Special Issue "Materials for Nuclear Waste Immobilization"

A special issue of Materials (ISSN 1996-1944).

Deadline for manuscript submissions: closed (1 June 2019).

Special Issue Editors

Dr. Michael I Ojovan
E-Mail Website
Guest Editor
Immobilisation Science Laboratory, Department of Materials Science and Engineering, University of Sheffield, Mappin Street, Sheffield S1 3JD, United Kingdom
Interests: structure and properties of amorphous materials including the viscosity and glass transition; materials for nuclear waste immobilization; waste processing technologies
Special Issues and Collections in MDPI journals
Prof. Neil C. Hyatt
E-Mail Website
Guest Editor
Immobilisation Science Laboratory, Department of Materials Science and Engineering, University of Sheffield, Mappin Street, Sheffield S1 3JD, United Kingdom
Interests: radioactive waste management and disposal; advanced nuclear materials; structure–property relations in mixed metal oxides

Special Issue Information

Dear Colleagues,

Nuclear energy is clean, reliable and competitive with many useful applications among which power generation is the most important, where it can gradually replace fossil fuels and avoid massive pollution to the environment. A useless by-product, resulting from utilization of nuclear energy in both power generation and other applications, such as in medicine, industry, agriculture, and research, is nuclear waste.

Safe and effective management of nuclear waste is crucial in ensuring sustainable utilization of nuclear energy. Nuclear waste must be processed to make it safe, which includes its conditioning, so it is immobilized and packaged before storage and disposal. Immobilization of waste radionuclides in durable wasteform materials provides the most important barrier to contribute to the overall performance of any storage and/or disposal system. Materials for nuclear waste immobilization are, thus, at the core of multibarrier systems of isolation of radioactive waste from environment aimed to ensure long term safety of storage and disposal.

This Special Issue aims to analyze the materials currently used, as well as novel materials for nuclear waste immobilization, including technological approaches utilized in nuclear waste conditioning pursuing to ensure efficiency and long-term safety of storage and disposal systems. It will focus on cementitious materials, geopolymers, glasses, glass composite materials, and ceramics developed and used in nuclear waste immobilization with performance of such materials of the utmost importance.

Dr. Michael I Ojovan
Prof. Neil C. Hyatt
Guest Editors

Manuscript Submission Information

Manuscripts should be submitted online at www.mdpi.com by registering and logging in to this website. Once you are registered, click here to go to the submission form. Manuscripts can be submitted until the deadline. All papers will be peer-reviewed. Accepted papers will be published continuously in the journal (as soon as accepted) and will be listed together on the special issue website. Research articles, review articles as well as short communications are invited. For planned papers, a title and short abstract (about 100 words) can be sent to the Editorial Office for announcement on this website.

Submitted manuscripts should not have been published previously, nor be under consideration for publication elsewhere (except conference proceedings papers). All manuscripts are thoroughly refereed through a single-blind peer-review process. A guide for authors and other relevant information for submission of manuscripts is available on the Instructions for Authors page. Materials is an international peer-reviewed open access semimonthly journal published by MDPI.

Please visit the Instructions for Authors page before submitting a manuscript. The Article Processing Charge (APC) for publication in this open access journal is 1800 CHF (Swiss Francs). Submitted papers should be well formatted and use good English. Authors may use MDPI's English editing service prior to publication or during author revisions.

Keywords

  • Radioactive waste
  • Nuclear waste management
  • Nuclear waste immobilization
  • Nuclear wasteforms
  • Plasma treatment for nuclear waste immobilization
  • Vitrification technologies for nuclear waste immobilization
  • Materials for nuclear waste immobilization
  • Cementitious materials for nuclear waste immobilization
  • Geopolymers for nuclear waste immobilization
  • Polymeric materials for nuclear waste immobilization
  • Bitumens and bituminization for nuclear waste immobilization
  • Glasses for nuclear waste immobilization
  • Glass composite materials for nuclear waste immobilization
  • Ceramics for nuclear waste immobilization
  • Metals for nuclear waste conditioning.

