Intermediate Level Nuclear Waste Research Articles

The suitability of a supersulfated cement for nuclear waste immobilisation

Composite cements based on ordinary Portland cement are used in the UK as immobilisation matrices for low and intermediate level nuclear wastes. However, the high pore solution pH causes corrosion of some metallic wastes and undesirable expansive reactions, which has led to alternative cementing systems being examined. We have investigated the physical, chemical and microstructural properties of a supersulfated cement in order to determine its applicability for use in nuclear waste encapsulation. The hardened supersulfated cement paste appeared to have properties desirable for use in producing encapsulation matrices, but the high powder specific surface resulted in a matrix with high porosity. Ettringite and calcium silicate hydrate were the main phases formed in the hardened cement paste and anhydrite was present in excess. The maximum rate of heat output during hydration of the supersulfated cement paste was slightly higher than that of a 9:1 blastfurnace slag:ordinary Portland cement paste commonly used by the UK nuclear waste processing industry, although the total heat output of the supersulfated cement paste was lower. The pH was also significantly lower in the supersulfated cement paste. Aluminium hydroxide was formed on the surface of aluminium metal encapsulated in the cement paste and ettringite was detected between the aluminium hydroxide and the hardened cement paste.

Assessment of the evolution of the redox conditions in a low and intermediate level nuclear waste repository (SFR1, Sweden)

The evaluation of the redox conditions in an intermediate and low level radioactive waste repository such as SFR1 (Sweden) is of high relevance in the assessment of its future performance.The SFR1 repository contains heterogeneous types of wastes, of different activity levels and with very different materials, both in the waste itself and as immobilisation matrices and packaging. The level of complexity also applies to the different reactivity of the materials, so that an assessment of the uncertainties in the study of how the redox conditions would evolve must consider different processes, materials and parameters.This paper provides an assessment of the evolution of the redox conditions in the SFR1. The approach followed is based on the evaluation of the evolution of the redox conditions and the reducing capacity in 15 individual waste package types, selected as being representative of most of the different waste package types present or planned to be deposited in the SFR1. The model considers different geochemical processes of redox relevance in the system. The assessment of the redox evolution of the different vaults of the repository is obtained by combining the results of the modelled individual waste package types.According to the model results, corrosion of the steel-based material present in the repository keeps the system under reducing conditions for long time periods. The simulations have considered both the presence and the absence of microbial activity. In the initial step after the repository closure, the microbial mediated oxidation of organic matter rapidly causes the depletion of oxygen in the system. The system is afterwards kept under reducing conditions, and hydrogen is generated due to the anoxic corrosion of steel. The times for exhaustion of the steel contained in the vaults vary from 5ky to more than 60ky in the different vaults, depending on the amount and the surface area of steel. After the complete corrosion of steel, the system still keeps a high reducing capacity, due to the magnetite formed as steel corrosion product.The redox potential in the vaults is calculated to evolve from oxidising at very short times, due the initial oxygen content, to very reducing at times shorter than 5years after repository closure. The redox potential imposed by the anoxic corrosion of steel and hydrogen production is on the order of −0.75V at pH 12.5. In case of assuming that the system responds to the Fe(III)/magnetite system, and considering the uncertainty in the pH due to the degradation of the concrete barriers, the redox potential would be in the range −0.7 to −0.01V.A Monte-Carlo probabilistic analysis on the rate of corrosion of steel shows that the reducing capacity of the system provided by magnetite is not exhausted at the end of the assessment period, even assuming the highest corrosion rates for steel.Simulations assuming presence of oxic water due to glacial melting, intruding the system 60ky after repository closure, indicate that magnetite is progressively oxidised, forming Fe(III) oxides. The time at which magnetite is completely oxidised varies depending on the amount of steel initially present in the waste package.The behaviour of Np, Pu, Tc and Se under the conditions foreseen for this repository is discussed.

