Final Disposal Of Radioactive Waste Research Articles

"Feels Like an Indie Game"-Evaluation of a Virtual Field Trip Prototype on Radioactive Waste Management Research for University Education.

This article describes the design and evaluation of a virtual field trip on the topic of radioactive waste management research for university education. We created an interactive virtual tour through the Mont Terri underground research laboratory by enhancing the virtual experiment information system, designed for domain experts, with background information, illustrations, tasks, tests, and an improved user interface. To put the tour's content into context, a conventional introductory presentation on the final disposal of radioactive waste was added. A user study with 22 participants proved a good perceived usability of the virtual tour and the virtual field trip's ability to transfer knowledge. These results suggest a benefit of employing virtual field trips in geoscientific university courses. In addition, it is conceivable to use the virtual field trip as a tool for science communication in the context of participatory processes during nuclear waste disposal site selection processes.

Requirements for inventory models of radioactive waste

Abstract. The aim of the final disposal of radioactive waste is to ensure the best possible safety for the long-term protection (i.e. 1 million years) of humans and the environment from ionizing radiation and other harmful effects of this waste. In Germany, the Site Selection Act (StandAG) regulates the procedure for determining the site with the best possible safety for a repository with high-level radioactive waste. One of the mandatory aspects of the StandAG is to perform various preliminary safety assessments in order to assess the expected extent of the safe confinement of the radioactive waste. For this purpose, all information (in particular on the quantity, type, composition and activity) of the waste (i.e. the inventory) is required to conduct the respective safety assessment. On behalf of the Federal Office for the Safety of Nuclear Waste Management (BASE) and within the framework of the research project “Significance of the radioactive waste inventory in the site selection for a repository (InvEnd)”, Brenk Systemplanung GmbH identified the characteristics of the inventory required for the preliminary safety assessments. To determine these inventory characteristics, two main data sources were analysed, namely the generic but host-rock-specific concepts for a German high-level radioactive waste repository and the inventory data and approaches of other countries with technologically developed repository programmes. The identified characteristics can be divided into different groups (e.g. geometric, chemical, radiological and physical) and categories such as primary inventory data that are essential for carrying out the preliminary safety assessments or secondary inventory data which validate the primary inventory data. The evaluation of the impact of the identified characteristics shows that especially data on component-specific chemical composition and component-specific radionuclide inventory have a high impact on the results of safety assessments. These characteristics are, in turn, dependent on other characteristics, such as the irradiation history and the original chemical composition of the fuel in addition to the structural parts of the fuel assemblies. Furthermore, the study assesses uncertainties of the inventory characteristics, such as data uncertainty and conceptual uncertainty. The extent of data uncertainty depends on the inventory characteristic. However, in most cases, only qualitative assessments are possible. Conceptual uncertainties depend on the degree of simplification used to define reference wastes but cannot be quantified yet, as no reference wastes have been defined. The results of this study enable authorities and waste management organizations to identify the key data of high-level radioactive waste inventories that are essential for conducting safety assessments of repositories, even during early stages of site selection and planning.

Influence of microbial uranium reduction processes on the final disposal of radioactive waste

