Intermediate-level Long-lived Waste Research Articles

Interaction between radio-oxidized polypropylene and gaseous HCl. Part 2. Kinetics and reaction mechanisms

In Part 1 of this article, evidence of cross-infection effects of polyolefins and halogenated polymers present in the long-lived intermediate-level waste (LL-ILW) packages was shown. We choose to study the interaction between polypropylene and one of the main degradation products of PVC, i.e. HCl. Polyolefins and chlorinated polymers were chosen because they are the two most abundant classes of polymer types found in the LL-ILW packages. We showed that the interaction between HCl and pristine or radio-oxidized PP results in HCl consumption and in CO2, H2O and CO formation. The object of the present article is to further study the reaction mechanisms. Using derivative methods at room temperature, the reaction of HCl with hydroperoxide was documented while alcohols seem stable. A model taking into account HCl diffusion, sorption in PP and reaction with hydroperoxide is insufficient to explain the observed HCl consumption kinetics. We concluded that HCl consumption is not only due to physical sorption and to chemical reaction with a unique oxidized compound, but multiple reactions with different oxidation products should be considered as well.

Analytical and Numerical Estimation of Fracture Initiation and Propagation Regions around Large-Diameter, Deep Boreholes for Disposal of Long-Lived Intermediate-Level Waste

The safety of high-level radioactive waste disposal has been studied across the world considering mined geologic repositories. Here, we introduce large-diameter, deep borehole disposal as one of the potential solutions for small volumes of long-lived intermediate-level waste (ILW). The short- and long-term stability of deep disposal boreholes is critical for environmental safety and public health. In this paper, we first use a recently revisited extensional strain criterion for fracture initiation and apply analytical solutions of a two-dimensional stress model to predict the fracturing region around a 2 km deep and 0.7 m diameter disposal borehole. Analytical solutions of fracture initiation are compared with results from the numerical simulator FRACOD, while the latter model also predicts dynamic effects such as fracture propagation. Both analytical and numerical methods predicted similar fracture initiation characteristics around the minor horizontal compressive stress springline, consistent with literature data. Numerical results showed deeper fracturing zones than those predicted by analytical solutions, mainly because the analytical predictions provide static snapshots under specific given conditions, while the numerical model calculates additional dynamic effects of fracture propagation. Including stress dynamics is shown to further weaken the rock around the borehole. At the bottom plane of the borehole, three-dimensional numerical simulations showed the development of fracturing zones around the major horizontal compressive stress springline. Borehole stability analyses are essential to plan the safe operation of drilling operations while also giving insights as to what borehole depths are more prone to fracturing and hence potentially less suitable as a waste disposal zone.

In-situ monitoring of micro-tunnel geometry change based on distributed optical fiber strain sensing

Underground radioactive waste disposal facilities must be monitored to guarantee their correct and safe exploitation during the early stage of their service life. This will be the case in Cigéo, the french industrial geological disposal facility for high-level and intermediate-level long-lived waste, which must be monitored for the first a hundred years at least. For this purpose optical fiber strain sensing cables can be chosen for their low intrusivity and resistance to harsh radiation environment. In this paper, a monitoring method able to follow shape evolution of the structure’s section, based on distributed strain sensing is tested on a high-level waste disposal cell mock-up, a steel lined micro-tunnel of 762 mm of outer diameter. The cell demonstrator is built in Meuse/Haute-Marne Underground Research Laboratory, in the same geological environment as the one envisaged for Cigéo but it is empty of waste. After 7 months of loading of the rock around the mock-up, a maximum of 2 mm of diameter reduction is observed with Rayleigh-scattering based strain sensing technique and 10 mm of spatial resolution around the circumference. This validation under realistic test conditions opens up to the use of the method in-situ for tunnel monitoring in harsh environment.