U–Pb, Rb–Sr, and U-series isotope geochemistry of rocks and fracture minerals from the Chalk River Laboratories site, Grenville Province, Ontario, Canada

As part of the Geologic Waste Management Facility feasibility study, Atomic Energy of Canada Ltd. (AECL) is evaluating the suitability of the Chalk River Laboratories (CRL) site in Ontario, situated in crystalline rock of the southwestern Grenville Province, for the possible development of an underground repository for low- and intermediate-level nuclear waste. This paper presents petrographic and trace element analyses, U–Pb zircon dating results, and Rb–Sr, U–Pb and U-series isotopic analyses of gneissic drill core samples from the deep CRG-series characterization boreholes at the CRL site. The main rock types intersected in the boreholes include hornblende–biotite (±pyroxene) gneisses of granitic to granodioritic composition, leucocratic granitic gneisses with sparse mafic minerals, and garnet-bearing gneisses with variable amounts of biotite and/or hornblende. The trace element data for whole-rock samples plot in the fields of within-plate, syn-collision, and volcanic arc-type granites in discrimination diagrams used for the tectonic interpretation of granitic rocks.Zircons separated from biotite gneiss and metagranite samples yielded SHRIMP-RG U–Pb ages of 1472±14 (2σ) and 1045±6Ma, respectively, in very good agreement with widespread Early Mesoproterozoic plutonic ages and Ottawan orogeny ages in the Central Gneiss Belt. The Rb–Sr, U–Pb, and Pb–Pb whole-rock errorchron apparent ages of most of the CRL gneiss samples are consistent with zircon U–Pb age and do not indicate substantial large-scale preferential element mobility during superimposed metamorphic and water/rock interaction processes. This may confirm the integrity of the rock mass, which is a positive attribute for a potential nuclear waste repository. Most 234U/238U activity ratios (AR) in whole rock samples are within errors of the secular equilibrium value of one, indicating that the rocks have not experienced any appreciable U loss or gain within the past 1Ma. However, 234U/238U AR in fracture mineral samples collected down to borehole lengths of about 740m deviate from the secular equilibrium value and 234U/238U model ages calculated for fracture mineral samples showing excess 234U range from 593 to 1415ka, thus providing evidence of fracture flow in the associated bedrock during the past 1.5Ma. Rare earth element patterns are variable in fracture-filling calcites and Fe oxides/hydroxides but are similar to those observed in associated whole-rock samples. The observed Ce anomalies are very small (CeN/CeN∗≈1), do not vary with depth, and, therefore, do not contain evidence that the studied fracture minerals precipitated from oxidizing waters at the conceptual depth of a repository.

Corrosion of steel drums containing cemented ion-exchange resins as intermediate level nuclear waste

Exhausted ion-exchange resins used in nuclear reactors are immobilized by cementation before being stored. They are contained in steel drums that may undergo internal corrosion depending on the presence of certain contaminants. The objective of this work is to evaluate the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins with different aggressive species. The corrosion potential and the corrosion rate of the steel, and the electrical resistivity of the matrix were monitored for 900days. Results show that the cementation of ion-exchange resins seems not to pose special risks regarding the corrosion of the steel drums.

The Effect of γ-radiation on Mechanical Properties of Model UK Nuclear Waste Glasses

ABSTRACTThe effect of γ-radiation on the mechanical properties of model UK intermediate and high level nuclear waste glasses was studied up to a dose of 8 MGy. It was determined that γ-irradiation up to this dose had no measurable effect upon the Young’s modulus, shear modulus, Poisson’s ratio, indentation hardness, or indentation fracture toughness. The absence of measurable radiation induced changes in mechanical properties was attributed to redox mediated healing of electron-hole pairs generated by γ-irradiation by multivalent transition metal ions, in particular the Fe3+ - Fe2+ couple.