Abstract. The safe disposal of high-level radioactive waste represents a major scientific and societal challenge. In addition to geological, geochemical and geophysical properties of such a repository, the influence of naturally occurring microorganisms from deep geological layers has to be taken into account for a comprehensive safeguard concept. Various sulfate- and iron-reducing bacteria are present in different clay formations, which can serve as a potential host rock for the long-term disposal of the waste, as well as in the backfill material bentonite. In the event of a worst-case scenario, if water enters the repository, those microorganisms can interact with the waste and change, for example, the oxidation state or the chemical speciation, which can influence the mobility of the radionuclides. In this study, the reduction of highly mobile, water-soluble U(VI) to less mobile U(IV) by the iron-reducing microorganism Desulfitobacterium sp. G1-2 and the sulfate reducer Desulfosporosinus hippei DSM 8344T were investigated. Desulfitobacterium sp. G1-2 has been isolated from a bentonite sample, and Desulfosporosinus hippei DSM 8344T represents a genus of sulfate-reducing bacteria present in clay rock and bentonite. During time-dependent experiments in a bicarbonate buffer (30 mM, 100/550 µM U(VI)), Desulfitobacterium sp. G1-2 showed a removal of up to 80 % of U within 5 d, whereas samples of Desulfosporosinus hippei DSM 8344T showed no decrease in U concentrations over time. Therefore, experiments were carried out in artificial Opalinus Clay pore water with this bacterium (100 µM U(VI), pH 5.5). In this case, the U concentration showed a decrease of up to 80 % of the radionuclide from the supernatants within 48 h. UV–Vis studies of dissolved cell pellets of both bacteria after U incubation showed an almost complete reduction to U(IV) for Desulfitobacterium sp. G1-2. On the other hand, samples of Desulfosporosinus hippei DSM 8344T exhibited only a partial reduction. Transmission electron microscopy (TEM) imaging combined with energy-dispersive X-ray spectroscopy analysis revealed the release of membrane vesicles from cells of Desulfosporosinus hippei DSM 8344T as a possible defense reaction against cell incrustation. Furthermore, TEM images of Desulfitobacterium sp. G1-2 cells showed the presence of two different U-containing aggregates inside the cells. These investigations showed clear differences in the reduction behavior of sulfate- and iron-reducing bacteria relevant to nuclear waste disposal. This highlights the importance of studies on the U(VI) interactions of different microorganisms present in deep geological layers. Moreover, new aspects for a safety concept for a nuclear repository in clay formations and for final disposal sites using bentonite as backfill material could be revealed.

Systematic laboratory investigations concerning the reduction of uncertainties related to the modeling of the material behavior of crushed salt

Abstract. The planned use of crushed salt as the backfill and sealing material in the context of the final disposal of radioactive waste in rock salt is justified and favored by the realization that the material properties of loosely deposited crushed salt will be transformed into the material properties of in situ rock salt in the long term by the convergence of the viscous rock salt mass. In other words, the convergence-induced compaction of crushed salt ultimately results in a nearly pore-free material, with a tightness and creep capacity corresponding to that of the ambient rock salt mass. The computational prognosis of the load-bearing and sealing behavior required for the long-term safety assessment of underground repositories requires a sufficiently proven physical modeling of the, in general, three-dimensional load bearing and permeability behavior of crushed salt. This requirement can only be met if the influencing variables on the load bearing and permeability behavior are sufficiently quantified and represented by mathematical–physical functions on the basis of suitable laboratory and field investigations. In order to eliminate or reduce the currently (still) existing scientific deficits and uncertainties concerning the general three-dimensional load bearing and permeability behavior of crushed salt, a systematized laboratory program was developed within the framework of the joint project KOMPASS (compaction of crushed salt for the safe containment), with the aim of ensuring an isolated recording of individual processes and influencing factors on the compaction behavior. Based on this, triaxial long-term compaction tests, TUC-V2 and TUC-V4, with a test duration of 750 and 180 d, respectively, were carried out within the framework of the joint project KOMPASS and of the follow-up project KOMPASS II. Both tests, but especially TUC-V2 with its 2-year duration, represent a sequence of several loading phases for separate analysis of the effects of individual influencing factors on the compaction behavior of crushed salt. By integrating several loading phases in one long-term test each, it was possible, on the one hand, to eliminate the scattering of the material behavior that would have been expected if several individual tests had been carried out as an alternative and, on the other hand, to provide metrological evidence of the range of validity of the compaction behavior as a function of individual influencing factors for a comparatively wide range of different porosities in each case. The computational reanalysis of the long-term tests opens up the possibility of validating and strengthening the currently available constitutive model approaches with regard to their ability to influence several parameters on the compaction behavior, while eliminating the scatter band of the material behavior. The improvement of the existing constitutive model approaches up to the sufficient reproduction of the measurement results equals an increase in the validation status or the credibility of the constitutive models or, vice versa, a reduction in the uncertainties remaining due to (not yet) validated model assumptions. Within the framework of the KOMPASS II joint project, a benchmark was carried out against the background outlined above for the comparative analysis of the constitutive models currently available to the project partners for depicting the compaction behavior of crushed salt with regard to their ability to computationally represent the measurement results generated within the TUC-V2 experiment. Specifically, the focus is on constitutive model functions that control the viscous compaction and shear deformation behavior and, within these functions, the influencing factors of mean stress, deviatoric stress, and temperature, which were recorded by laboratory tests for medium ranges of porosities from ≈ 16 % to very low porosities of ≈ 2.5 %. Within the framework of the continuation of the investigations planned under the acronym MEASURES (Multi-scale Experimental and numerical Analysis of crushed Salt material Used as engineered backfill for a nuclear waste REpository in rock Salt), it is intended to continue the benchmark study with a more detailed investigation of the influence of the deviatoric stress, which includes considering the data from TUC-V4. It is also intended to supplement the laboratory data base with further experiments by (a) investigations on the compaction behavior in the range of high-to-medium porosities and (b) investigations on the influence of the moisture content on the compaction capability (to be performed with KOMPASS reference material).

