High-level Waste Repository Research Articles

Corrosion processes of C-steel in long-term repository conditions

ABSTRACTIn the context of the French repository for high-level wastes, carbon steel (C-steel) materials are currently considered among the most cost effective and reliable materials for the prediction of long-term corrosion behaviour. An in situ experiment was conducted at the Meuse/Haute Marne underground research laboratory to investigate the corrosion behaviour of different grades of C-steel up to a maximum temperature of 85°C. C-steel coupons were placed in a two-phase test chamber containing liquid and gas emanating from the Callovo-Oxfordian host rock, at given times and for specific durations, under anoxic conditions. After their exposure in the test chamber, various characterisations were carried out. Overall, after 36 months, the results highlight a potential acidic transient. This could promote high corrosion rates of C-steel. Therefore, an alternative concept has been considered since 2014, which involves the addition of a cement–bentonite-based grout in order to buffer the possible acidic transient.This paper is part of a supplement on the 6th International Workshop on Long-Term Prediction of Corrosion Damage in Nuclear Waste Systems.

Implementation of the strain hardening model into buffer material in high level waste repository

Engineered Barrier System (EBS) primarily provides a protective cover for the High Level Waste (HLW) canisters. In addition, it functions as an efficient hydraulic and chemical barrier system. For in situ disposal applications, the ideal buffer material is bentonite and it is fabricated into blocks to build EBS. Currently, traditional methods introduce significant errors in constitutive modeling of those modeled blocks. In this paper, the particular failure laws and stress-strain laws of bentonite blocks were furnished from a lot of uniaxial and triaxial tests. Also, it showed that the stress-strain relations given by the traditional constitutive relations did not correspond to the failure forms of bentonite specimens in the experiments. On this basis, the relations between internal friction angle, cohesion and plastic strain were analyzed so that a new strain hardening model was established under the framework of Mohr-Coulomb criterion. In the end, a series of 3D computation of HLW disposal were carried out using the software FLAC3D, where the proposed strain hardening model led to satisfactory prediction of the buffer materials.

The Issue of Underground Depositing of High Radioactive Waste

Nowadays manipulation and depositing of high-level radioactive waste has become the most important issue, which needs to be solved. High-level radioactive waste consists mainly of spent fuel elements from nuclear power plants, which cannot be deposited for long time in surface repositories in the same way as it is possible in case of low and medium level radioactive waste. The most effective and safe solution in longer time horizon seems to be deep geological repository of high level waste. In this process of deposition, large amount of specific conditions needs to be taken into account while designing the whole underground complex, because the materials and structures must fulfil all necessary requirements. Then adequate safety will be ensured.

Effect of pH on Plutonium Migration Behavior in Compacted Bentonite

ABSTRACTIn disposing of high-level radioactive waste, the drop in pH in the repository as the iron overpack corrodes must be considered. Plutonium migration behavior may be affected by the pH of pore water in compacted bentonite barriers in high-level waste repositories. To examine the effect of pH on migration behavior, H-bentonite was prepared by treating Japanese Na-bentonite, Kunipia-F, with hydrochloric acid. Diffusion experiments were performed with mixtures of Na- and H-bentonites. The pH value in the pore water of the water-saturated bentonite mixtures decreased from 8 to 3 as the mixing ratio of H-bentonite increased. Diffusion experiments were carried out by using238Pu then apparent diffusion coefficients were determined from the plutonium distribution in the specimens. The apparent diffusion coefficients were on the order of 10-13to 10-12m2/s at pH values lower than 4, whereas they were less than 10-14m2/s at pH values higher than 6.5. These results indicate that plutonium diffused faster as Pu3+or PuO22+due to disproportionation at lower pH while plutonium could be retarded as Pu(OH)40by sorption on bentonite at higher pH.

