Intermediate-level Waste Research Articles

Detritiation of JET Beryllium and Tungsten

During the operation of a fusion reactor, first wall components are exposed to the plasma and therefore tritium, resulting in generation of tritiated materials that would be classified as intermediate level waste (ILW) following their removal from the vessel. Investigations were undertaken into the thermal treatment of beryllium and tungsten representative of the materials used within the Joint European Torus (JET) fusion reactor to assess if tritium from these materials can be removed in the Material Detritiation Facility at the United Kingdom Atomic Energy Authority. This detritiation process may allow the reclassification of these materials as low level waste (LLW). When heated in the presence of oxygen, both tungsten and beryllium readily oxidize as temperature increases. The oxide layers that are formed on tungsten and beryllium surfaces are thought to act as a tritium barrier, reducing the amount of tritium that can be removed by thermal treatment. As such, the generation of oxide layers may need to be minimized for treatment of tungsten and beryllium, potentially via thermal treatment at lower temperatures. Additionally, the formation of beryllium oxide presents health and safety concerns due to its toxicity and physical form. Experiments were undertaken using tungsten and beryllium samples from previous JET campaigns. The samples were heated in a pyrolyzer, and the tritium released was captured in a series of bubblers. The remaining tritium in the material was characterized by acid dissolution to allow for detritiation factors (which are defined as the fraction of tritium inventory in the sample before and after the thermal treatment) to be calculated. Tritium was successfully removed from the samples by thermal treatment in air. Future trials will use samples with larger tritium inventory to confirm obtained results and demonstrate the feasibility of thermal treatment as a detritiation method for tungsten and beryllium on higher-activity samples. This should allow for samples representative of the JET ITER-like wall (current JET configuration) to be detritiated and could demonstrate the ability of the process to reduce the tritium inventory of JET materials and allow reclassification of components from ILW to LLW.

Back-Analysis of Rock Mass Strength at a Radioactive Waste Disposal Site Using Acoustic Emission Monitoring Data and 3D Numerical Modelling

In this study, a new method was examined that used acoustic emission (AE) monitoring data, in combination with a primary Boundary Element stress analysis, to back-calculate rock mass strength. The presented AE data came from the National Radioactive Waste Repository (NRWR) for low- and intermediate-level waste (LLW/ILW). AE monitoring is able to detect the pre-peak, peak, and post-peak stress changes in rock mass. The presented method used AE monitoring data to back-calculate parameters, such as uniaxial compressive strength and the rock mass deformation modulus. The AE initiation threshold was used to develop an objective function that considered the stress in the rock mass and the rock mass strength. The findings of this research propose that most AE events can be related to the crack initiation threshold, and in space, most of them are located at points away from the excavated walls. In the vicinity of the excavation damage zones around the cavities, the stress conditions beyond the crack damage boundary of the rock mass occur in many areas, leading to significant irreversible deformations. This novel method was demonstrated to aid in the prediction of rock mass strength and is a valuable, non-invasive method for improving the spatial prediction of rock mass parameters, which will lead to safer underground storage facilities.

Uranium surface processes with sandstone and volcanic rocks in acidic and alkaline solutions

Understanding the behaviour of uranium waste, for disposal purposes, is crucial due to the correlation between pH values and the disposal of distinct types of waste, with low level waste typically associated with acidic pH values, and higher and intermediate level waste commonly related to alkaline pH values. We studied the adsorption of U(VI) on sandstone and volcanic rock surfaces at pH 5.5 and 11.5 in aqueous solutions with and without bicarbonate (2 mM HCO3–) using XAS and FTIR. In the sandstone system, U(VI) adsorbs as a bidentate complex to Si at pH 5.5 without bicarbonate and as uranyl carbonate species with bicarbonate. At pH 11.5 without bicarbonate, U(VI) adsorbs as monodentate complexes to Si and precipitates as uranophane. With bicarbonate at pH 11.5, U(VI) precipitated as a Na-clarkeite mineral or remained as a uranyl carbonate surface species. In the volcanic rock system, U(VI) adsorbed to Si as an outer sphere complex at pH 5.5, regardless of the presence of bicarbonate. At pH 11.5 without bicarbonate, U(VI) adsorbed as a monodentate complex to one Si atom and precipitated as a Na-clarkeite mineral. With bicarbonate at pH 11.5, U(VI) sorbed as a bidentate carbonate complex to one Si atom. These results provide insight into the behaviour of U(VI) in heterogeneous, real-world systems related to the disposal of radioactive waste.