Modeling and simulation of ductile damage in forged steel used to encapsulate high-level radioactive waste

Andra, the French national radioactive waste management agency, is in charge of studying the disposal of high-level and long-lived intermediate-level waste (HLW and ILW-LL) in a deep geological repository. According to the reference concept, it is planned to encapsulate high-level waste in non-alloy P285NH steel overpacks before inserting them into horizontal steel cased micro-tunnels. This work is a part of the study about the long-term behavior of a welded steel overpack subjected to external hydrostatic pressure and several localized loading paths. Indeed, the main objective of this work is to develop the most suitable model for non-alloy steel P285NH to be used in the prediction of the long-term overpack behavior. Dealing with a ductile steel, elastoplastic constitutive equations accounting for mixed nonlinear isotropic and kinematic hardening strongly coupled with ductile isotropic damage are adopted. They are formulated based on the classical thermodynamics of irreversible processes framework with state variables at the macroscopic scale, (Germain, 1973) (Lemaitre 1985, Saanouni 2012). In this paper, a new coupling formulation between the scalar isotropic ductile damage and the deviatoric and spherical part of the Cauchy stress and elastic strain tensors is proposed. In order to calibrate the developed model on P285NH steel, multiple tensile tests are performed using classical cylindrical specimens, notched specimens and double notched specimens. In the last part, some experimental fields are measured using digital image correlation. Application is made to a simplified overpack represented by thick walled cylinder subject to compressive loading path. A FEM (Finite Element method) crushing operation of an overpack’s cylindrical part has simulated and analysed.

Deep borehole disposal of intermediate-level waste

Abstract. Around the world, deep borehole disposal is being evaluated for intermediate-level waste (ILW), high-level waste (HLW), spent nuclear fuel (SNF), separated plutonium waste and some very high specific activity fission product waste. In Australia, long-lived ILW from research reactors and radiopharmaceutical production represents the principal waste stream that requires deep geologic disposal. Whilst the Australian government has not yet made a decision on its preferred strategy for ILW disposal, deep borehole disposal of small volumes of ILW would be a more cost-effective and modular solution compared to a conventional geologic disposal facility (GDF). CSIRO, ANSTO and SANDIA have created an international partnership to execute a full-scale borehole research, development and demonstration (RD&D) project in Australia. The project will demonstrate the technical feasibility of the long-term safety of borehole disposal in deep geological formations. The execution of this project could also demonstrate options for nuclear waste disposal that would reduce proliferation risks, potentially up to the termination of compliance with international safeguards requirements. The RD&D includes demonstration of surface handling and waste/seal emplacement capabilities, basic research on foundational science areas, and full-scale field testing in both a deep characterization borehole and a larger-diameter (0.7 m or 27.5 inch) 2000 m deep demonstration borehole. The multi-barrier system designed for such a deep disposal borehole concept places much less reliance on engineered barriers at the disposal zone to achieve safety as compared to a conventional GDF. It rather relies on geological features for waste containment. The concept being explored uses disposal containers with primary waste packages, such as vitrified waste canisters, inside; to be both cost effective and fit for purpose, such a container could have a mild steel-based structural component with copper coating. A critical review of six coating technologies showed that cold spray has the greatest advantages, such as minimal porosity and compressive residual stress. The RD&D has delivered novel enabling tools that assist with site screening, borehole design and post-closure safety assessments. For instance, an automated geological fault mapping and meshing tool was developed that assists with ranking the suitability of potential disposal sites based on proximity to faults. New codes were developed for better representation of fault zones in 2D/3D numerical flow and transport models, while also being more efficient to execute. Post-closure safety assessments tested the sensitivity of long-term safety with respect to disposal depth, rock permeability and sorption. Heat transport calculations explored the sensitivity of temperature evolution within the borehole to parameters such as heat load, borehole depth, geothermal gradients and rock thermal conductivity. For verification of host rock tightness while also demonstrating the absence of recent groundwater, a new noble gas analytical facility has been established for measuring rare noble gases in mineral fluid inclusions as indicators of very old pore fluids.