Corrosion susceptibility of steel drums to be used as containers for intermediate level nuclear waste

The present work is a study of the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins contaminated with different types and concentrations of aggressive species. A special type of specimen was manufactured to simulate the cemented ion-exchange resins in the drum. The evolution of the corrosion potential and the corrosion rate of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 900 days. The aggressive species studied were chloride ions (the main ionic species of concern) and sulphate ions (produced during radiolysis of the cationic exchange-resins after cementation). The work was complemented with an analysis of the corrosion products formed on the steel in each condition, as well as the morphology of the corrosion products. When applying the results obtained in the present work to estimate the corrosion depth of the steel drumscontaining the cemented radioactive waste after a period of 300 years (foreseen durability of the Intermediate Level Radioactive Waste facility in Argentina) , it is found that in the most unfavourable case (high chloride contamination), the corrosion penetration will be considerably lower than the thickness of the wall of the steel drums.

The potential impact of anaerobic microbial metabolism during the geological disposal of intermediate-level waste

AbstractMicrobial metabolism has the potential to control the biogeochemistry of redox-active radionuclides in a range of geodisposal scenarios. In this study, sediments from a high pH lime workings site were incubated under carefully controlled anaerobic conditions, at a range of alkali pH values with added electron donors and electron acceptors, to explore the limits and rates of bioreduction in a sediment system analogous to intermediate-level nuclear waste. There was a clear succession in the utilization of electron acceptors (in the order nitrate > Fe(III)-citrate > Fe(III) oxyhydroxide > sulfate), in accordance with calculated free energy yields and Eh values over the pH range 10–12. The rate and extent of bioreduction decreased at higher pH, with an upper limit for the processes studied at pH 12. The biochemical limits for such processes are discussed, alongside the potential impact of such forms of microbial metabolism on the solubility of a range of redox active radionuclides that will feature heavily in the safety case for the geological disposal of intermediate-level nuclear waste.

Radial Flow Permeability Testing of an Argillaceous Limestone

Argillaceous Lindsay limestone is the geologic storage formation that will be encountered at the site for the construction of a deep ground repository in Ontario, Canada, for the storage of low to intermediate level nuclear waste. The permeability of the Lindsay limestone is a key parameter that will influence the long-term movement of radionuclides from the repository to the geosphere. This paper describes the use of both steady-state and transient radial flow laboratory tests to determine the permeability of this argillaceous limestone. The interpretation of the tests is carried out using both analytical results and computational models of flow problems that exhibit radial symmetry. The results obtained from this research investigation are compared with the data available in the literature for similar argillaceous limestones mainly found in the Lindsay (Cobourg) formation. The experiments give permeabilities in the range of 1.0 × 10(-22) to 1.68 × 10(-19) m(2) for radial flows that are oriented along bedding planes under zero axial stress. The factors influencing transient pulse tests in particular and the interpretation of the results are discussed.

Corrosion of Steel Drums Containing Cemented Ion-Exchange Resins as Intermediate Level Nuclear Waste

ABSTRACTIon-exchange resins are used for purification of radioactive liquid waste from nuclear reactors. After exhaustion, resins become intermediate level radioactive waste to be managed. They have to be immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability for safety requirements. The immobilized resins are thus contained in steel drums that can undergo internal corrosion depending on the presence of certain contaminants. This work shows an study of the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins with different types and contents of aggressive species. Results show that the corrosion depth of the steel drums after a period of 300 years (foreseen life-span of the radioactive waste disposal facility), in the most unfavorable case (high chloride contamination), will be considerably lower than the thickness of the wall of the drums.

Preliminary Investigation on Atmospheric-Induced Stress Corrosion Cracking of British Intermediate Level Nuclear Waste Containers: Residual Stress Analysis and Four Point Bend Exposure Test

There are approximately 45,000 Intermediate Level Nuclear Waste (ILW) containers currently stored in the UK at Sellafield and elsewhere. These stainless steels containers should last up to 150 years of surface storage. In this study the general structural integrity of 500L ILW containers were investigated. The associated degradation issue under ambient conditions known as atmospheric-induced stress corrosion cracking (AISCC) has been studied. Preliminary investigation included profiling of residual stress distribution in parent and weld specimens, atmospheric exposure testing under various salts. So far residual stress profile shows high degree of complexity. No cracks have been observed until this moment in time.