Stress state estimation – new data and variability assessment of model results

Abstract. The upper crust of the Earth is used more and more to transport and extract raw materials and energy. It is also used for the final disposal of radioactive waste in deep geological repositories with a particularly crucial requirement for long-term safety over 1×106 years. An accurate knowledge of the hosting rock and surrounding formation coupled with comprehensive modelling is fundamental to demonstrate the geological site has the required properties for safe and long-term underground storage. Besides other criteria, geomechanics play an important role. Even if the determination of rock properties is already a challenge, the estimation of the contemporary stress state in the upper crust is even more challenging. To predict the stress state 3-D geomechanical–numerical models are used, but high-quality underground structure, rock properties and in situ stress data for model calibration are needed. In particular, data of maximum and minimum horizontal stress magnitudes (SHmax and Shmin) are required to find suitable lateral displacement boundary conditions to produce a best-fit model. During the recent exploration phase for a deep geological repository for radioactive waste in Switzerland, a unique dataset of stress magnitude data has been acquired from eight cored boreholes. Rock mechanical properties were constrained from geophysical logging and laboratory testing. The empirically correlated rock properties were not simply averaged, but a probability distribution was provided. The stress field was explored by conducting more than 120 micro-hydraulic fracturing (MHF) and dry sleeve re-opening (SR) tests in different stratigraphic units, to estimate the magnitudes of Shmin and SHmax. We present the results of a 3-D geomechanical–numerical model that shows the best fit with respect to the measured stress magnitudes. Considering the uncertainties of the MHF/SR tests and the ones resulting from rock property variability, the model can reproduce most of the measurements. However, we do show not only the best-fit result but also a bandwidth of individual stress components within a P05–P95 probability range. We also show that there are still outliers for this bandwidth, but we will show that these can be explained by scaling effects.

Identification of areas of interest for radiological impact assessments in an evolving landscape context