Published Papers (10 papers)

Order results
Result details
Select all
Export citation of selected articles as:

Research

Jump to: Review

Open AccessArticle
The Effect of Heavy Ion Irradiation on the Forward Dissolution Rate of Borosilicate Glasses Studied In Situ and Real Time by Fluid-Cell Raman Spectroscopy
Materials 2019, 12(9), 1480; https://doi.org/10.3390/ma12091480 - 07 May 2019
Abstract
Borosilicate glasses are the favored material for immobilization of high-level nuclear waste (HLW) from the reprocessing of spent fuel used in nuclear power plants. To assess the long-term stability of nuclear waste glasses, it is crucial to understand how self-irradiation affects the structural [...] Read more.
Borosilicate glasses are the favored material for immobilization of high-level nuclear waste (HLW) from the reprocessing of spent fuel used in nuclear power plants. To assess the long-term stability of nuclear waste glasses, it is crucial to understand how self-irradiation affects the structural state of the glass and influences its dissolution behavior. In this study, we focus on the effect of heavy ion irradiation on the forward dissolution rate of a non-radioactive ternary borosilicate glass. To create extended radiation defects, the glass was subjected to heavy ion irradiation using 197Au ions that penetrated ~50 µm deep into the glass. The structural damage was characterized by Raman spectroscopy, revealing a significant depolymerization of the silicate and borate network in the irradiated glass and a reduction of the average boron coordination number. Real time, in situ fluid-cell Raman spectroscopic corrosion experiments were performed with the irradiated glass in a silica-undersaturated, 0.5 M NaHCO3 solution at temperatures between 80 and 85 °C (initial pH = 7.1). The time- and space-resolved in situ Raman data revealed a 3.7 ± 0.5 times increased forward dissolution rate for the irradiated glass compared to the non-irradiated glass, demonstrating a significant impact of irradiation-induced structural damage on the dissolution kinetics. Full article
(This article belongs to the Special Issue Materials for Nuclear Waste Immobilization)
Show Figures

Figure 1

Open AccessArticle
An Assessment of Initial Leaching Characteristics of Alkali-Borosilicate Glasses for Nuclear Waste Immobilization
Materials 2019, 12(9), 1462; https://doi.org/10.3390/ma12091462 - 06 May 2019
Cited by 1
Abstract
Initial leaching characteristics of simulated nuclear waste immobilized in three alkali- borosilicate glasses (ABS-waste) were studied. The effects of matrix composition on the containment performance and degradation resistance measures were evaluated. Normalized release rates are in conformance with data reported in the literature. [...] Read more.
Initial leaching characteristics of simulated nuclear waste immobilized in three alkali- borosilicate glasses (ABS-waste) were studied. The effects of matrix composition on the containment performance and degradation resistance measures were evaluated. Normalized release rates are in conformance with data reported in the literature. High Li and Mg loadings lead to the highest initial de-polymerization of sample ABS-waste (17) and contributed to its thermodynamic instability. Ca stabilizes non-bridging oxygen (NBO) and reduces the thermodynamic instability of the modified matrix. An exponential temporal change in the alteration thickness was noted for samples ABS-waste (17) and Modified Alkali-Borosilicate (MABS)-waste (20), whereas a linear temporal change was noted for sample ABS-waste (25). Leaching processes that contribute to the fractional release of all studied elements within the initial stage of glass corrosion were quantified and the main controlling leach process for each element was identified. As the waste loading increases, the contribution of the dissolution process to the overall fractional release of structural elements decreases by 43.44, 5.05, 38.07, and 52.99% for Si, B, Na, and Li respectively, and the presence of modifiers reduces this contribution for all the studied metalloids. The dissolution process plays an important role in controlling the release of Li and Cs, and this role is reduced by increasing the waste loading. Full article
(This article belongs to the Special Issue Materials for Nuclear Waste Immobilization)
Show Figures