Carbonation durability of blended cement pastes used for waste encapsulation

Blended cement pastes are currently used for encapsulation of low level and intermediate level nuclear waste in the UK. However, there is still little information on the long-term durability of those mixes to some chemical attacks. Accelerated testing may predict the long-term durability or at least help the selection of more durable formulations. In this work, blended blastfurnace slag (BFS)/Portland cement (OPC) pastes containing 60, 75 and 90% BFS and pulverised fuel ash (PFA)/OPC pastes with 40, 55 and 75% PFA were cured at 20 and 60°C for 90 days then submitted to natural and accelerated carbonation (5% CO2). The effects of the curing temperature as well as the OPC replacement level on the carbonation ratio are presented. Results showed a good correlation between natural and accelerated carbonation for the pastes studied. Carbonation was found to be governed by the amount of calcium hydroxide available in the mixes before the process started.

Axial flow hydraulic pulse testing of an argillaceous limestone

Lindsay Limestone is an argillaceous limestone encountered in Southern Ontario, Canada. The Lindsay formation is regarded as a suitable geologic medium for the construction of a Deep Geologic Repository for storing low to intermediate level nuclear waste. The Lindsay Limestone is a nodular argillaceous rock with very low permeability, requiring the use of hydraulic pulse tests for the measurement of its permeability characteristics. The paper describes the triaxial testing facility, the theoretical basis for the test and the procedures used to analyze the experimental results for estimating the permeability of the Lindsay Limestone. The results are compared with the data available in the literature.

Corrosion susceptibility of steel drums containing cemented intermediate level nuclear wastes

Cementation processes are used as immobilization techniques for low or intermediate level radioactive waste for economical and safety reasons and for being a simple operation. In particular, ion-exchange resins commonly used for purification of radioactive liquid waste from nuclear reactors are immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability. Combustible solid radioactive waste can be incinerated and the resulting ashes can also be immobilized before storage. The immobilized resins and ashes are then contained in steel drums that may undergo corrosion depending on the presence of certain contaminants. The work described in this paper was aimed at evaluating the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins and incineration ashes containing different concentrations of aggressive species (mostly chloride and sulphate ions). A special type of specimen was designed to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 1 year. The results show the deleterious effect of chloride on the expected lifespan of the waste containers.

Immobilisation matrices for intermediate level nuclear wastes using sulphate activated BFS/OPC and PFA/OPC composite cements

The highly alkaline pore solution of ordinary Portland cement composites currently used by the nuclear waste immobilisation industry causes corrosion of encapsulated reactive metals which generates hydrogen, causes loss of wasteform integrity and may release radioactive material to the environment. This work has investigated using near neutral salts as additives to composite cement powders in order to develop less alkaline cement pastes and reduce the corrosion of reactive metals. The effect of adding anhydrite and gypsum to ordinary Portland cement partially replaced by blast furnace slag or pulverised fuel ash was investigated. Adding the sulphate salts significantly changed the composition and microstructure of the bulk cement and the aluminium corrosion layer with ettringite the main phase detected. The amount of aluminium hydroxide formed during corrosion was significantly reduced by adding the sulphate salts as was the rate of hydrogen generation.