Regulations concerning potential future health impacts of the final disposal of radioactive wastes in geological disposal facilities are written in terms of annual dose to individuals who are representative of small groups living in the landscape in the vicinity of the repository site. As disposal programmes in Sweden and Finland have progressed towards licensing and construction, so too has detail describing the state and evolution of surface biosphere and the landscape around the proposed disposal sites increased. Simple and generic biosphere dose assessment models in early iterations have grown in complexity with increasing site-specific detail that aims to capture the radiologically significant features of the landscape into which future releases of radionuclides might credibly occur. Current dose assessment models used in support of license applications for disposal programmes in Sweden and Finland are highly complex and their application consequently lacks transparency. An alternative simpler approach to characterising landscape objects for dose assessment models would be beneficial in that it would offer an additional line of reasoning and would add clarity, thereby supporting the decision-making process of the regulatory authorities.In the context of coastal Fennoscandia, landscape change is relatively rapid and dramatic with post-glacial landrise transforming areas of the coastal seabed into terrestrial ecosystems over a period of a few thousand years, global sea level rise notwithstanding. The locations of the geosphere-biosphere interfaces for deep geologic disposal can be estimated with some precision but the nature of the receiving ecosystems at the time of the release is less certain.The approach described here provides a statistical quantification of key morphological characteristics of areas in the landscape where doses could arise, so as to better express uncertainties in dose modelling. The proposed method assumes that the variation in the morphology of potential release locations can be described by the variation in landscape objects seen in the landscape on a wider scale, providing a statistical description of the possible landscape objects, so allowing a more comprehensive range of potential future evolutions to be addressed.Our understanding of the evolution of the landscape, based on the kinds of terrain and ecosystem development models used by POSIVA in Finland and SKB in Sweden, suggests that objects identified in present-day maps can be used as analogues for a statistical characterisation of objects in the future landscape; objects identified in the observed topography and bathymetry can therefore serve as the basis for the statistical description of landscape dose objects over the period during which doses are likely to arise. Using digital elevation models around a disposal site in Finland, we show that the statistical descriptions of landscape dose objects at three times over a period of 10 kyear of the evolved landscape are sufficiently similar to establish the suitability of the approach.The aim of this statistical analysis is to supplement current methods for defining radiological assessment models so as to provide additional numerical support to both the simpler and more complex methods employed by implementors and regulators. The method has been developed in the context of the Swedish and Finnish regulatory review process and is referred to in the IAEA's revised BIOMASS methodology. We briefly address how the method might be applied in other landscape contexts.

Radioactive waste management: Societal challenges in the era of green nuclear energy

Abstract Recently, there have been international debates over the idea of labeling nuclear power plants as a green investment. To be considered sustainable, to get a green investment label and to contribute to green growth, nuclear energy should be based on clear plans regarding safe disposal of radioactive waste. Final disposal of radioactive waste is witness to many disputes, one of the most important being that of public acceptance. Society represents a critical and a decisive stage in the process of radioactive waste management. Management and disposal process of radioactive waste requires community confidence and acceptance. The socio-political context must be addressed continuously through stakeholder commitment and public concerns. The final stage of radioactive waste management is especially characterized by population involvement, stakeholder availability, and mutual dialog. The objective of this study is based on finding the relationship between the location of a radioactive waste disposal facility (RWDF) and the perception of the population regarding various aspects related to nuclear energy and radioactive waste. A series of questions in the form of a questionnaire were used for data collection. A total of 200 valid questionnaires were used for this analysis. The study was focused on the Romanian population following a homogenous distribution throughout the country. Therefore, to better understand the issue, the paper investigates the correlation between the location of a radioactive waste disposal facility and a variety of factors, such as: attitude, perceived benefits, perceived risks, overall knowledge and perceived costs. Spearman rank correlation was used for data analysis. Bringing together theoretical information about nuclear energy and radioactive waste and the empirical data collected at the national level, this research study showed that the acceptance of a radioactive waste disposal facility is much stronger if the population is in favor of nuclear energy. As well, perceived risks create major concerns throughout the population and influence the location of a radioactive waste disposal facility. Cost aspects are controversial and create disagreements in relation with radioactive waste management options. That being said, this study brings value to the nuclear field by emphasizing the important societal factors which influence the location of a radioactive waste disposal facility in the green energy era.

Challenges and Opportunities in the Sustainable Radioactive Waste Management

Nuclear power plants are worldwide operated. While some countries consider phasing out nuclear power, others are considering building new power plants. Controversial debates have always arisen over the sustainability of nuclear energy, and radioactive waste has been the main topic. The possibility of long term management of radioactive waste considering the population behavioural and demographic characteristics is of particular importance in this study. The aim of the study was to examine if demographic factors and behavioural factors influence the sustainable management of radioactive waste, more precisely their final disposal. In particular, we analysed the relationship between each variable and the acceptance of the final disposal of radioactive waste and the location of the disposal facility. To find the answer of the present research study, we set the acceptance of the final disposal of radioactive waste and the location of the disposal facility as dependent variable and two demographic variables and two behavioural variables as independent variables. Statistical correlations were performed between the selected variables by using SPSS Statistics to establish the relationship between the factors at the individual level and the acceptance of the final disposal of radioactive waste in the context of sustainable development. Questionnaires were used as research tool. Based on this investigation, it was found that there is no correlation between perceived risks and the acceptance of the final disposal of radioactive waste and the location of the disposal facility. Likewise, this study shows that individuals’ acceptance of radioactive waste is based on individual behaviour, but it is also influenced by the demographic characteristics.