Figure 1

Open AccessArticle
Investigating the Durability of Iodine Waste Forms in Dilute Conditions
Materials 2019, 12(5), 686; https://doi.org/10.3390/ma12050686 - 26 Feb 2019
Cited by 1
Abstract
To prevent the release of radioiodine during the reprocessing of used nuclear fuel or in the management of other wastes, many technologies have been developed for iodine capture. The capture is only part of the challenge as a durable waste form is required [...] Read more.
To prevent the release of radioiodine during the reprocessing of used nuclear fuel or in the management of other wastes, many technologies have been developed for iodine capture. The capture is only part of the challenge as a durable waste form is required to ensure safe disposal of the radioiodine. This work presents the first durability studies in dilute conditions of two AgI-containing waste forms: hot-isostatically pressed silver mordenite (AgZ) and spark plasma sintered silver-functionalized silica aerogel (SFA) iodine waste forms (IWF). Using the single-pass flow-through (SPFT) test method, the dissolution rates respective to Si, Al, Ag and I were measured for variants of the IWFs. By combining solution and solid analysis information on the corrosion mechanism neutral-to-alkaline conditions was elucidated. The AgZ samples were observed to have corrosion preferentially occur at secondary phases with higher Al and alkali content. These phases contained a lower proportion of I compared with the matrix. The SFA samples experienced a higher extent of corrosion at Si-rich particles, but an increased addition of Si to the waste led to an improvement in corrosion resistance. The dissolution rates for the IWF types are of similar magnitude to other Si-based waste form materials measured using SPFT. Full article
(This article belongs to the Special Issue Materials for Nuclear Waste Immobilization)
Show Figures

Figure 1

Open AccessArticle
Preliminary Assessment of Criticality Safety Constraints for Swiss Spent Nuclear Fuel Loading in Disposal Canisters
Materials 2019, 12(3), 494; https://doi.org/10.3390/ma12030494 - 05 Feb 2019
Abstract
This paper presents preliminary criticality safety assessments performed by the Paul Scherrer Institute (PSI) in cooperation with the Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) for spent nuclear fuel disposal canisters loaded with Swiss Pressurized Water Reactor (PWR) UO2 [...] Read more.
This paper presents preliminary criticality safety assessments performed by the Paul Scherrer Institute (PSI) in cooperation with the Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) for spent nuclear fuel disposal canisters loaded with Swiss Pressurized Water Reactor (PWR) UO2 spent fuel assemblies. The burnup credit application is examined with respect to both existing concepts: taking into account actinides only and taking into account actinides plus fission products. The criticality safety calculations are integrated with uncertainty quantifications that are as detailed as possible, accounting for the uncertainties in the nuclear data used, fuel assembly and disposal canister design parameters and operating conditions, as well as the radiation-induced changes in the fuel assembly geometry. Furthermore, the most penalising axial and radial burnup profiles and the most reactive fuel loading configuration for the canisters were taken into account accordingly. The results of the study are presented with the help of loading curves showing what minimum average fuel assembly burnup is required for the given initial fuel enrichment of fresh fuel assemblies to ensure that the effective neutron multiplication factor, k e f f , of the canister would comply with the imposed criticality safety criterion. Full article
(This article belongs to the Special Issue Materials for Nuclear Waste Immobilization)
Show Figures

