced BIOMASS Methodology Report in press).

A research and development roadmap to support applications of the enhanced BIOMASS methodology

The International Atomic Energy Agency has coordinated an international project addressing enhancements of methods for modelling the biosphere in post-closure safety assessments of solid radioactive waste disposal. This has resulted in the enhanced BIOMASS methodology that is described elsewhere in this special issue. To a large degree, the enhancements to the BIOMASS methodology arose from experience gained in applying the original methodology, both in the context of other international projects and in assessments of existing or proposed disposal facilities for solid radioactive wastes. Here, this experience is used, together with information on the status of solid radioactive waste disposal programmes worldwide, to identify opportunities for applying the enhanced methodology and for learning from those applications. This provides a basis for identifying research and development to support application of the enhanced methodology in a variety of environmental settings. These research and development requirements include aspects related to climate change under a variety of forcing scenarios, landform development in climatic regimes ranging from cold, polar to arid, tropical, modelling of groundwater flow and contaminant transport in surface-water catchments where both fractured rock and porous sediments are present, and studies of the transport of key radioisotopes of elements central to major biogeochemical cycles, such as those of carbon, chlorine, sulphur and iodine. In addition, some remarks are made on aspects of the application of the enhanced methodology that could imply review and updating of regulations and regulatory guidance, e.g. in relation to the definition of representative persons or groups to be considered in assessments and in respect of approaches to the assessment of radiological impacts on non-human biota. Furthermore, consideration is given as to how the scientific and technical experience that has been gained and methods that have been developed in the context of solid radioactive disposal facilities could support management of contaminated sites and legacy facilities that are likely to require long-term management and stewardship.

Evaluation of thermal plasma process for treatment disposal of solid radioactive waste

The management of radioactive waste is a worldwide activity based on the guidelines of the International Atomic Energy Agency (IAEA), and all stages of management require scientifically proven methods for possible deployment. The management of radioactive waste is a huge challenge due to the high risk in the collection, gathering, transport, handling, and storage. In this study, a thermal plasma treatment process was evaluated for its efficiency to process solid radioactive waste. Experiments were carried out with the application of stable isotopes of Lead, Iodine, Cobalt, and Cesium. After the thermal plasma treatments, the slag and the residual gas were analyzed to verify the influence of process time and discharge power on the efficiency of the process. The treatment for 25 min and 10 kW was sufficient to reduce the mass by 50% of the slag. When the applied power was increased to 15 kW, an expressive reduction in the treatment time (10 min) was able to promote the same mass reduction. The results indicated that the treatment of radioactive waste by thermal plasma is a promising method to manage and reduce the mass and volume for the final disposal.

Safety assessments undertaken using the BIOMASS methodology: lessons learnt and methodological enhancements

The International Atomic Energy Agency has coordinated an international project addressing enhancements of methods for modelling in post-closure safety assessments of solid radioactive waste disposal. The project used earlier published work from the IAEA biosphere modelling and assessment (BIOMASS) project to further develop methods and techniques. The task was supported by a parallel on-going project within the BIOPROTA forum. The output from the project is described in detail in a forthcoming IAEA report. Here an overview of the work is given to provide researchers in the broader fields of radioecology and radioactive waste disposal with a summarised review of the enhanced BIOMASS methodology and the work that has been undertaken during the project. It is hoped that such dissemination will support and promote integrated understanding and coherent treatment of the biosphere component within the overall assessment process. The key activities undertaken in the project were: review and identification of those parts of the original BIOMASS methodology that needed enhancement, discussions on lessons learned from applying the BIOMASS method, using real examples to assess the methodology and its usefulness, and writing of those parts of the methodology that were considered could benefit from refinement or for which new guidance was required to take account of scientific developments. The work has shown that the overall approach in the original BIOMASS methodology has proven sound. However, the enhanced version clarifies the need for an iterative and holistic approach with system understanding central to the approach. Specifically, experience, especially in site-specific contexts, has emphasised that adequate system understanding is essential in underpinning safety assessments for radioactive waste disposal. The integral role of the biosphere within safety assessment is also emphasised in the enhanced methodology.