Программный модуль оценки достижения радиологической эквивалентности (РОЗА-РАО)

The main problem of the large-scale development of nuclear energy in the world is the safe dis-posal of accumulated radioactive waste (RW). The hazard of RW at the time of their formation to human health far exceeds the hazard of the source natural uranium ore, after the use of which these wastes were formed. Due to the natural processes of radioactive decay, the hazard of radi-oactive waste decreases over time. The hazard of waste can also be reduced by artificial trans-mutation or incineration of the radionuclides that make up RW composition. For uranium nuclear fuel cycles, the natural reference level of RW hazard is the hazard level of the uranium ore, upon reaching which decisions can be made on the final disposal of RW. The article presents a meth-od for estimating the time of achieving radiological equivalence of the mass of accumulated RW of nuclear energy and the source mass of natural uranium raw materials used in this case. A de-scription of the software module "Radiological support of protection – radioactive waste" (PM ROZA-RAO), which implements this method, is also presented. Radiological equivalence herein refers to the equality of the corresponding radiation risks for the population. The methodology used is determined by the current recommendations of the International Commission on Radio-logical Protection (ICRP) and the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) on the calculation of human radiation risks from internal exposure based on equivalent doses in organs and tissues. The determination of equivalent doses to organs and tissues was based on dose coefficients of radioactive substances provided for widespread use by the United States Environmental Protection Agency (EPA). Calculations by means of the PM ROZA-RAO showed that within the framework of the most likely scenario for the development of two-component nuclear energy system in Russia, for radioactive waste accumulated by 2100, the time of onset of radiological equivalence is less than 100 years (98.8 years). With the help of the developed software module, for the first time in the nuclear industry, the parameters of the processing of exposed nuclear fuel (the composition and activity of radionuclides for controlled storage, combustion and transmutation) and the management of RW can be determined, minimizing the potential radiation risks of the population, which corresponds to the best practice of developed countries in the field of radiation protection.

Predisposal conditioning, treatment, and performance assessment of radioactive waste streams

Before the final disposal of radioactive wastes, various processes can be implemented to optimise the waste form. This can include different chemical and physical treatments, such as thermal treatment for waste reduction, waste conditioning for homogenisation and waste immobilisation for stabilisation prior to packaging and interim storage. Ensuring the durability and safety of the waste matrices and packages through performance and condition assessment is important for waste owners, waste management organisations, regulators and wider stakeholder communities. Technical achievements and lessons learned from the THERAMIN and PREDIS projects focused on low- and intermediate-level waste handling is shared here. The recently completed project on Thermal Treatment for Radioactive Waste Minimization and Hazard Reduction (THERAMIN) made advances in demonstrating the feasibility of different thermal treatment techniques to reduce volume and immobilise different streams of radioactive waste (LILW) prior to disposal. The Pre-Disposal Management of Radioactive Waste (PREDIS) project addresses innovations in the treatment of metallic materials, liquid organic waste and solid organic waste, which can result from nuclear power plant operation, decommissioning and other industrial processes. The project also addresses digitalisation solutions for improved safety and efficiency in handling and assessing cemented-waste packages in extended interim surface storage.