Effect of thermo-hydro-mechanical coupling on the evolution of stress in the concrete liner of an underground drift in the Cigéo project

The French National Radioactive Waste Management Agency (Andra) is in charge of studying the disposal of high-level and intermediate-level long-lived waste (HLW and ILW-LL) in a deep geological repository (Cigéo project) within the host formation is the Callovo-Oxfordian claystone (COx). The heat emitted from waste packages induces a thermo-hydro-mechanical (THM) coupling within the structural elements and the host rock. This study focuses on the behavior of the concrete lining of an ILW-LL cell subjected to THM loading during its construction and operational phases. The mechanical behavior of the host rock is represented by an elasto-visco-plastic model taking into account the anisotropies in stiffness and strength. The coupled THM formulation is based on the Biot theory. Different simulations including full THM coupling and HM coupling with or without creep behavior of COx claystone have been performed to show the effect of the thermal load (generated by the waste packages), of the water seepage and of the creep strain of the host rock on the stress evolution in the concrete liner. The results show the preponderant role of the creep strain of COx claystone on the stress state of the liner during the operational phase, while the effect of the heat loading is moderate and that of the seepage is not significant.

Study of Neutron Background in order to Improve Radioactive Waste Drum Characterization

A usual way of radioactive waste drums characterization combines gamma spectrometry measurements with passive neutron measurements. The CEA facility for waste drum characterization hereby described has been operated for more than 30 years. In this framework, a large variety of waste drums has been characterized in terms of spectra, densities, materials and radioactivity levels. As the facility was first dedicated to measure Intermediate-Level Long-lived Waste, the neutron background was not significant compared to expected neutron emitters from waste packages. These last years, Dismantling and Decommissioning operations have been advanced in this CEA site to the point where they are now associated with mostly Very Low Level Waste production. Therefore, neutron background is becoming significant. Using the large variety of past characterized drums brings the opportunity to study this background. The present study has been led over a sample of almost 1500 drums over a wide waste density range. These drums have been selected over the last 20 years by taking into account only one criterion: without any expected neutron emitters from the waste itself. This work first presents the technical settings of the measurement facility before describing the raw data of the measurements. Next, a statistical study over raw data enables to better acknowledge the neutron spallation background behavior in terms of time, density and materials. Ensues a way of using this new knowledge in order to improve how to take into account neutron spallation background in passive neutron measurements of packages of low actinides activities and high densities.

Alkaline hydrolysis of radio-oxidized aliphatic polymers

Organic molecules are part of the Long-Lived Intermediate-Level Waste (LL-ILW) packages. These packages are planned to be delivered at the French deep geological repository, consequently their degradation has to be thoroughly understood in order to support the safety studies. In a first period, these organic molecules, which include polymers, are radio-oxidized because of their close proximity with radionuclides. In a second period, water will diffuse inside the site up to the packages, leading to the alkaline hydrolysis of the radio-oxidized polymers. The objective of this study is to evaluate the parameters having a major influence on these radio-oxidized polyolefins hydrolysis at high pH. Aliphatic polymers are chosen because of their relative importance, in quantity, in the French LL-ILW packages.In this work, we have first radio-oxidized aliphatic polymers in conditions that are representative of their ageing in the first period of the LL-ILW packages: we have employed different doses up to 10 MGy, different emitters - or their equivalent -, i.e. γ-rays and swift heavy ions (SHI) irradiation to model α-rays. To model the second period of ageing of these LL-ILW packages, we have hydrolyzed under alkaline conditions and under inert atmosphere these radio-oxidized polymers.We have evaluated the influence of the materials’ composition (the resin chemical structure, presence of charges and fillers) and of the irradiation conditions (irradiation type, linear energy transfer (LET) and dose). We observed that hydrolysis is influenced not only by the molecular structure of the polymer, but also by the presence of inorganic fillers and organic additives. The hydrolysis increases with increasing dose and increasing LET for SHI. The most penalizing conditions were encountered under low LET ionizing rays. This is related to the higher oxidation level induced under γ-rays: the higher the oxidation rate, the higher the polyolefins sensitivity to alkaline hydrolysis.