Numerical modeling of colloid-facilitated radionuclide decay chain transport in a coupled fracture–matrix system

A numerical model is developed to simulate the colloid-facilitated transport of radionuclides in a single fracture-host rock coupled system with planar geometry. Such a system may be encountered in case of leaching and transport of radionuclides from a high-level radioactive waste repository in deep geological formations. The presence of colloids in the system reduces the retardation effects on the radionuclide migration and can change the concentration distribution of radionuclides in the system significantly. The processes such as advection, dispersion, surface sorption, radioactive decay and diffusive loss to the host rock are considered for transport in the fracture, whereas dispersion, adsorption and radioactive decay are included for transport in the host rock. Since the radionuclides to be disposed into the high-level waste repository may include a decay chain of progeny, the performance assessment methodology must incorporate the ingrowth of progeny with time to carry out realistic estimates. The inclusion of decay chain buildup and transport is essential to avoid underestimation with respect to the environmental impact assessment for the repository. Hence, the numerical model based on implicit finite difference scheme is developed with the inclusion of decay chain transport. The model is verified by comparing with well-known analytical solutions. The application of the model is demonstrated by applying it to a five-member decay chain of 238U. With increase in the concentration of colloids, the concentrations of the radionuclides in the fracture water increase and also reach the steady state at an earlier time. The concentrations of radionuclides almost become steady throughout the distance of 200 m in the fracture as the colloid concentration reaches 5 kg m−3 or more. Sensitivity analysis is carried out with respect to various model parameters such as dispersion coefficient in fracture, dispersion coefficient in the porous host rock, porosity of the host rock and distribution coefficient of radionuclide with respect to the mobile colloids. The developed model can be a useful tool for the performance assessment of high-level radioactive waste repositories.

Gas migration modelling in geological repository module in clay formation and sensitivity analysis

Corrosion of steel canisters disposed of in a geological repository for high-level waste leads to generation and accumulation of hydrogen gas which may significantly affect long-term safety of the repository. Numerical modelling can be used to predict the likely hydraulic evolution of such a disposal facility and to estimate the influence of generated gas on repository safety.This article presents a two-phase flow (not taking into account mechanical coupling) gas migration model developed in the original version of the computer code TOUGH2-MP (USA). An important and new aspect of this model is its capability to introduce a network of very thin interfaces in three-dimensional analysis of gas migration. This model was verified in the international benchmark exercises in the recent European Commission project FORGE (Fate Of Repository GasEs). The analysis was focused on the results of a second benchmark exercise, i.e. hydrogen gas behaviour in a part of a geological repository (module) situated in clay formation, and sensitivity analysis. The results showed that the interfaces, backfill and excavation disturbed zone are the main migration and accumulation media for gaseous hydrogen in the repository module. The peak pressure in the module did not exceed the geological environment pressure at a depth of 500m, and is not sufficient to disturb mechanical stability and functionality of the system of engineered and natural barriers.

Dismantling of the EB experiment: Experimental research on the retrieved GBM and bentonite blocks

The Engineered Barrier Emplacement Experiment in Opalinus Clay (EB experiment) was a full-scale test for the demonstration of a new concept of high-level waste (HLW) repositories in horizontal drifts in the Opalinus Clay formation. After 10.5 years of hydration, the EB experiment was dismantled in autumn 2012. Samples obtained from the granular bentonite material (GBM), and bentonite blocks were sent to a laboratory for further analysis. The bentonite samples analyzed at the Laboratory of Mechanic of Lille (LML) were obtained from the CMT1, CMT2, CMT3 and RW sections. Their physical states were determined, as were their effective gas permeability and swelling capacity at different relative humidity (RH) levels. The results indicate that the water contents of the GBM determined in the laboratory ranged between 25.63% and 44.88% and that the dry densities ranged between 1.13 and 1.44g/cm3. The blocks had water contents similar to (or slightly higher than) those of the GBM, and their dry densities had decreased from an initial value of 1.69g/cm3 to values close to 1.30g/cm3, which were similar to the average values found in the GBM. The effective gas permeabilities of the GBM samples were within the range of 1.50×10−22m2 and 1.03×10−17m2, whereas, the corresponding values of the samples obtained from the blocks were between 2.20×10−21m2 and 5.12×10−21m2. The permeability values are primarily related to the dry densities and water contents of the samples. Contact with the Opalinus clay (OPA) does not appear to affect the saturation of the GBM, which is not the case for the samples near the concrete plug. The water retention characters of the GBM and blocks were extremely close, which indicates the homogeneity of the materials and a good control in their preparation.