Research and Development of Processing and Disposal of Solid Radioactive Waste

Research and Development of Processing and Disposal of Solid Radioactive Waste

Reprocessing of Irradiated [18O]H2O under the Conditions of a PET Center

The possibility of reprocessing of irradiated [18O]H2O after isolation of [18F]fluoride ion from it (regenerate) under the conditions of the cyclotron-radiochemical laboratory of an operating PET center was examined. The radionuclide and chemical composition of [18O]H2O and the degree of its enrichment in 18O before and after distillation at atmospheric pressure were studied comprehensively. The suggested method allows efficient removal of long-lived γ-emitting radionuclides (56Co, 57Co, 58Co, 65Zn, 54Mn, 51Cr) from the [18O]H2O regenerate. The distillation reduces the content of residual solvents (acetaldehyde, acetone, isopropanol, ethanol, acetonitrile), whereas the content of 3H and 18O in [18O]H2O changes within the measurement uncertainty. The distillation residues and decontamination solutions after neutralization can be jointly cemented to incorporate the γ-emitters into a compact matrix with the aim of its subsequent delivery to specialized organizations performing long-term storage and disposal of solid radioactive waste.

Climate change and landscape development in post-closure safety assessment of solid radioactive waste disposal: Results of an initiative of the IAEA

The International Atomic Energy Agency has coordinated an international project addressing climate change and landscape development in post-closure safety assessments of solid radioactive waste disposal. The work has been supported by results of parallel on-going research that has been published in a variety of reports and peer reviewed journal articles. The project is due to be described in detail in a forthcoming IAEA report. Noting the multi-disciplinary nature of post-closure safety assessments, here, an overview of the work is given to provide researchers in the broader fields of radioecology and radiological safety assessment with a review of the work that has been undertaken. It is hoped that such dissemination will support and promote integrated understanding and coherent treatment of climate change and landscape development within an overall assessment process.The key activities undertaken in the project were: identification of the key processes that drive environmental change (mainly those associated with climate and climate change), and description of how a relevant future may develop on a global scale; development of a methodology for characterising environmental change that is valid on a global scale, showing how modelled global changes in climate can be downscaled to provide information that may be needed for characterising environmental change in site-specific assessments, and illustrating different aspects of the methodology in a number of case studies that show the evolution of site characteristics and the implications for the dose assessment models.Overall, the study has shown that quantitative climate and landscape modelling has now developed to the stage that it can be used to define an envelope of climate and landscape change scenarios at specific sites and under specific greenhouse-gas emissions assumptions that is suitable for use in quantitative post-closure performance assessments. These scenarios are not predictions of the future, but are projections based on a well-established understanding of the important processes involved and their impacts on different types of landscape. Such projections support the understanding of, and selection of, plausible ranges of scenarios for use in post-closure safety assessments.

Особенности сорбции цезия в бентонитовых барьерных системах при захоронении твердых радиоактивных отходов

Особенности сорбции цезия в бентонитовых барьерных системах при захоронении твердых радиоактивных отходов

Stochastic Seepage Model in Repository of Low and Intermediate Level Solid Radioactive Wastes

Repository of Low and Intermediate Level Solid Radioactive Wastes (ILRW) is accepted to deal with the ILRW produced from fast development of nuclear power industry and other fields. In performance assessment of the potential geologic repository, modeling plays a very important role in addition to experimental inverstigations. In addition to experimental inverstigations modeling has also to be used for these purposes. In the paper, stochastic mathematical theories were used for calculating the radionuclide transport in fractured rock masses at Repository of Low and Intermediate Level Solid Radioactive Wastes. To better understand about the dangers of public for radiation and the nuclides transport in fractured rock masses at low and intermediate level solid radioactive waste disposal, the paper builds three dimensional stochastic seepage grid model to describe radionuclide transport in fractured rock masses based on three-dimensional directional seepage theory that is built into percolation mechanics, sets up a rigorous computer simulating system to calculate the radionuclide transport and safety performance of the Repository of Low and Intermediate Level Solid Radioactive Wastes based on computing technology. Finally, To illustrate the precision and validity of this model that is built into the stochastic mathematical theories or computing technology, the simulation experiments are carried out to understand law of radionuclide transport.

