Use of poly(methyl methacrylate) in radioactive waste management: I. Radiation stability and degradation

Advances in Materials with Self-Healing Properties: A Brief Review

The development of materials with self-healing capabilities has garnered considerable attention due to their potential to enhance the durability and longevity of various engineering and structural applications. In this review, we provide an overview of recent advances in materials with self-healing properties, encompassing polymers, ceramics, metals, and composites. We outline future research directions and potential applications of self-healing materials (SHMs) in diverse fields. This review aims to provide insights into the current state-of-the-art in SHM research and guide future efforts towards the development of innovative and sustainable materials with enhanced self-repair capabilities. Each material type showcases unique self-repair mechanisms tailored to address specific challenges. Furthermore, this review investigates crack healing processes, shedding light on the latest developments in this critical aspect of self-healing materials. Through an extensive exploration of these topics, this review aims to provide a comprehensive understanding of the current landscape and future directions in self-healing materials research.

Comparative analysis on application conditions of indium (III) oxide-reinforced glasses in nuclear waste management and source transportation: A Monte Carlo simulation study

This study's primary objective is to provide the preliminary findings of novel research on the design of Indium (III) oxide-reinforced glass container that were thoroughly developed for the purpose of a nuclear material container for transportation and waste management applications. The shielding characteristics of an Indium (III) oxide-reinforced glass container with a certain elemental composition against the ⁶⁰Co radioisotope was thoroughly evaluated. The energy deposition in the air surrounding the designed portable glass containers is measured using MCNPX general-purpose Monte Carlo code. Simulation studies were carried out using Lenovo-P620 workstation and the number of tracks was defined as 10⁸ in each simulation phase. According to results, the indium oxide-doped C6 (TZI8) container exhibits superior protective properties compared to other conventional container materials such as 0.5Bitumen-0.5 Cement, Pb Glass composite, Steel-Magnetite concrete. In addition to its superiority in terms of nuclear safety, it is proposed that the source's simultaneous observation and monitoring, as well as the C6 (TZI8) glass structure's transparency, be underlined as significant advantages. High-density glasses, which may replace undesirable materials such as concrete and lead, provide several advantages in terms of production ease, non-toxic properties, and resource monitoring. In conclusion, the use of Indium (III) oxide-reinforced glass with its high transparency and outstanding protection properties may be a substantial choice in places where concrete is required to ensure the safety of nuclear materials.

Removal of radioactive contaminants by polymeric microspheres

Radionuclide removal from radioactive liquid waste by adsorption on polymeric microspheres is the latest application of polymers in waste management. Polymeric microspheres have significant immobilization capacity for ionic substances. A laboratory study was carried out by using poly(N-isopropylacrylamide) for encapsulation of radionuclide in the liquid radioactive waste. There are numbers of advantages to use an encapsulation technology in radioactive waste management. Results show that polymerization step of radionuclide increases integrity of solidified waste form. Test results showed that adding the appropriate polymer into the liquid waste at an appropriate pH and temperature level, radionuclide was encapsulated into polymer. This technology may provide barriers between hazardous radioactive ions and the environment. By this method, solidification techniques became easier and safer in nuclear waste management. By using polymer microspheres as dust form, contamination risks were decreased in the nuclear industry and radioactive waste operations.

Silica with immobilized phosphinic acid-derivative for uranium extraction

A novel adsorbent benzoimidazol-2-yl-phenylphosphinic acid/aminosilica adsorbent (BImPhP(O)(OH)/SiO2NH2) was prepared by carbonyldiimidazole-mediated coupling of aminosilica with 1-carboxymethylbenzoimidazol-2-yl-phenylphosphinic acid. It was obtained through direct phosphorylation of 1-cyanomethylbenzoimidazole by phenylphosphonic dichloride followed by basic hydrolysis of the nitrile. The obtained sorbent was well characterized by physicochemical methods, such as differential scanning calorimetry-mass spectrometry (DSC-MS), surface area and pore distribution analysis (ASAP), scanning electron microscopy (SEM), X-ray photoelectron (XPS) and Fourier transform infrared (FTIR) spectroscopies. The adsorption behavior of the sorbent and initial silica gel as well as aminosilica gel with respect to uranium(VI) from the aqueous media has been studied under varying operating conditions of pH, concentration of uranium(VI), contact time, and desorption in different media. The synthesized material was found to show an increase in adsorption activity with respect to uranyl ions in comparison with the initial compounds. In particular, the highest adsorption capacity for the obtained modified silica was found at the neutral pH, where one gram of the adsorbent can extract 176mg of uranium. Under the same conditions the aminosilica extracts 166mg/g, and the silica - 144mg/g of uranium. In the acidic medium, which is common for uranium nuclear wastes, the synthesized adsorbent extracts 27mg/g, the aminosilica - 16mg/g, and the silica - 14mg/g of uranium. It was found that 15% of uranium ions leached from the prepared material in acidic solutions, while 4% of uranium can be removed in a phosphate solution.

The synthesis and characterization of tributyl phosphate grafted carbon nanotubes by the floating catalytic chemical vapor deposition method and their sorption behavior towards uranium

An efficient method for grafting tributylphosphate onto carbon nanotubes (CNT-TBP) is described. CNT-TBP exhibited excellent adsorption towards U(vi) ions.