Figure 1

Open AccessArticle
Treatment Technology of Hazardous Water Contaminated with Radioisotopes with Paper Sludge Ash-Based Geopolymer—Stabilization of Immobilization of Strontium and Cesium by Mixing Seawater
Materials 2018, 11(9), 1521; https://doi.org/10.3390/ma11091521 - 24 Aug 2018
Cited by 2
Abstract
Long-term immobilization ratios of strontium (Sr2+) and cesium (Cs+) in paper sludge ash-based geopolymer (PS-GP) were investigated in one year. PS-GP paste specimens were prepared in the conditions of 20 °C and 100% R.H., using two kinds of paper [...] Read more.
Long-term immobilization ratios of strontium (Sr2+) and cesium (Cs+) in paper sludge ash-based geopolymer (PS-GP) were investigated in one year. PS-GP paste specimens were prepared in the conditions of 20 °C and 100% R.H., using two kinds of paper sludge ash (PS-ash). Two kinds of alkaline solution were used in the PS-GP as activator. One was prepared by diluting aqueous Na-disilicate (water glass) with seawater. Another was a mixture of this solution and caustic soda of 10 M concentration. When seawater was mixed into the alkaline solution, unstable fixations of Sr2+ and Cs+ were greatly improved, resulting stable and high immobilization ratios at any age up to one year, no matter what kind of PS-ash and alkaline solution were used. Element maps obtained by EPMA exhibited nearly even distribution of Cs+. However Sr2+ was biased, making domains so firmly related to Ca2+ presence. The mechanism that seawater stabilizes immobilization of Sr2+ and Cs+ was discussed in this study, but still needs to further investigation. Chemical composition analyses of PS-GP were also conducted by SEM-EDS. Two categories of GP matrix were clearly observed, so called N-A-S-H and C-A-S-H gels, respectively. By plotting in ternary diagrams of SiO2-(CaO + Na2O)-Al2O3 and Al2O3-CaO-Na2O, compositional trends were discussed in view of ‘plagioclase gels’ newly found in this study. As a result, it is suggested that the N-A-S-H and C-A-S-H gels should be strictly called Na-rich N-C-A-S-H and Ca-rich N-C-A-S-H gels, respectively. Full article
(This article belongs to the Special Issue Materials for Nuclear Waste Immobilization)
Show Figures

Figure 1

Open AccessArticle
Compact Storage of Radioactive Cesium in Compressed Pellets of Zeolite Polymer Composite Fibers
Materials 2018, 11(8), 1347; https://doi.org/10.3390/ma11081347 - 03 Aug 2018
Abstract
To facilitate the safe storage of radioactive Cs, a zeolite–poly(ethersulfone) composite fiber was fabricated to be a compact storage form of radioactive Cs, and an immobilization was investigated with respect to the effects of volume reduction and stability of the fiber’s adsorbent matrix. [...] Read more.
To facilitate the safe storage of radioactive Cs, a zeolite–poly(ethersulfone) composite fiber was fabricated to be a compact storage form of radioactive Cs, and an immobilization was investigated with respect to the effects of volume reduction and stability of the fiber’s adsorbent matrix. Using compressed heat treatment at 100–800 °C for a zeolite polymer composite fiber (ZPCF) containing Cs, the fabrication changed its form from a fiber into a pellet, which decreased the matrix volume to be about one-sixth of its original volume. The Cs leakage behavior of the ZPCF matrix was examined in its compact pellet form for non-radioactive Cs and radioactive Cs when different fabrication conditions were carried out in the immobilization. The elution ratio of non-radioactive Cs from the matrix was minimal, at 0.05%, when the ZPCF was compressed with heat treatment at 300 °C. When using radioactive Cs for the compression at below 300 °C, the pellet form also had no elution of the pollutants from the matrix. When the compressed treatment was at 500 °C, the matrix exhibited elution of radioactive Cs to the outside, meaning that the plastic component was burning and decomposed in the pellet. A comparison of ZPCF and natural zeolite indicated that the compressed heating process for ZPCF was useful in a less-volume-immobilized form of the compact adsorbent for radioactive Cs storage. Full article
(This article belongs to the Special Issue Materials for Nuclear Waste Immobilization)
Show Figures