Temperature and LET effects on radiation-induced modifications in non-perfect polyethylenes

Regarding the nuclear safety, the potential accumulation of explosive gases during storage, transport and final disposal of Intermediate Level Long Lived Wastes (IL-LLW) has to be well understood and precisely evaluated. In this paper, we have investigated the Linear Energy Transfer (LET) effect that takes into account the different emitters present in the packages, but also the temperature impact in transport safety cases. We have focused our study on polyethylene, as this polymer presents one of the highest hydrogen radiation chemical yield (G0(H2)), and because of the explosive and inflammable nature of this gas. Chemical defects included in the polymer in the form of carbonyl (C=O) and vinyl (C=C) groups were explored. Depending on the atmosphere of irradiation, they are the main defects formed in polyethylene under irradiation and they are known to be effective energy and radical scavengers. Characterization of the chemical structure of pristine materials has been done. Hydrogen radiation chemical yields were quantified after polymers irradiation. Additionally, the carbon monoxide release was measured for polyethylenes with C=O moieties. Whatever the type of double bonds present in the polymer and whatever the irradiation conditions, hydrogen emission decreases compared to neat materials. Double bonds are effective scavenging groups to which energy is transferred regardless the irradiation nature. Underlying phenomena are explained in this work.

Hydration of Eurobitum bituminized waste under free swelling conditions: osmosis-induced swelling and NaNO3 leaching

ABSTRACTIn Belgium, Eurobitum bituminized radioactive waste is an important intermediate-level long-lived waste form that contains, besides bitumen and radionuclides, large amounts of soluble salts (mainly NaNO3). Geological disposal in a water-saturated sedimentary formation will induce swelling of Eurobitum due to water uptake by the hygroscopic salts embedded in a highly efficient semi-permeable bitumen membrane. Initially, while there is still free space in the primary waste containers, free swelling will occur. To improve our understanding of the water uptake processes under free swelling conditions, water uptake and leaching tests are being performed in which inactive Eurobitum was contacted at the top with 0.1 M KOH (simplified representation of young cement water) without effective stress on the sample. Under these test conditions, the evolution of the water uptake and NaNO3 leaching processes shows several stages: (1) initial fast ingress of water into and leaching of NaNO3 from pores close to the surface; (2) slower progression of the hydration front, the swelling and the leaching of NaNO3, but behaving linearly with the square root of time, indicating that these processes are controlled by diffusion; (3) after more than 2 to 3 years, the leaching, hydration and swelling rates re-increase, attributed to the formation of interconnecting pores and (micro)cracks in the bitumen matrix in response to its increasing deformation.

Radioactive waste management in France: safety demonstration fundamentals

The main challenge in development of the safety case for deep geological disposal is associated with the long periods of time over which high- and intermediate-level long-lived wastes remain hazardous. A wide range of events and processes may occur over hundreds of thousands of years. These events and processes are characterised by specific timescales. For example, the timescale for heat generation is much shorter than any geological timescale. Therefore, to reach a high level of reliability in the safety case, it is essential to have a thorough understanding of the sequence of events and processes likely to occur over the lifetime of the repository. It then becomes possible to assess the capability of the repository to fulfil its safety functions. However, due to the long periods of time and the complexity of the events and processes likely to occur, uncertainties related to all processes, data, and models need to be understood and addressed. Assessment is required over the lifetime of the radionuclides contained in the radioactive waste.