Shrinkage behaviour and related corrosion performance of low-pH cementitious materials based on OPC or CAC

Prior to using low-pH cementitious materials in underground repositories for high level waste, the characteristics determining their long-term durability must be analysed in depth. In this sense, different shrinkage tests have been made on mortar and concrete specimens using low-pH cement formulations based on ordinary portland cement (OPC) or calcium aluminate cement (CAC), with high mineral admixtures contents. They showed similar autogenous shrinkage than samples without mineral admixtures but higher drying shrinkage when materials based on OPC with high silica fume contents were considered. Besides, as the use of reinforced concrete could be required in underground repositories, the susceptibility of reinforcement to corrosion when using low-pH cementitious materials based on OPC was analyzed, considering carbon steel and galvanized steel. In the formers corrosion was detected due to the low pore solution pH but any problem was detected when galvanized reinforcement were used.

Sorption of plutonium on geological materials associated with a Chinese radioactive waste repository: influence of pH

The aim of this study is to characterize plutonium sorption particularly 239Pu distribution coefficients and the influence of pH on sorption processes in saturated geological materials used as a radioactive waste repository in China. The examined geological media is from a southwestern Chinese repository and consists of soil and slates. The backfill material consists of Na-bentonite from Inner Mongolia Gaomiaozi, and will be used at the Beishan high-level waste (HLW) repository, Gansu Province. The results show that the sorption capacity of Pu by geological medium in an acidic environment is low, but significantly increases with increasing pH.

Temperature and tortuosity effect on gas migration in a high-level waste disposal tunnel

AbstractCorrosion of steel canisters, disposed of in a repository for high-level waste (HLW), leads to generation of hydrogen gas for a long period after the repository's closure. The accumulation of hydrogen gas may lead to significant desaturation and unacceptable build-up of pressure in the backfilled disposal tunnels if the gas cannot escape through the low-permeability host rock. Consequently, the investigation of gas migration is of high relevance in the assessment of the repository's performance.In this paper, the results of numerical investigations on gas migration performed using the computer code TOUGH2 (USA) are presented. The objective was to investigate migration of gas generated in a single disposal tunnel of a conceptual geological repository in a clay formation, which was suggested for benchmark studies in the European Commission project FORGE (Fate Of Repository GasEs). The analysis was focused on evaluation of the impact of an initial temperature in the repository and of different tortuosity models on gas migration. It was revealed that gas migration results were dependent on tortuosity model, while temperature variation in the repository had minor impact.

Evaluation of temperature-induced effects on safety-relevant properties of clay host rocks with regard to HLW/SF disposal

AbstractDBE TECHNOLOGY, BGR and GRS are developing a methodology to demonstrate the safety of a repository for high-level waste and spent fuel (HLW/SF) in clays according to the requirements of the German regulating body. In particular, these requirements prescribe that the barrier effect of host rocks must not be compromised by a thermal impact resulting from HLW/SF emplacement. To substantiate and quantify this requirement, we carried out a literature survey of research on thermally-induced changes on clay properties. Effects thus compiled can be divided into thermo-hydro-mechanical and chemical-biological-mineralogical effects and were analysed with regard to their relevance to the integrity of clay host rocks. This analysis identified one effect of major influence within each group: thermal expansion and compaction as well as results of microbial activities. Importantly, it further revealed that a moderate temperature increase above 100°C cannot be expected to compromise the integrity of the geological barrier according to the current knowledge state. Evidence is presented in this paper that temperature increases up to 150°C can actually contribute to an improved performance of a radioactive waste repository by increasing the consolidation of the clay and sterilizing the repository's near-field to depress the deteriorative microbial effects. A quantitative temperature criterion for thermal impact of HLW/SF on clay host rocks is accordingly proposed.