A study on the effect of material impurity concentration on radioactive waste levels for plans for decommissioning of nuclear power plant

Co and Eu impurities in the SSCs are nuclides that dominantly influence the neutron-induced radioactive inventory in metal and concrete radwastes (radioactive wastes) during NPP decommission. The impurity concentrations provided by NUREG/CR-3474 were used for the practical range of Co and Eu impurity concentrations to be applied to the code calculations. Metal structures near the core were evaluated to be ILW (intermediate-level waste) for the whole range of Co impurity concentration, so the boundary line between ILW and LLW (low-level waste) has no change for the whole concentration range provided by NUREG/CR-3474. Also, the boundary line between VLLW (very low-level waste) and CW (clearance waste) in the concrete shield could alter a little depending on the Eu impurity concentration within the range provided by NUREG/CR-3474. From this work, it is found that the concentration of material impurities of SSCs gives no critical impact on determining radwaste levels.

Disposal of short-lived low- and intermediate-level waste: analysis of foreign experience

The article summarizes and analyzes the main approaches and practical experience of foreign countries regarding the disposal of short-lived low- and medium-level waste. Their place in the category of low-level waste according to the recommended classification of the IAEA, for which disposal in near-surface storage facilities is provided, is shown. Currently, in countries that operate nuclear facilities and develop nuclear technologies, there is no one approach to the disposal of such waste. The type of repository and its structural and technological solutions are determined by the specifics of the waste itself, the nature of the site, the state strategy in the nuclear industry, as well as social and economic factors. To date, several types of storage facilities have been developed for the disposal of short-lived low- and medium-level waste: relatively simple trench-type near-surface storage facilities; modular-type near-surface storage facilities with a complicated system of engineering barriers and criteria for selecting containing rocks (natural barriers) compared to the trench type; geological storages, which are mainly built in mining operations with a complex system of tunnels and barrier protection. Bentonite clays are widely used as components of engineering barriers. In the simplest case of near-surface repositories, the bentonite layer serves as the base material of the repositories and helps to reduce filtration and increase the sorption of radionuclides and other components. In the organization of geological repositories, the bentonite layer contributes to the reduction of water filtration into the repositories, increases the sorption of radionuclides, prevents the destruction of concrete structures during a shift by containing rocks, and, in the event of the destruction of cement-concrete structures of engineering barriers, leads to a decrease in the removal of radionuclides from the repositories into the environment. The work gives examples of different types of storage facilities in countries around the world.

SAFETY IMPROVEMENT OF THE RADIOACTIVE WASTE STORAGE FACILITY IN ALBANIA.

The Albanian National Radioactive Waste Storage Facility (NRWSF) constructed in 1999 consists of an interim waste storage facility for very low level waste, low level waste, intermediate level waste and disused sealed radioactive sources (DSRS) coming from research institutions, agriculture and industrial activities and from applications in nuclear medicine in Albania. The safety assessment of this facility is performed considering its impact to workers, public and environment. In 2018, Albania started a new National project Alb 9010 titled 'Upgrading the Radioactive Waste Storage Building According to International Standards'. The two fundamental objectives are to ensure the protection of the public, workers and the environment and to ensure retrievability of waste packages for final disposal. The project focuses on ensuring the safety of the treatment, conditioning and storage of radioactive waste and DSRS. Under this project has been done the improvement of the safety of the RWSF and upgrading of the existing building infrastructure. In the end of the project the objective is to improve the safety of the personnel in the RWSF, protection of public and environment, Upgrade the Equipment's used in the RW Management Activities, Upgrade the existing Infrastructure of the RWSF Building and also ensure retrievability of waste packages for final disposal.