Knowledge Management in the Federal Company for Radioactive Waste Disposal (BGE)

Abstract. With completion of the restructuring of the nuclear waste disposal landscape in Germany, the competences for construction and running of repositories for radioactive waste products are since 2017 with the Federal Company for Radioactive Waste Disposal (BGE). The BGE has over 50 years of experience in the disposal of nuclear waste in Germany. Over a period of several decades substantial effort was put into the research and developmental work on nuclear waste disposal in Germany, which led to a considerable state of knowledge, both thematically and quantitatively; however, so far no consistent overview exists on the various works and therefore on the total state of knowledge, as these works only partially exist or existed in the very different organizations (diffuse and local) within the nuclear waste disposal landscape. The emerging loss of specialist knowledge in Germany caused by the withdrawal from nuclear energy and the recession of national mining activities is doing the rest. This imminent loss of knowledge must be taken into consideration as a critical factor, especially considering the age structure within the BGE and the simultaneously occurring processes on selection of a site for disposal of highly radioactive waste products. For this reason, in the division of research and development/knowledge management (R&D/KM) an independent department for knowledge management was created in order to establish an infrastructure for knowledge management and a connection between knowledge management platforms and knowledge carriers in the company. In addition to the collection and distribution of the available knowledge, current results from research and development are also to be fed into the “learning organization”. For this purpose, knowledge management provides a digital information basis, into which current results from research and development are also entered as a knowledge store. This knowledge store consisting of over 14 000 documents, contains mostly research reports and scientific publications, which are concerned with diverse topics for the final disposal of radioactive waste products in geological formations. The total stock of internal company documents available can be accessed with a browser-based text analysis software. Intelligent search algorithms render the textual contents accessible, combine them with synonyms and dictionaries deposited in the system and make the resulting hits of the search queries available for the user in order of importance in summarized and full text versions. Using specific query terms, as known from search machines or library servers, this software analyzes the available documents of the digital information basis and in a brief summary of results, in addition to a brief description of the contents, the naming of relevant keywords, the identification of sources, compilers, institutions, knowledge carriers and an extended optimized information analysis of hits, also provides the possibility to retrieve the complete document and the download. The research options can, in addition to the queries via the established internet search engines also be combined with queries via the incorporated information or databank catalogue of national and international scientific institutions or libraries, which are concerned with research programs relevant for repositories. In order to further increase and optimize the information possibilities for employees of the BGE, a variety of knowledge landscapes has been implemented in the intranet, which enable specific queries on topics, expert information, institution, country and persons. Using these tools a first essential target of knowledge management, the collection, availability and distribution of written down knowledge in the company has been achieved. For the development of person-related and group-related knowledge the department of knowledge management is compiling concepts, which can only be implemented together and jointly with them, as they border on certain interfaces in the organization/company. This particularly concerns measures that fall into the areas of human resources, such as personnel management and personnel strategies or in the area of the particular project; keyword lessons learnt.

High acidity- and radiation-resistant triazine-based POPs for recovery of Pd(II) from nuclear fission products

The recovery of platinum group metals (PGMS) from high-level radioactive liquid waste (HLLW) not only has a complementary role in alleviating the shortage of rare precious metals but also is helpful for the final disposal of radioactive waste. In this work, a new cationic covalent triazine framework (CTF) named as cationic pyridyl triazine polymeric network (CPTPN) is designed and synthesized by ZnCl2-catalyzed ionothermal method for recovering of Pd(II) from HLLW. The excremental results shows that palladium can be efficiently adsorbed by CPTPN from 3 M HNO3 solution via anion exchange with fast adsorption kinetics with adsorption equilibrium reached in 10 min. The adsorption isotherm for Pd(II) follows the Langmuir model, which reveals the maximum adsorption capacity (qm) of 339.8 mg g−1 and 389.7 mg g−1 at 3 M HNO3 for CPTPN-Cl and CPTPN-NO3 , respectively. Further studies show that the qm of CPTPN-NO3 in 6 M HNO3 can reach a higher level at 428.6 mg g−1, which is the highest record for Pd(II) recovery in 6 M HNO3. In addition to high thermal stability, CPTPN exhibits a good irradiation stability with excellent adsorption performance maintained after 1000 kGy of γ-irradiation. Furthermore, CPTPN has an excellent adsorption selectivity for Pd(II) in simulated HLLW. Although the selectivity of CPTPN-Cl for Pd(II) is found to be unsatisfactory due to the significant co-adsorption of Ag(I), this problem can be well solved by changing Cl− ions in CPTPN-Cl to NO3−. Both CPTPN-Cl and CPTPN-NO3 show excellent reusability in 5 adsorption–desorption cycles. Thus, CPTPN can be considered as a candidate adsorbent that can tolerate high irradiation and high acidity for selective recovery of Pd(II) from HLLW.