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.

Environmental Risk of Nuclear Power and Policy Proposal on Disposal of Solid Radioactive Waste

The difficulty occurred in nuclear power plants that the accumulated radioactive solid waste is beyond the design capacity and unable to be sent to disposal is focused on in this paper. The deep reasons for the difficulty occurred are concluded to be the unclear responsibility for disposal of radioactive waste and the divided national function of nuclear power development and radioactive waste management, by analyzing the disposal demand of radioactive solid waste caused by continuous development of nuclear power and the current situation and existing problems for the disposal of low-intermediate level radioactive solid waste in China. The policy suggestions of issuing the disposal siting plan of radioactive solid waste, forming independent firms of radioactive waste storage and disposal and improving radioactive waste management fund system are proposed based on above analysis and investigation.

ICRP PUBLICATION 122: Radiological Protection in Geological Disposal of Long-lived Solid Radioactive Waste

This report updates and consolidates previous recommendations of the International Commission on Radiological Protection (ICRP) related to solid waste disposal (ICRP, 1985, 1997b, 1998). The recommendations given apply specifically to geological disposal of long-lived solid radioactive waste. The report explains how the ICRP system of radiological protection described in Publication 103 (ICRP, 2007) can be applied in the context of the geological disposal of long-lived solid radioactive waste. Although the report is written as a standalone document, previous ICRP recommendations not dealt with in depth in the report are still valid.The 2007 ICRP system of radiological protection evolves from the previous process-based protection approach relying on the distinction between practices and interventions by moving to an approach based on the distinction between three types of exposure situation: planned, emergency and existing. The Recommendations maintains the Commission’s three fundamental principles of radiological protection namely: justification, optimisation of protection and the application of dose limits. They also maintain the current individual dose limits for effective dose and equivalent dose from all regulated sources in planned exposure situations. They re-enforce the principle of optimisation of radiological protection, which applies in a similar way to all exposure situations, subject to restrictions on individual doses: constraints for planned exposure situations, and reference levels for emergency and existing exposure situations. The Recommendations also include an approach for developing a framework to demonstrate radiological protection of the environment.This report describes the different stages in the life time of a geological disposal facility, and addresses the application of relevant radiological protection principles for each stage depending on the various exposure situations that can be encountered. In particular, the crucial factor that influences the application of the protection system over the different phases in the life time of a disposal facility is the level of oversight or ‘watchful care’ that is present. The level of oversight affects the capability to control the source, i.e. the waste and the repository, and to avoid or reduce potential exposures. Three main time frames are considered: time of direct oversight, when the disposal facility is being implemented and is under active supervision; time of indirect oversight, when the disposal facility is sealed and oversight is being exercised by regulators or special administrative bodies or society at large to provide additional assurance on behalf of society; and time of no oversight, when oversight is no longer exercised in case memory of the disposal facility is lost.

Intercomparison of Models of 14C in the Biosphere for Solid Radioactive Waste Disposal

Radiocarbon is present in solid radioactive wastes arising from the nuclear power industry, in reactor operating wastes, and in graphite and activated metals that will arise from reactor decommissioning. Its half-life of 5730 yr, among other factors, means that 14C may be released to the biosphere from radioactive waste repositories. These releases may occur as 14C-bearing gases, especially methane, or as aqueous species, and enter the biosphere from below via natural processes or via groundwater pumped from wells. Assessment of radiation doses to humans due to such releases must take account of the major role of carbon in biological processes, requiring specific 14C assessment models to be developed. Therefore, an intercomparison of 5 14C assessment models was organized by the international collaborative forum, BIOPROTA. The intercomparison identified significantly different results for the activity concentrations in the soil, atmosphere, and plant compartments, based upon the different modeling approaches. The major source of uncertainty was related to the identification of conditions under which mixing occurs and isotopic equilibrium is established. Furthermore, while the assumed release area plays a role in determining the calculated atmospheric 14C concentrations, the openness of the plant canopy and the wind profile in and above the canopy are the key drivers. The intercomparison has aided understanding of the processes involved and helped to identify areas where further research is required to address some of the uncertainties.