Impact of Water Consumption and Saturation-Dependent Corrosion Rate on Hydrogen Generation and Migration from an Intermediate-Level Radioactive Waste Repository

The corrosion of metals contained in intermediate-level long-lived waste (ILW) under reduced chemical condition will lead to the production of hydrogen gas during the post-closure phase of a deep geological repository for radioactive waste. According to previous investigations by Talandier et al. (Proceedings of TOUGH symposium 2006, Berkeley, 2006), the period of concern covers several 1,000 years after closure of a repository in a clay host rock (Callovo-Oxfordian). The limited hydrogen transport efficiency of the host rock will lead to significant saturation of the concrete waste canister pore space and voids with a gas phase and pressure build-up within the emplacement drifts. On the other hand, the water availability is limited as a result of (i) the low permeability of the clay host rock and (ii) the desaturation of the rock mass close to the drift wall due to the ventilation of the drifts during the operational phase of the repository. In former numerical simulations it was assumed that under the reducing chemical conditions prevailing in the repository, the corrosion rate would be a function of the available metal surface and the temperature only. In this paper, simulation results based on new phenomenological functions are presented, which were implemented in TOUGH2. These allow taking into account (i) a water saturation dependency of the hydrogen generation rate, (ii) the water consumption due to the corrosion process, and (iii) the total metal mass available for corrosion. The paper presents results of 1D radial and 2D vertical simulations of a typical cross-section of a waste emplacement drift and the surrounding rock mass. The interactions between water availability from the low permeable clay, the dependency of the hydrogen generation rates from the water saturation of the waste package and the hydrogen migration in the host rock are demonstrated.

Physico-chemical investigation of clayey/cement-based materials interaction in the context of geological waste disposal: Experimental approach and results

Within the concepts under study for the geological disposal of intermediate-level long-lived waste, cement-based materials are considered as candidate materials. The clayey surrounding rock and the cement-based material being considered differ greatly in their porewater composition. Experiments are conducted on the diffusion of solutes constituting those porewaters in a confined clay/cement composite system using cells. The test temperature was set at 25 °C and 2, 6 and 12 months. Results supply new information: carbonation is low and not clog the interface. Such absence of carbonation allows for the diffusion of aqueous species and, thus, for the degradation of the cement paste and the illitisation of illite/smectite interstratifications. The cement material is subjected to a decalcification: portlandite dissolution and a CaO/SiO 2 reduction in the calcium silicate hydrate. The sulphate in diffusion induces non-destructive ettringite precipitation in the largest pores. After 12 months, about 800 µm of cement material is concerned by decalcification.

Nuclear Power – A Panacea for Future Energy Needs?

Each week seems to bring further evidence that the Earth is warming at a faster rate than previously estimated. Pressure is building to replace power sources that emit carbon dioxide with those tha...

Specific transport and storage solutions for residues return programmes

AbstractThe treatment of used nuclear fuel, performed at AREVA's La Hague plant, allows recovering uranium 95% and plutonium 1% for recycling, the remaining 4% being considered as ultimate waste that can be sorted into two categories: high level activity waste (HLW) which is vitrified, and long-lived intermediate level waste (ILW) composed of structural elements of used nuclear fuel which is compacted. Whether vitrified or compacted, the waste is conditioned in the same universal and multipurpose container, named the Universal Canister. The resulting residue is named CSD-V for vitrified waste and CSD-C for compacted waste; they both remain property of the utilities and must be returned to countries of origin. In order to transport Universal Canisters in the best technical and economical conditions, TN International designs two kinds of cask solutions for its customers, either for transport only or for dual purpose, storage and transport, depending on the facility. Since the mid-1990s, TN International has...

Anisotropic Diffusion in Layered Argillaceous Rocks: A Case Study with Opalinus Clay

Anisotropic diffusion was studied in Opalinus Clay, a potential host rock for disposal of spent fuel, vitrified high-level waste, and long-lived intermediate-level waste in Switzerland. Diffusion parallel to the bedding was measured using a radial through-diffusion technique and diffusion perpendicular to the bedding bythe classical (planar) through-diffusion technique. The samples used were from Mont Terri (MT) and from Benken (BE). Diffusion of HTO, 36Cl-, and 22Na+ parallel and perpendicular to the bedding was studied under confining pressures of 7 MPa (MT) and 14 MPa (BE). The results indicate that diffusion parallel to the bedding is faster than diffusion perpendicular to the bedding by a factor of 4-6 for the three radionuclides, indicating that the Opalinus Clay is anisotropic. This might be explained by smaller path lengths (tortuosity) for species diffusing parallel to the fabric. The degree of anisotropy is slightly smaller for Opalinus Clay from Mont Terri than from Benken. This is due to the lower overburden pressure in Mont Terri resulting in a lower preferential orientation of the clay platelets.