QUANTOM® − Optimization of the online neutron flux measurement system

For the final disposal of radioactive waste, the waste packages have to meet the acceptance requirements defined by national licensing and supervisory authorities. Nondestructive methods are very much preferred over destructive methods for the qualification or re-qualification. Existing nondestructive methods as integral or segmented gamma scanning or neutron counting only determine the isotope specific activity but do not allow quantifying other non-radioactive hazardous substances. These should have been documented during creation, conditioning, and packaging of the waste. But especially for legacy waste, this documentation is often poor or even missing. This gap is to be filled by the QUANTOM® measurement device that will determine the mass fraction of elements within a 200-l-drum using the Prompt- and Delayed- Gamma-Neutron-Activation-Analysis. In order to obtain a spatially resolved characterization, it will employ a segmented scanning approach. For the determination of the absolute mass fractions, the neutron flux inside the drum has to be known accurately. As the waste itself will alter the neutron distribution and flux, it is not possible to calculate the latter a priori from the gamma measurement. Hence the neutron flux has to be measured simultaneously with the gamma radiation. In this presentation, we will introduce the system for measuring the thermal neutron flux surrounding the waste drum from which the flux within the waste package has to be reconstructed. We performed a simulation study to score several possible detector placements for an improved reconstruction performance. We will show the outcome of these calculations and present the final design of the detector arrangement.

Feasibility of Using Steel Drums for the Confinement of Burned Low-Level Radioactive Waste Ashes

The selection of suitable materials to construct a facility for the final disposal of radioactive wastes (repository) is a challenge in materials science, since these facilities must be designed to provide confinement to the radionuclides for long periods of time. In the case of the low-level radioactive waste (LLRW) repository, the confinement of the species must be guaranteed for more than 300 years. In this context, the evaluation of the proposed materials must be performed under accelerated laboratory tests. One type of LLRW consists of ashes that result from the incineration of solid combustible radioactive waste. In Argentina, one of the proposed projects is to place these ashes inside steel drums and immobilized them by a cementation process. The issue to be resolved is whether the steel drums will undergo internal corrosion depending on the presence of certain deleterious species. The aim of the present work is to evaluate the corrosion susceptibility of steel drums containing simulated cemented incineration ashes. To this purpose, a special type of specimen was designed and the corrosion process was followed up for almost 4 years. It was found that, although the deleterious effect of chloride is undeniable, corrosion rates are not as high as to compromise the integrity of the steel drums for a design period of 300 years.

Long-term safety of the extended SFR - Methodology and conclusions from the SR-PSU project

ABSTRACTSince 1988 the Swedish Nuclear Fuel and Waste Management Co. operates a repository for low- and intermediate-level short-lived radioactive waste, SFR, in Forsmark, Sweden. Due to decommissioning of the nuclear power plants additional storage capacity is needed. In December 2014, an application to extend the repository was therefore submitted. One key component of this application was an assessment of post-closure safety of the extended SFR. For this safety assessment, a methodology based on that developed by SKB for the spent nuclear fuel repository was used and the impact of the degradation of repository components, the evolution of the surface system and changes of future climate on the radiological safety of the repository was assessed over a period of 100,000 years. The central conclusion of the SR-PSU safety assessment is that the extended SFR repository meets requirements on protection of human health and of the environment that have been established by the Swedish radiation safety authority for the final disposal of radioactive waste. Furthermore, the design of the repository was shown suitable for the waste selected and the applied methodology suitable for the safety assessment.