Nuclear Subsidies and Proliferation: Clarifying the Link

ABSTRACTIn March 2015, the South Australian state government established a royal commission to investigate the financial, social, technical, diplomatic, and nonproliferation benefits and risks of expanding its nuclear industry, including activities related to uranium mining; enriching, reprocessing, and fabricating nuclear fuels for both domestic use and export; producing nuclear power; and storing radiological waste, including foreign spent reactor fuel. Given its enormous uranium reserves and current mining activities, some Australians have argued that Australia could benefit financially by expanding the mining sector and by adding value to its uranium exports by enriching the material and fabricating it into reactor fuel assemblies. Others have maintained that Australia can realize significant economic benefits by recycling and storing foreign spent fuel and producing carbon-free nuclear power. In the end, the commission recommended that Australia consider opening up a high-level waste repository to take in foreign spent fuel. It did not recommend any other nuclear activities at this time. The following viewpoint is based on testimony I delivered to the commission on the nuclear weapon proliferation implications of the proposed activities. If Australia wants to avoid the temptation of selling nuclear goods to states that might use these goods to make bombs, it should only consider new nuclear activities that can be entirely financed by the private sector and turn a profit without having to resort to foreign sales. This policy would also enable Australia to set an important, new international nonproliferation standard.

고준위폐기물처분장 공학적방벽의 갭채움재 기술현황

고준위폐기물처분장에서 갭채움재는 완충재와 뒷채움재의 성능을 좌우하는 중요한 공학적방벽의 구성요소이다. 본 논문에서는 갭채움재에 대한 해외 기술현황을 조사하고, 이를 통하여 갭채움재의 개념, 제조기술, 성형특성, 설치기술에 대한 연구결과들을 정리하였다. 갭채움재 개념은 처분방식과 처분개념에 따라서 나라마다 약간씩 차이가 있었다. 갭채움재 물질로는 대부분 벤토나이트를 사용하였고, 충전제로 점토를 사용하였다. 갭채움재는 펠렛, 과립상, 또는 펠렛-과립상 혼합물의 형태로 사용되었다. 갭채움재 펠렛 제조에는 정압축, 롤러압축, 압출-컷팅 방법 등이 사용되었으며, 이 중, 실험과 실제 현장에서의 펠렛 소요량을 감안하여 많은 나라들이 롤러압축방법과 압출-컷팅방법에 대한 기술 확보에 집중하였다. 펠렛 성형특성 실험결과, 펠렛의 건조밀도와 건전성은 수분함량, 구성물질, 제조방법, 펠렛 크기에 민감하였고, 제작 시 압축하중에는 상대적으로 덜 민감하였다. 갭채움재의 설치방법으로는 수직처분공 완충재 갭에서는 부어넣기(pouring) 방법, 붓고 다지기(pouring and tamping) 방법, 진동을 주며 부어넣기(pouring with vibration) 방법 등이 시도되었으며, 수평처분공 완충재와 처분터널의 뒷채움재 갭에서는 숏크리트 기술을 이용한 불어넣기(blowing by use of shotcrete technology) 방법과 오거를 이용한 정치 및 다지기(auger placement and compaction) 방법 등이 시도되었다. 그러나 이 방법들은 아직 기술적으로 초기단계에 있어 앞으로도 계속적인 연구가 이루어질 것으로 예상되었다. In a high-level waste repository, the gap fill of the engineered barrier is an important component that influences the performance of the buffer and backfill. This paper reviewed the overseas status of R&D on the gap fill used engineered barriers, through which the concept of the gap fill, manufacturing techniques, pellet-molding characteristics, and emplacement techniques were summarized. The concept of a gap fill differs for each country depending on its disposal type and concept. Bentonite has been considered a major material of a gap fill, and clay as an inert filler. Gap fill was used in the form of pellets, granules, or a pellet-granule blend. Pellets are manufactured through one of the following techniques: static compaction, roller compression, or extrusion-cutting. Among these techniques, countries have focused on developing advanced technologies of roller compression and extrusion-cutting techniques for industrial pellet production. The dry density and integrity of the pellet are sensitive to water content, constituent material, manufacturing technique, and pellet size, and are less sensitive to the pressure applied during the manufacturing. For the emplacement of the gap fill, pouring, pouring and tamping, and pouring with vibration techniques were used in the buffer gap of the vertical deposition hole; blowing through the use of shotcrete technology and auger placement and compaction techniques have been used in the gap of horizontal deposition hole and tunnel. However, these emplacement techniques are still technically at the beginning stage, and thus additional research and development are expected to be needed.