Development of a sorption data base for the cementitious near-field of a repository for radioactive waste

The migration of radionuclides within a repository for radioactive waste is retarded due to interaction with the engineered barrier system. Sorption processes play a decisive role in the retardation of radionuclides in the repository environment, and thus, the development of sorption data bases (SDBs) is an important task and an integral part of performance assessment. The methodology applied in the development of a SDB for the cementitious near-field of a repository for long-lived intermediate-level waste is presented in this study. The development of such a SDB requires knowledge of the chemical conditions of the near-field and information on the uptake process of radionuclides by hardened cement paste. The principles upon which the selection of the “best available” laboratory sorption values is based are outlined. The influence of cellulose degradation products, cement additives and cement-derived colloids on the sorption behaviour of radionuclides is addressed in conjunction with the development of the SDB.

Building the Safety Case for a Disposal Facility for Spent Fuel, HLW and Long-lived ILW in Switzerland

ABSTRACTThis paper describes a generic methodology for building the safety case for a geological repository, which is currently being applied to a possible facility for spent fuel, vitrified highlevel waste and long-lived intermediate-level waste in the Opalinus Clay of Northern Switzerland. The methodology involves:1. the identification of certain basic disposal principles,2. the choice of a disposal system, via a flexible repository development strategy,3. the derivation of the system concept, based on current understanding of the phenomena that characterise, and may influence, the disposal system and its evolution,4. the derivation of a safety concept, based on reliable, well understood and effective pillars of safety,5. the illustration of the radiological consequences of the disposal system through the definition and analysis of a wide range of assessment cases, and6. the compilation of the arguments and analyses that constitute the safety case, as well as guidance for future stages of the repository programme. A range of measures, including audits, are used to promote completeness of the phenomena considered in the safety case, and to avoid inadvertent bias.

Preliminary Safety Studies for Conventional Spent Fuel, MOX, and Vitrified High-Level Waste

This paper discusses the results of post-closure safety studies for two different waste streams that, according to current Swiss waste-management concepts, may be co-disposed in a single deep geological repository. The waste streams are:• directly disposed spent UO2 and mixed-oxide (MOX) fuel• vitrified high-level waste from the reprocessing of spent fuelThe inventories are based on a consideration of the anticipated arisings from nuclear power generation in Switzerland. A part (at least) of these arisings will be reprocessed, with the resultant vitrified high-level waste and long-lived intermediate-level waste returned to Switzerland. The Swiss electricity utilities have placed contracts with BNFL (UK) and COGEMA (France) for the reprocessing of about one third of the total arisings of spent fuel from nuclear power generation, assuming an electricity production scenario of 120 GWa(e). The decision as to whether to reprocess the remainder is currently left open, implying that up to two thirds of the arisings could be in the form of unreprocessed spent fuel for direct disposal. For the purposes of the present study, however, it is assumed that the arisings from the full 120 GWa scenario will be either directly disposed or reprocessed.

Nuclear waste management in Switzerland / Nukleare Entsorgung in der Schweiz

The Swiss waste management concept foresees geological disposal for all categories of radioactive waste. Two repositories are planned, one for short-lived low- and intermediate-level wastes (L/ILW), and one for high-level waste and long-lived intermediate-level waste (HLW/TRU) or spent fuel elements. The Wellenberg in Central Switzerland was proposed as the site for the L/ILW repository. There are three priorities for the current decade: providing centralised interim storage facilities, commencing construction of a L/ILW repository, and identifying a site for a HLW/TRU repository