In situ X-ray diffraction observation of smectite hydration under constant volume

Smectite (especially, montmorillonite) is the major clay mineral constituent of bentonite, which is designed to play a key role as a buffer material in geological repositories for the final disposal of radioactive waste in Japan. It is therefore crucial to understand the hydration behavior of smectite in terms of swelling during hydration and saturation processes.Against such a background, the authors simultaneously observed behaviors of smectite swelling at the micro-level (i.e., both the generation of swelling pressure and the change of hydration state). In the experiments, deionized water was allowed to permeate into a dried specimen of smectite (named Kunipia-F®) with different dry densities (ρd: 0.97, 1.23, 1.43, 1.64 and 1.88 Mg/m3) under conditions of constant temperature and volume. The swelling pressure was measured using an in situ uniaxial consolidation apparatus during the water feeding process. Changes in local hydration states (i.e., one-molecular-layer hydration states to three-molecular-layer hydration states) were also simultaneously observed. Hydration among these different states propagated from the inlet side to the outlet side of the specimen. The authors discussed the relationships governing the hydration state, swelling pressure, the number of hydration moles, dry density, equilibrium final pressure, and then the dynamic mechanism behind pressure propagation.

Determination of scaling factors to estimate the radionuclide inventory of wastes from the IEA-R1 research reactor

Regulations for transfer and final disposal of radioactive waste require that the inventory of radionuclides for each package enclosing such waste must be estimated and declared. In this work, the scaling factor methodology was applied successfully to estimate the inventory of radionuclides in spent ion-exchange resins and spent activated charcoal, both with low and intermediate-activity level, from the IEA-R1 nuclear research reactor. Scaling factors or correlation functions were obtained linking the activity concentrations of 15 difficult to measure nuclides with two gamma-rays emitting key nuclides, reducing exposure to ionizing radiation, contamination risks and operational costs.

Investigation of radioactivity concentration in spent technetium generators

This study was carried out to survey and measure radioactivity concentration and estimate radiation dose level at the surface of spent technetium generator columns for the safe final disposal of radioactive waste. High resolution γ-spectrometry with the aid of handheld radiation survey meters has been used. The radioactivity measurements has shown that 238U, 40K and 137Cs were only measurable in one sample whereas 125Sb was found in 14 samples out of total of 20 samples with an activity concentration which ranged from 21 to 7404 with an average value of 1095Bq/kg. The activity concentration of 125Sb is highly variable indicating that the spent 99mTc generator columns are of different origin. This investigation highlighted the importance of radiation monitoring of spent technetium generators in the country in order to protect workers, and the public from the dangers posed by radioactive waste.

Coupled multiphase flow and elasto-plastic modelling of in-situ gas injection experiments in saturated claystone (Mont Terri Rock Laboratory)

Understanding gas migration in host rock is one of the most important issues to consider in the final disposal of radioactive waste in deep geological formations. In order to investigate gas transport in water-saturated argillaceous rock, constant gas pressure injection tests on the samples from the Mont Terri Rock Laboratory (Switzerland), as well as pulse injection tests with a stepwise increasing of gas pressure, were carried out. During gas migration, desaturation and hydraulic mechanical interaction can be observed. To simulate gas migration in water-saturated porous media, a coupled multiphase flow and mechanical method within the FEM code family OpenGeoSys (OGS) was developed including the van Genuchten function for the relationship between capillary pressure and water saturation, the approach of Mualem for the relative permeability of both gas and water phases, and an elasto-plastic deformation model. Two approaches were applied, taking into consideration the permeability change by pore gas pressure and by deformation, respectively. Both permeability approaches are based on experimental observations and allow for the significant increase in permeability when gas pressure exceeds the minimal stress. The material parameters for the multiphase model are determined using the measured data from laboratory experiments under well-defined test conditions. With the calibrated model parameters, the in-situ experiment was simulated. Good agreement between the numerical results and the experimental data is achieved.

Final disposal of radioactive waste

Summary. — In this paper the origin and properties of radioactive waste as well as its classification scheme (low-level waste - LLW, intermediate-level waste - ILW, high-level waste - HLW) are presented. The various options for conditioning of waste of different levels of radioactivity are reviewed. The composition, radiotoxicity and reprocessing of spent fuel and their effect on storage and options for final disposal are discussed. The current situation of final waste disposal in a selected number of countries is mentioned. Also, the role of the International Atomic Energy Agency with regard to the development and monitoring of international safety standards for both spent nuclear fuel and radioactive waste management is described.