The socio-economic and communication challenges of spent nuclear fuel management in Finland: The post site selection phase of the repository project in Eurajoki

One of the key factors behind the expansion of nuclear energy policy in Finland is nuclear waste management. As in many countries, nuclear waste management agencies are still struggling with the siting of intermediate and low level nuclear waste management facilities, in Finland Posiva, the nuclear waste company owned by the two nuclear power utilities Teollisuuden Voima (TVO) and Fortum Power and Heat (FPH), has been excavating a underground characterization facility since 2004 in the Olkiluoto site in the municipality of Eurajoki. The facility under construction is meant to be for a part of the final repository for disposal of spent nuclear fuel (SNF). The site selection process of the repository was launched in Finland in the early 1980s. The site selection process included different measures to inform and involve local citizens. For example, an Environmental Impact Assessment process was implemented by the nuclear waste company in the late 1990s. The siting process, which proceeded step by step, was successful as in January 2000 the local council of the municipality of Eurajoki approved the siting of the repository by 20 votes to 7. The Council of State made a favourable decision-in-principle (DiP) in December 2000 and Parliament ratified the decision in May 2001. Posiva has also applied for a decision-in-principle to extend the facility. The aim of the paper is to introduce the first observations of the survey focused on local inhabitants. The paper addresses such questions as: What are the main impacts of the repository according to local residents? How can they obtain information about the repository? Which actors are seen as reliable sources of information regarding the final disposal of spent nuclear fuel? The survey was carried out in June 2008. The respondents ( N = 3000) were selected from the residents of the municipality of Eurajoki and the neighbouring municipalities using stratified random sampling. The information provided by the survey is important as the municipality of Eurajoki is the very first municipality in the world where the views of local residents on the construction of a repository for the final disposal of spent nuclear fuel could be elicited after the decision of the site selection. The post site selection phase the municipality of Eurajoki is currently experiencing can provide information useful for other countries likely to face this phase in the future. The paper is a part of the co-operation research project of the University of Jyväskylä and the University of Tampere. The research project “Follow-up research regarding socio-economic effects and communication of final disposal facility of spent nuclear fuel in Eurajoki and its neighbouring municipalities” is funded by the Finnish Research Programme on Nuclear Waste Management (KYT2010).

Modeling the competition between solid solution formation and cation exchange on the retardation of aqueous radium in an idealized bentonite column

Clays and clay rocks are considered viable geotechnical barriers in radioactive waste disposal. One reason for this is the propensity for cation exchange reactions in clay minerals to retard the migration of radionuclides. Although another retardation mechanism, namely the incorporation of radionuclides into sulfate or carbonate solid solutions, has been known for a long time, only recently has it been examined systematically. In this work, we investigate the competitive effect of both mechanisms on the transport of radium (Ra) in the near-field of a low- and intermediate level nuclear waste repository. In our idealized geochemical model, numerical simulations show that barium (Ba) and strontium (Sr) needed for Ra sulfate solid solutions also partition to the cation exchange sites of montmorillonite (Mont), which is the major mineral constituent of bentonite that is used for tunnel backfill. At high Mont content, most Ra tends to attach to Mont, while incorporation of Ra in sulfate solid solutions is more important at low Mont content. To explore the effect of the Mont content on the transport of radium, a multi-component reactive transport model was developed and implemented in the scientific software OpenGeoSys-GEM. It was found that a decrease of fixation capacity due to low Mont content is compensated by the formation of solid solutions and that the migration distance of aqueous Ra is similar at different Mont/water ratios.

Gross alpha determination in radioactive wastes from nuclear power plants using the track registration technique

Low and intermediate level nuclear wastes (ion-exchange resins and evaporator concentrates) essentially contain beta and gamma emitters, with very few alpha emitters. Several techniques may be used to determine gross alpha activity but, in this case, solid-state nuclear track detectors (SSNTDs) are a suitable technique for gross alpha determination because track detectors are not sensitive to beta and gamma emitters. Also, this technique is simple and inexpensive. In this paper, we studied the parameters (background, efficiency and self-absorption) that could affect the gross alpha determination using SSNTDs for both sample preparation methods, the “dry method” with tensioactives and the “wet method”. For the “dry method”, a self-absorption curve for 241Am standard was prepared using a set of varying thickness of sodium salt and for two different tensioactives: Tween ® 20 and Teg. The results showed that, below 1 mg / cm 2 , the self-absorption factor can be considered similar for both tensioactives and equal to unity. Several detectors for gross alpha determination were compared and we found that the most suitable techniques were ZnS(Ag) solid scintillator and track detectors. Both detectors were used to compare radioactive waste samples. Finally, the proposed methods (“dry method” with Teg tensioactive and “wet method”) using track detectors were tested by analysing the gross alpha activity of several radioactive wastes.