Report of ICRP Task Group 80: ‘Radiological protection in geological disposal of long-lived solid radioactive waste’

The report of International Commission on Radiological Protection (ICRP) Task Group 80 entitled ‘Radiological protection in geological disposal of long-lived solid radioactive waste’ updates and consolidates previous ICRP recommendations related to solid waste disposal (ICRP Publications 46, 77, and 81). The recommendations given in this report apply specifically to geological disposal of long-lived solid radioactive waste. The report explains how the 2007 system of radiological protection, described in ICRP Publication 103, can be applied in the context of the geological disposal of long-lived solid radioactive waste. The report is written as a self-standing document. It describes the different stages in the lifetime of a geological disposal facility, and addresses the application of relevant radiological protection principles for each stage depending on the various exposure situations that can be encountered. In particular, the crucial factor that influences application of the protection system over the different phases in the lifetime of a disposal facility is the level of oversight that is present. The level of oversight affects the capability to reduce or avoid exposures. Three main time frames have to be considered for the purpose of radiological protection: time of direct oversight when the disposal facility is being implemented and active oversight is taking place; time of indirect oversight when the disposal facility is sealed and indirect oversight is being exercised to provide additional assurance on behalf of the population; and time of no oversight when oversight is no longer exercised because memory is lost.

A mathematical model for the behaviour of Se-79 in soils and plants that takes account ofseasonal variations in soil hydrology

Se-79 is a long-lived radionuclide of potential radiological significance in relation to thedeep geological disposal of solid radioactive wastes. In the context of release to theterrestrial environment, its main radiological impact is delivered through food chainpathways. Therefore, its accumulation in soils and uptake by plants is an importantconsideration in post-closure safety assessment studies. However, representation of itsbehaviour in the soil–plant system requires consideration of the multiple valencestates that it can exhibit under different redox conditions and its susceptibility tovolatilisation. A simple model is described that includes seasonal variations in soilhydrology and their effects on the mobility and root uptake of Se-79. Illustrativecalculations are undertaken with the model, to demonstrate its capabilities forinterpreting experimental data on the behaviour of Se-79 in soils and plants, and formaking projections on the long-term behaviour of Se-79 transported to soils.

Discussions about safety criteria and guidelines for radioactive waste management

In Japan, the clearance levels for uranium-bearing waste have been established by the Nuclear Safety Commission (NSC). The criteria for uranium-bearing waste disposal are also necessary; however, the NSC has not concluded the discussion on this subject. Meanwhile, the General Administrative Group of the Radiation Council has concluded the revision of its former recommendation 'Regulatory exemption dose for radioactive solid waste disposal', the dose criteria after the institutional control period for a repository. The Standardization Committee on Radiation Protection in the Japan Health Physics Society (The Committee) also has developed the relevant safety criteria and guidelines for existing exposure situations, which are potentially applicable to uranium-bearing waste disposal. A new working group established by The Committee was initially aimed at developing criteria and guidelines specifically for uranium-bearing waste disposal; however, the aim has been shifted to broader criteria applicable to any radioactive wastes.

Modeling the inhalation enhancement factor in prospective radiological risk assessment

Inhalation of suspended particulates originating from contaminated soil is an important exposure pathway that may result from disposal of long-lived solid radioactive waste. A hypothetical scenario was evaluated involving long-term irrigation of agricultural land with contaminated water and ensuing accumulation of radionuclides in surface soil. One of the factors affecting the dose from inhalation of suspended soil particulates is the radionuclide enhancement (enrichment) in airborne soil particles relative to that of the underlying soil. The enhancement factor (enrichment factor) is defined as a ratio of airborne particle activity concentration (Bq kg-1 ) to surface soil activity concentration (Bq kg-1 ). The values of this parameter and its effect on inhalation dose were evaluated for different soil types and inhalation exposure conditions.