Three-dimensional numerical simulation by upscaling of gas migration through engineered and geological barriers for a deep repository for radioactive waste

Abstract This paper presents the results of a benchmark study that compares a number of numerical models applied to a specific problem in the context of hydrogen flow and transport in a nuclear waste repository. The processes modelled are two-phase (water and hydrogen) immiscible compressible two-component transient flow in a heterogeneous porous medium under isothermal conditions. The three-dimensional (3D) model represents a module of a repository for high-level waste in a clay host rock. An upscaling technique and a vertex-centred finite-volume method are employed to yield very accurate solutions. Since the full range of results required in the benchmark is too large to be displayed in this paper, we focus on the evolution of the pressures, the saturations, the fluxes and the comparison of the numerical results with the other participants. A homemade C++ upscaling code and the parallel multiphase flow simulator DuMu X have been adopted for this study.

Study of cement-grout penetration into fractures under static and oscillatory conditions

Grouting of the rock surrounding high-level waste (HLW) can serve to minimize groundwater flow around it and thereby to retard erosion of waste-embedding clay (buffer) and transport of possibly released radionuclides. Earlier attempts have shown the efficiency of superimposing the injection pressure with oscillations for bringing cement-rich grouts into narrow fractures using organic superplasticizers. However, these are short-lived and can produce radionuclide-bearing organic colloids, and should be replaced by inorganic agents. Portland cement in grouts is not long lived and low-pH cements are preferable as is also reduction of the cement content to an absolute minimum. The present study describes the composition and performance of candidate grouts in laboratory experiments with injection into plane-parallel slots with different aperture. The study included development of a simple and quick method for estimating the viscosity on the construction site for adapting the grout recipe to the injectability of the rock. A simple theoretical model for predicting grout penetration gives fair agreement with laboratory data. The longevity of the grout under various conditions is believed to be sufficient for use in HLW repositories implying waste placement in very deep holes.

Coupled thermal-hydro analysis of unsaturated buffer and backfill in a high-level waste repository

The buffer and backfill are the major components of an engineered barrier system (EBS) for a high level waste (HLW) repository. Reliable EBS performance assessment requires the delicate numerical modeling of the buffer and backfill. This study carried out the sensitivity analysis of thermal-hydro (TH) model parameters, and based on its results, the coupled TH analysis of an unsaturated buffer and backfill in an HLW repository. The temperature distribution was sensitive to a change in the degree of saturation and thus the thermal conductivity, and the degree of saturation distribution was sensitive to a variation in temperature and thus water viscosity. The decay heat of HLW from a canister dissipated out through the buffer and backfill into the surrounding rock. The temperature was higher closer to the canister and was lower farther from the canister. The temperature in the backfill was overall lower than that in the buffer, but both temperatures were approached slowly over a long time. The peak temperature was located at the center of the interface line between the canister and buffer, and was 107°C at the elapsed time of 0.47year. Re-saturation occurred in the order of the backfill and then a buffer as groundwater was intruded from the surrounding rock. The wetting of the backfill was initiated from the wall of the tunnel and the upper wall of deposition hole, and it then proceeded toward the inside. The buffer was wetted from the wall of the deposition toward the canister. The latest re-saturation location was a bit above from the center axis of the canister, at which the re-saturation period of time was estimated to be about 10years.