A statistical method for passive gamma ray tomography of nuclear waste vaults

This paper demonstrates a parameter identification problem encountered in the assay of intermediate-level nuclear waste vaults. A lumped parameter model of the vault contents is employed in which a likely geometry is assumed so as to reduce the number of unknown quantities in the resulting inverse problem. The inverse problem is tackled using Bayesian statistics, and the statistical distributions are sampled using a Markov chain Monte Carlo approach. The Bayesian approach affords flexibility in the specification of prior information: i.e. information that is known in advance of recording any measurements. Two parameterizations of the problem are considered: one involving five parameters and the other involving three parameters. It is shown that in order to obtain a satisfactory solution to the five-parameter problem a very specific type of prior information has to be specified, whereas for the three-parameter problem much less specific prior information is required. The inverse problem of identifying the model parameters in both the three-parameter and five-parameter formulations is solved with limited, but quantifiable precision.

An optimised multiple test framework for project selection in the public sector, with a nuclear waste disposal case-based example

Since 1967 the real (net of inflation) discount rate used to assess public sector projects in the UK has changed four times: from 8% to a peak of 10%, then to 5%, then to 6%, and then to the current position of 3.5% for years 1–30 declining in increments to 1% by year 300. This paper argues that the trend is in the right direction, but the associated decision process remains inappropriate, because the rationale for the changes has been flawed. Many other countries use discount rates of 6% or more in decision processes which are probably even more inappropriate. The fundamental problem stems from attempts by economists to embed too many conflicting considerations in a discount rate which is used in a single hurdle rate test. A multiple test approach is needed to address all the issues that the associated decision processes need to consider. This paper offers a way forward based on alternative economics perspectives, operational research perspectives, and established practice in a project risk management context. A case-based example concerned with the disposal of intermediate level nuclear waste in the UK illustrates how it works, and the implications of current UK practice, suggesting a very different view of a decision a decade ago which is a topical again. The proposed framework should be of interest to anyone interested in public sector projects, in the UK and elsewhere, and its generalisation has implications for private sector projects and private public partnerships (PPPs). Generic project risk management processes concerned with the whole of the project life cycle could embed a generalised form.

Probing the Water Phases and Microstructure in a Model Cement Blend Matrix used for the Encapsulation of Intermediate Level Nuclear Wastes

ABSTRACTThe changes in microstructure and content of water phases during hydration of a 3:1 BFS:OPC blend are investigated by Mercury Intrusion Porosimetry (MIP), freeze-drying, Thermal Gravimetric Analysis (TGA) and 1H Nuclear Magnetic Resonance (NMR) relaxometry. MIP indicates that during the blend hydration, a reduction in the population of capillary pores (larger than about 100 nm) occurs while the population of gel pores (smaller than few tens of nanometres) increases. Between 3 and 90 days, the porosity estimated by MIP decreases from about 36% down to 18% while the median pore size decreases from about 140 nm down to 6 nm.1H NMR relaxometry shows that after 1 day of hydration, nearly 70% of the evaporable water is held in capillary pores while about 30% is present in gel pores. After two weeks, most of the evaporable water (90%) is found in pores smaller than few tens of nanometres.The amount of evaporable water detected by freeze drying decreases from less than 20 wt.% after one week of hydration down to about 16.3 wt.% after 90 days while the amount of chemically bound water related to the degree of advancement of the cement hydration and detected by TGA increases from 8 wt.% to 10.3 wt.%.During hydration the BFS:OPC blend matrix evolves from an open microporous network to one of a poorly connected network of water rich nanopores with increasing amounts of chemically bound water.