Development of a specialised calibration and infra-red emissivity correction to enable the temperature monitoring of intermediate level nuclear waste using thermal imaging

The temperature measurement of intermediate level nuclear waste containers is vital to monitor thermal stability and track temperature trends over time. The use of thermal imaging for this monitoring has numerous advantages compared to using a single spot radiation thermometer, and Sellafield wish to employ this approach for their ‘3 m3 box’ intermediate level waste containers. In order to achieve this objective with confidence a custom temperature calibration is required for low uncertainty temperature measurement. Another key consideration for practical non-contact temperature measurement is the emissivity of the surface being measured at the relevant infrared wavelengths, as such the emissivity of the Sellafield 3 m3 box materials was measured and then corrected for. The implementation of a calibrated instrument and emissivity correction were validated using temperature controlled, 1:10 scale model 3 m3 box containers within a climatic chamber. The surface temperature of the scale containers was determined to an uncertainty of ± 6.5 °C (k = 2.1) for the container walls, ± 3.5 °C (k = 2.1) for the container lid and ± 1.5 °C (k = 2.1) for the container vents. The measurement techniques used to determine the temperature of the 1:10 scale 3 m3 box containers were therefore successful and there is a clear development path to monitoring the containers in storage using thermal imaging.

IMPROVED CONCEPTUAL DESIGN OF LILW REPOSITORY

Given the forthcoming need for the construction of a repository for low and intermediate radioactive level waste in the Republic of Croatia, this paper proposes a repository design which is, from a geotechnical point of view, simple, practical and safe, and significantly improved considering current conceptual designs. Existing low and intermediate radioactive level waste repositories are mostly vault-type, near-surface constructions with some kind of covering (top) system of protective layers. However, most of these repositories do not have a bottom protective system, apart from concrete flooring (base). The reasons for such designs include the presumed longevity of the waste packages (containers), which are mostly reinforced concrete and/or steel containers. Considering that the concrete is a material which will, under certain conditions, deteriorate (e.g. dissolution of the cement matrix), and so potentially release radionuclides to the environment, it is essential to design the repository in such a manner that all forms of early release of radionuclides are prevented. The improved conceptual design of low and intermediate radioactive level waste repository presented in this paper is intended to provide an improved containment of radionuclides from waste and ensure the long term safety of the repository. This paper is the first in a series which will cover the basic design of the repository, systems of protective layers and preliminary slope stability analyses.

Waste-specific volume for retrospective, predictive assessments, and ranking of practices during normal operation of Russian nuclear power plants

Possibilities of applying the generated radioactive waste-specific volume per unit of produced electricity are shown. This waste specific volume is used for retrospective assessment and forecasting of radio active waste volumes generated at Russian nuclear power plants. According to the available data period covering 2008-2021, the mean and median values of the annual waste-specific volume for each nuclear power plant were obtained. The medians for solid radioactive wastes divided into the categories of very low-level wastes, low-level wastes, intermediate level wastes and high-level wastes are equal to 3.6?10-2, 3.2?10-2, 3.2?10-3, 3.0?10-4 m3(GWh)-1, respectively. For liquid radioactive wastes of the low-level waste and intermediate level waste categories - 1.3?10-3 m3(GWh)-1, 2.4?10-2 m3(GWh)-1, respectively. The highest mean and median values of waste-specific volume for all radioactive waste categories are typical for nuclear power plants with LWGR (RBMK) reactor installations. The forecast based on the plans to increase electricity production by Russian nuclear power plants indicates a likely increase in the volume of radioactive waste generation by 0.7-7.4 % (depending on the waste category) in the period from 2022 to 2027. The waste-specific volume use makes it possible to rank the existing practices of nuclear power plant operation by the volume of radioactive waste generation to justify the criteria for compliance with the International Project on Innovative Nuclear Reactors sustainability methodology.