The influence of stable element inventory on the migration of radionuclides in the vicinity of a high level nuclear waste repository exemplified for 59Ni

A generic modelling approach has been used to estimate the influence of the stable inventory, i.e. stable isotopes in a radioactive waste repository on the migration of radionuclides from waste canisters into the surrounding bentonite or Opalinus Clay. The model radionuclide chosen was bivalent 59Ni(II); the stable isotopes Ni(II), Fe(II), Mn(II), Zn(II) and Cu(II) are considered to be competitive for the same sorption sites on bentonite or Opalinus Clay. A simplified one-dimensional modelling approach in space was used for reactive transport calculations using the MCOTAC code incorporating the 2SPNE SC/CE sorption model. Calculated 59Ni(II) breakthrough curves in bentonite and Opalinus Clay are compared to estimates of the influences of the individual competing metals present in the porewaters.Generally, faster migration, i.e. a reduced sorption, for 59Ni(II) was calculated – up to two orders of magnitude in arrival time at specified locations in the bentonite or Opalinus Clay. This influence is a maximum for highest specified stable isotope concentrations.Fe, Zn and Mn have approximately the same effect on the migration of 59Ni(II); Cu has the potential for a much stronger effect. However, their individual effects at reducing the retardation of 59Ni(II) through sorption competition do not sum up linearly. In the various scenarios calculated, an upper limit for the reduction of the retardation of 59Ni(II) has been assessed for the combined sorption competition influence of all the stable isotopes.Although the calculated scenarios include several simplifications, they cover a wide range of combinations of sorption competition effects of stable isotopes present in a high-level waste repository on the migration of radioactive 59Ni(II). More detailed scenario calculations would be possible if a more detailed “geochemical inventory” of radionuclides and stable isotopes were to become available. Nevertheless, upper limits for the effects of sorption competition of bivalent stable isotopes on the migration of 59Ni in the vicinity of a high-level nuclear waste repository were assessed.

Materials in Nuclear Waste Disposition

Commercial nuclear energy has been used for over 6 decades; however, to date, none of the 30+ countries with nuclear power has opened a repository for high-level waste (HLW). All countries with nuclear waste plan to dispose of it in metallic containers located in underground geologically stable repositories. Some countries also have liquid nuclear waste that needs to be reduced and vitrified before disposition. The five articles included in this topic offer a cross section of the importance of alloy selection to handle nuclear waste at the different stages of waste processing and disposal.

On area-specific underground research laboratory for geological disposal of high-level radioactive waste in China

Underground research laboratories (URLs), including “generic URLs” and “site-specific URLs”, are underground facilities in which characterisation, testing, technology development, and/or demonstration activities are carried out in support of the development of geological repositories for high-level radioactive waste (HLW) disposal. In addition to the generic URL and site-specific URL, a concept of “area-specific URL”, or the third type of URL, is proposed in this paper. It is referred to as the facility that is built at a site within an area that is considered as a potential area for HLW repository or built at a place near the future repository site, and may be regarded as a precursor to the development of a repository at the site. It acts as a “generic URL”, but also acts as a “site-specific URL” to some extent. Considering the current situation in China, the most suitable option is to build an “area-specific URL” in Beishan area, the first priority region for China's high-level waste repository. With this strategy, the goal to build China's URL by 2020 may be achieved, but the time left is limited.

Water Retention and Gas Migration of Two High-Performance Concretes after Damage

In the context of long-term repository of high-level and long-lived nuclear waste, authors investigate different concrete properties related to fluid transport, in order to determine which is able to detect damage earliest. To this purpose, different protocols are tested, which impose progressive damage to two different Andra high-performance concretes (HPCs), based on pure portland (CEMI) or composed cement (CEMV)-type cements. The properties investigated are pore size distributions and porosity [by mercury intrusion porosimetry (MIP)], water retention curves, relative gas permeability, and gas migration properties (gas breakthrough pressure). Gas breakthrough pressure (GBP) is assessed rather than gas entry, by accurately measuring gas presence on the downstream side of a confined sample subjected to slowly increasing gas pressure on its upstream side. From MIP data, authors show that CEMI concrete has smaller porosity but greater pore sizes than CEMV. For CEMI concrete, all damage procedures significantly affect water retention curves and gas breakthrough pressures, yet they have no effect upon the relationship between relative gas permeability and water saturation. For CEMV concrete subjected to low damage levels, gas breakthrough measurements detect damage, whereas water retention and gas permeability do not.