Figure 1

Open AccessArticle
Synthesis and Physical Property Characterisation of Spheroidal and Cuboidal Nuclear Waste Simulant Dispersions
Materials 2018, 11(7), 1235; https://doi.org/10.3390/ma11071235 - 18 Jul 2018
Cited by 1
Abstract
This study investigated dispersions analogous to highly active nuclear waste, formed from the reprocessing of Spent Nuclear Fuel (SNF). Non-radioactive simulants of spheroidal caesium phosphomolybdate (CPM) and cuboidal zirconium molybdate (ZM-a) were successfully synthesised; confirmed via Scanning Electron Microscopy (SEM), powder X-ray diffraction [...] Read more.
This study investigated dispersions analogous to highly active nuclear waste, formed from the reprocessing of Spent Nuclear Fuel (SNF). Non-radioactive simulants of spheroidal caesium phosphomolybdate (CPM) and cuboidal zirconium molybdate (ZM-a) were successfully synthesised; confirmed via Scanning Electron Microscopy (SEM), powder X-ray diffraction (PXRD) and Fourier transform infrared (FTIR) spectroscopy. In addition, a supplied ZM (ZM-b) with a rod-like/wheatsheaf morphology was also analysed along with titanium dioxide (TiO2). The simulants underwent thermal gravimetric analysis (TGA) and size analysis, where CPM was found to have a D50 value of 300 nm and a chemical formula of Cs3PMo12O40·13H2O, ZM-a a D50 value of 10 μm and a chemical formula of ZrMo2O7(OH)2·3H2O and ZM-b to have a D50 value of 14 μm and a chemical formula of ZrMo2O7(OH)2·4H2O. The synthesis of CPM was tracked via Ultraviolet-visible (UV-Vis) spectroscopy at both 25 °C and 50 °C, where the reaction was found to be first order with the rate constant highly temperature dependent. The morphology change from spheroidal CPM to cuboidal ZM-a was tracked via SEM, reporting to take 10 days. For the onward processing and immobilisation of these waste dispersions, centrifugal analysis was utilised to understand their settling behaviours, in both aqueous and 2 M nitric acid environments (mimicking current storage conditions). Spheroidal CPM was present in both conditions as agglomerated clusters, with relatively high settling rates. Conversely, the ZM were found to be stable in water, where their settling rate exponents were related to the morphology. In acid, the high effective electrolyte resulted in agglomeration and faster sedimentation. Full article
(This article belongs to the Special Issue Materials for Nuclear Waste Immobilization)
Show Figures

Figure 1

Open AccessArticle
Analysis of the Secondary Phases Formed by Corrosion of U3Si2-Al Research Reactor Fuel Elements in the Presence of Chloride Rich Brines
Materials 2018, 11(7), 1121; https://doi.org/10.3390/ma11071121 - 30 Jun 2018
Abstract
Corrosion experiments with non-irradiated U3Si2-Al research reactor fuel samples were carried out in synthetic MgCl2-rich brine to identify and quantify the secondary phases because depending on their composition and on their amount, such compounds can act as [...] Read more.
Corrosion experiments with non-irradiated U3Si2-Al research reactor fuel samples were carried out in synthetic MgCl2-rich brine to identify and quantify the secondary phases because depending on their composition and on their amount, such compounds can act as a sink for the radionuclide release in final repositories. Within the experimental period of 100 days at 90 °C and anoxic conditions the U3Si2-Al fuel sample was completely disintegrated. The obtained solids were subdivided into different grain size fractions and non-ambient X-ray diffraction (XRD) was applied for their qualitative and quantitative phase analysis. The secondary phases consist of lesukite (aluminum chloro hydrate) and layered double hydroxides (LDH) with varying chemical compositions. Furthermore, iron, residues of non-corroded nuclear fuel (U3Si2), iron oxy hydroxides and chlorides were also observed. In addition to high amorphous contents (>45 wt %) hosting the uranium, the quantitative phase analysis showed, that LDH compounds and lesukite were the major crystalline phases. Scanning electron microscopy (SEM) and energy dispersive -Xray spectroscopy (EDS) confirmed the results of the XRD analysis. Elemental analysis revealed that U and Al were concentrated in the solids. However, most of the iron, added as Fe(II) aqueous species, remained in solution. Full article
(This article belongs to the Special Issue Materials for Nuclear Waste Immobilization)
Show Figures