Structural stability of stacked disposal containers in silo-type low- and intermediate-level waste repository under weight of backfilling materials

The long-term structural stability of radioactive-waste repositories is important for isolating radionuclides from the biosphere. The weight of backfilling materials affects the stability of these repositories. This paper reports the results of fracture experiments and simulations performed on a 1:6 scale 16-pack disposal container to investigate its structural stability. The fracture experiment was performed by piling steel balls and lead blocks atop the container. In addition, fracture experiments and finite-element simulations were performed on a container reinforced with additional poles between 200-liter drum positions. The obtained results demonstrate the structural instability of the original container under the backfilling-material weight. Moreover, the reinforced container is observed to withstand a load that exceeds the backfilling-material weight. The benchmarking between the experimental and simulation results was accomplished by comparing the yield stresses and strains observed in both cases. These results can serve as background data to relieve the compressive-strength criterion of silo-type repositories.

Nuclear fusion waste classification of K-DEMO under the Korean radioactive waste regulations

The nuclear fusion waste generated by the activation of the constitutent components of the Korean fusion DEMOnstration reactor (K-DEMO) is assessed after the end of life (EOL). Based on the results, an attempt is made to classify the radioactive waste (RW) according to the Korean regulations on RW classification and self-disposal standards. One hundred years after EOL, most (∼90%) of the reduced activation ferritic/martensitic steel structures in the tritium-breeding blanket can be classified as intermediate-level waste (ILW). One-third of the total weight of borated stainless steel inside the shielding blanket can also be classified as ILW 100 years after EOL. The remaining components of the K-DEMO are assessed as low-level waste. Based on this study, fusion waste disposal strategies will be planned in parallel with the advance of the K-DEMO design to minimize the fraction of ILW.

Evaluation of the Resistance to Crack Propagation of High-Performance Concretes by the Wedge Splitting Test Method

<p>This paper describes the key elements of the results of investigations on the resistance to crack propagation carried out as part of a large-scale project to develop high-performance concretes (HPCs) for the secondary lining of a low and intermediate-level waste disposal shaft. Four HPCs were investigated, in which the quantity of the binder component and the proportions of the two cements and the silica fume in it were varied. We have also added steel fibers to one HPC. The wedgesplitting test method was used to determine the resistance to crack propagation. The results obtained show that all the HPC investigated achieve good resistance to crack propagation. The addition of steel fibers further improves this resistance.</p>

STORAGE: A Source Term Model for Intermediate-Level Radioactive Waste

The French Alternative Energies and Atomic Energy Commission and Orano have developed a modeling tool named the Simulation TOol Of RAdiolysis Gas Emission (STORAGE) for assessing gas generation of intermediate-level waste (ILW). The first version of this model was designed to estimate gas (more specifically hydrogen) production by radiolysis of organic and water-containing materials. The code deals with different types of waste packages: metal drums, concrete drums, bitumen packages, and compacted waste containers. Diverse radioactive waste can be handled: spent fuel cladding, reprocessing sludge, contaminated technological waste (gloves, bags, bottles, etc.), ion exchange resin, etc. The validity of the model was evaluated using a series of measurements performed on U,Pu–contaminated solid waste from the Orano plutonium laboratories at the MELOX facility. A benchmark study for compacted waste containers was also implemented; the results of STORAGE were compared with reference calculations performed by Orano Projets. Future improvements of the STORAGE model are also presented.

Solubility of niobium(V) in cementitious systems relevant for nuclear waste disposal: Characterization of the solubility-controlling solid phases

Neutron activation of 93Nb in structural materials of nuclear reactors generates the radioactive, β‒‒γ emitting 94Nb isotope (t1/2= 2.04·104 a). Waste streams containing 94Nb are disposed in repositories for low- and intermediate-level waste, which are characterized by the extensive use of cementitious materials. The chemical behavior of niobium in these systems remains ill-defined, especially with respect to the solid phases controlling its solubility. Oversaturation solubility experiments in synthetic solutions representative of cementitious pore water (10 ≤ pH ≤ 12.5) show the precipitation of Nb, with [Nb] < 10–7 M at equilibrium. The increase in solubility observed at pH > 12.5 is attributed to negatively charged hydrolysis species stabilized by Ca. Solids controlling Nb solubility were characterized by a multi-method approach. XRD, Raman and quantitative chemical analysis support the predominance of cubic pyrochlore Ca2Nb2O7(cr) at pH ≈ 13.5, as well as in systems at pH ≈ 12 and high [Ca]. The incorporation of K in the pyrochlore structure was observed in pore water systems containing potassium, whereas CaNb2O6(cr) plays a predominant role at pH ≈ 10–11. This work provides an improved scientific basis to assess solid phase formation and solubility behavior of Nb(V) in cementitious systems relevant for nuclear waste disposal.

Radiation Resistance of Solid Extractant Axionit MND40T during Deep Separation of Americium from Liquid Intermediate-Level Waste of Chemical and Metallurgical Production

Radiation Resistance of Solid Extractant Axionit MND40T during Deep Separation of Americium from Liquid Intermediate-Level Waste of Chemical and Metallurgical Production

Simulation of radiation damage via alpha decay in BFS:PC grouts using 4He2+ ion acceleration

The impact of alpha radiation on cements used to encapsulate intermediate-level waste (ILW) is not well understood. ILW wastes can contain high levels of alpha-emitting radionuclides, meaning that the grouts used to encapsulate them are exposed to significant ionising radiation. Thus, a damaged region could develop in the grout adjacent to the alpha-emitting species. This work attempted to recreate this behaviour through nonradioactive 4He2+ ion-accelerator experiments, which have not previously been applied to common encapsulation grouts. The influence of this irradiation on a slag-Portland cement was investigated at different ages via transmission electron microscopy energy-dispersive x-ray spectroscopy (TEM-EDX) and supporting techniques, to assess whether 4He2+ irradiation caused textural or chemical zonation. No significant changes in hydrate phases or textures were observed, other than minor variations associated with carbonation. This paper provides a proof of concept for using ion acceleration techniques on cements and furthers knowledge on their radiation response.

Overview on the management of radioactive waste from fusion facilities: ITER, demonstration machines and power plants

In the absence of official standards and guidelines for nuclear fusion plants, fusion designers adopted, as far as possible, well-established standards for fission-based nuclear power plants (NPPs). This often implies interpretation and/or extrapolation, due to differences in structures, systems and components, materials, safety mitigation systems, risks, etc. This approach could result in the consideration of overconservative measures that might lead to an increase in cost and complexity with limited or negligible improvements. One important topic is the generation of radioactive waste in fusion power plants. Fusion waste is significantly different to fission NPP waste, i.e. the quantity of fusion waste is much larger. However, it mostly comprises low-level waste (LLW) and intermediate level waste (ILW). Notably, the waste does not contain many long-lived isotopes, mainly tritium and other activation isotopes but no-transuranic elements. An important benefit of fusion employing reduced-activation materials is the lower decay heat removal and rapid radioactivity decay overall. The dominant fusion wastes are primarily composed of structural materials, such as different types of steel, including reduced activation ferritic martensitic steels, such as EUROFER97 and F82H, AISI 316L, bainitic, and JK2LB. The relevant long-lived radioisotopes come from alloying elements, such as niobium, molybdenum, nickel, carbon, nitrogen, copper and aluminum and also from uncontrolled impurities (of the same elements, but also, e.g. of potassium and cobalt). After irradiation, these isotopes might preclude disposal in LLW repositories. Fusion power should be able to avoid creating high-level waste, while the volume of fusion ILW and LLW will be significant, both in terms of pure volume and volume per unit of electricity produced. Thus, efforts to recycle and clear are essential to support fusion deployment, reclaim resources (through less ore mining) and minimize the radwaste burden for future generations.

Development of a Sorption Technique for the Selective Separation of Plutonium and Americium from Nitric Acid Intermediate-Level Wastes of Chemical and Metallurgical Production

Development of a Sorption Technique for the Selective Separation of Plutonium and Americium from Nitric Acid Intermediate-Level Wastes of Chemical and Metallurgical Production

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.

Safety verification tests of low and intermediate level radioactive solid waste drum

Steel drums or boxes are used to transport, store, and dispose of low- and intermediate-level waste. According to relevant standards, before the solid waste container design is finalized, a series of type tests should be performed, including leakage test, water spray test, stacking test, penetration test, free-drop test, etc. This article focuses on the acceptance test and test methods for the LID- II b type waste drum finalization test. The tests in accordance with the standard requirements are the first time reported in China. It found the inherent design defects, and had high guiding significance to improve design and test verification operation.