Figure 1

Open AccessArticle
Magnesium Potassium Phosphate Compound for Immobilization of Radioactive Waste Containing Actinide and Rare Earth Elements
Materials 2018, 11(6), 976; https://doi.org/10.3390/ma11060976 - 08 Jun 2018
Cited by 4
Abstract
The problem of effective immobilization of liquid radioactive waste (LRW) is key to the successful development of nuclear energy. The possibility of using the magnesium potassium phosphate (MKP) compound for LRW immobilization on the example of nitric acid solutions containing actinides and rare [...] Read more.
The problem of effective immobilization of liquid radioactive waste (LRW) is key to the successful development of nuclear energy. The possibility of using the magnesium potassium phosphate (MKP) compound for LRW immobilization on the example of nitric acid solutions containing actinides and rare earth elements (REE), including high level waste (HLW) surrogate solution, is considered in the research work. Under the study of phase composition and structure of the MKP compounds that is obtained by the XRD and SEM methods, it was established that the compounds are composed of crystalline phases—analogues of natural phosphate minerals (struvite, metaankoleite). The hydrolytic stability of the compounds was determined according to the semi-dynamic test GOST R 52126-2003. Low leaching rates of radionuclides from the compound are established, including a differential leaching rate of 239Pu and 241Am—3.5 × 10−7 and 5.3 × 10−7 g/(cm2∙day). As a result of the research work, it was concluded that the MKP compound is promising for LRW immobilization and can become an alternative material combining the advantages of easy implementation of the technology, like cementation and the high physical and chemical stability corresponding to a glass-like compound. Full article
(This article belongs to the Special Issue Materials for Nuclear Waste Immobilization)
Show Figures

Figure 1

Review

Jump to: Research

Open AccessReview
Ceramic Mineral Waste-Forms for Nuclear Waste Immobilization
Materials 2019, 12(16), 2638; https://doi.org/10.3390/ma12162638 - 19 Aug 2019
Abstract
Crystalline ceramics are intensively investigated as effective materials in various nuclear energy applications, such as inert matrix and accident tolerant fuels and nuclear waste immobilization. This paper presents an analysis of the current status of work in this field of material sciences. We [...] Read more.
Crystalline ceramics are intensively investigated as effective materials in various nuclear energy applications, such as inert matrix and accident tolerant fuels and nuclear waste immobilization. This paper presents an analysis of the current status of work in this field of material sciences. We have considered inorganic materials characterized by different structures, including simple oxides with fluorite structure, complex oxides (pyrochlore, murataite, zirconolite, perovskite, hollandite, garnet, crichtonite, freudenbergite, and P-pollucite), simple silicates (zircon/thorite/coffinite, titanite (sphen), britholite), framework silicates (zeolite, pollucite, nepheline /leucite, sodalite, cancrinite, micas structures), phosphates (monazite, xenotime, apatite, kosnarite (NZP), langbeinite, thorium phosphate diphosphate, struvite, meta-ankoleite), and aluminates with a magnetoplumbite structure. These materials can contain in their composition various cations in different combinations and ratios: Li–Cs, Tl, Ag, Be–Ba, Pb, Mn, Co, Ni, Cu, Cd, B, Al, Fe, Ga, Sc, Cr, V, Sb, Nb, Ta, La, Ce, rare-earth elements (REEs), Si, Ti, Zr, Hf, Sn, Bi, Nb, Th, U, Np, Pu, Am and Cm. They can be prepared in the form of powders, including nano-powders, as well as in form of monolith (bulk) ceramics. To produce ceramics, cold pressing and sintering (frittage), hot pressing, hot isostatic pressing and spark plasma sintering (SPS) can be used. The SPS method is now considered as one of most promising in applications with actual radioactive substances, enabling a densification of up to 98–99.9% to be achieved in a few minutes. Characteristics of the structures obtained (e.g., syngony, unit cell parameters, drawings) are described based upon an analysis of 462 publications. Full article
(This article belongs to the Special Issue Materials for Nuclear Waste Immobilization)
Show Figures

Figure 1

Back to TopTop