Radioactive waste with widely varying characteristics is generated from the operation and maintenance of nuclear reactors, nuclear fuel cycle facilities, research facilities and medical facilities and the through the use of radioisotopes in industrial applications. The waste needs to be treated and conditioned appropriately to provide wasteforms acceptable for safe storage and disposal. Conditioning of radioactive waste is an important step to prepare waste for long-term storage or disposal and includes the following processes:

• ▪ Immobilization which may or may not also provide volume reduction, including

  • a) Low temperature processes and

  • b) Thermal processes;

• ▪ Containerization for

  • a) Transport,

  • b) Storage, and

  • c) Disposal;

• ▪ Overpacking of primary containers

  • a) Prior to disposal and

  • b) In a disposal facility as part of disposal process.

Conditioning consists of operations that produce a waste package suitable for handling, transportation, storage and/or disposal and may be performed for a variety of reasons including standardization of practices and/or wasteforms, technical requirements for waste stability in relation to a repository design or safety case, technical requirements related to waste transportation, societal preferences, regulatory preferences, etc. This paper gives an overview of recent advances in conditioning of low- and intermediate-level radioactive waste.

The paper is based on the new IAEA Handbook “Conditioning of Low- and Intermediate-Level Liquid, Solidified and Solid Waste” which is one of eight IAEA handbooks intended to provide guidance for evaluating and implementing various characterisation and radioactive waste processing and storage technologies before final disposal

Current practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-century when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.

The application of safeguards measures by the International Atomic Energy Agency (IAEA) involves analytical measurements of samples taken during inspections of nuclear facilities. Thus constant development and advancement of analytical techniques is required. For quality control purposes, the IAEA has implemented a dedicated project to enhance its analytical capabilities by producing tailor-made reference materials for the analysis of uranium isotope signatures in (single) particles.

To this end, a particle production set-up was developed and implemented at Forschungszentrum Juelich capable to produce uranium oxide microparticles which are intended to be used as (certified) reference materials for particle analysis methods. A step towards the certification process is the evaluation of consistency of the size distribution and homogeneity. A monodisperse particle size distribution as well as the single phase triuranium octoxide structure was confirmed using SEM, µ-XRD and µ-Raman spectroscopy, respectively. Analysis performed on single uranium oxide microparticles confirmed consistency of the uranium isotopic ratios in comparison to the initial precursor solutions. To improve the homogeneity and particle handling, the particles are transferred into suspensions, for which the stability was investigated with respect to dissolution.

Leaching results on fragments of high burnup (65 MWd/kgU) UO2 fuel from a commercial pressurized water reactor are presented. The experiment was performed in simplified granitic groundwater under a hydrogen pressure of up to 5 MPa, representing conditions in a water intrusion scenario for a Swedish KBS-3 design spent nuclear fuel repository. The freshly crushed fragments were pre-washed for 6 days to remove pre-oxidized matrix and part of the instant release fraction of the radionuclide inventory, and then transferred to an autoclave for leaching under hydrogen conditions. Following an initial release of U attributed to dissolution of a pre-oxidized fuel layer caused by the aerated handling mainly during the transfer from pre-washing to autoclave, the U concentration decreased with time to levels of 2-5x10-9 M, which corresponds, approximately, to the solubility of amorphous UO2. The release of radionuclides such as Cs and Sr gradually declined indicating a transition to inhibition of the fuel matrix dissolution.

The Supercontainer (SC) is the reference concept for the post-conditioning of vitrified high-level nuclear waste and spent fuel in Belgium. It comprises a prefabricated concrete buffer that completely surrounds a carbon steel overpack. Welding is being considered as a final closure technique of the carbon steel overpack in order to ensure its water tightness. Welding is known to induce residual stresses near the weld zone, which may lead to an increased susceptibility to stress corrosion cracking (SCC). In this study, slow strain rate tests were conducted to study the SCC behavior of plain and welded P355 QL2 grade carbon steel exposed to an artificial concrete pore water solution that is representative of the SC concrete buffer environment. The tests were performed at 140°C, a constant strain rate of 5 × 10-7 s-1 and at open circuit potential under anoxic conditions. The effect of thiosulfate on the SCC behavior was investigated up to levels of 600 mg/L S2O32-.

Cladding tubes of water-cooled nuclear reactors are usually made of Zircaloy and are an important retaining element for radionuclides present in the fuel both during predisposal activities such as reloading of fuel assemblies from interim storage casks to final disposal casks and during final disposal in the case of canister breaching. However, cladding integrity is affected by various processes during reactor operation and beyond, e.g. fuel cladding chemical interaction and fission product precipitation onto the inner cladding surface. Using experimental and modelling methods, the radionuclide inventory of an irradiated Zircaloy-4 plenum section is analyzed. Quantities of 235/238U, 237Np, 238/239/240/241/242Pu, 241/243Am, 243/244Cm besides 14C, 55Fe, 125Sb, 154Eu, and 134/137Cs were (radio-)chemically determined in digested Zircaloy-4 subsamples. Measured inventories of activation products in the Zr-alloy are in good agreement with calculated values. However, amounts of actinides and fission products exceed the calculated inventory by factor ∼57 (minor actinides and non-volatile fission products) and ∼114 (137Cs). Excess of minor actinides and part of enhanced Cs inventory originate from fuel residues deposited on the inner cladding surface during fuel rod fabrication, whereas vast amount of cesium is volatilized from subjacent fuel pellets and transported to the plenum.

The formation mechanism and chemical form of insoluble C-14 found in PWR need to be examined in order to predict its environmental behavior after disposal. This study investigates the alteration of ion-exchange resin by heating and irradiation, because past studies indicated the ion-exchange resin may be the origin of insoluble C-14.

Resin was heated at 300 °C in solution with low oxygen content to simulate the environment of PWR coolant. The sulfo group was found to detach within 8 h, and structures similar to polystyrene were remained. This is followed by detachment of H from the alkyl group, condensation reaction, and the formation of amorphous carbon-like structure. After heating for 24 and 96 h, the resin was irradiated by 60Co γ-rays in the solution. The FT-IR and TG measurements after irradiation suggested that OH and COOH groups were formed on the surface of the resin. These functional groups may be involved in reactions that finally form the amorphous carbon.

In addition, the characteristics of heated and irradiated resin were compared to real insoluble-C (CRUD) sample in PWR (in Appendix).

In the safety assessment of a radioactive waste disposal, it is extremely important to assess the migration behavior of long-half-life radionuclides in the disposal environment. The migration behavior in the disposal environment for a radionuclide varies with the chemical form of the nuclide. In particular, C-14 has various chemical forms, and its migration behavior in the disposal environment substantially varies with the chemical forms. It has been reported that hardly soluble C-14 is generated in PWRs. However, the chemical form of this hardly soluble C-14 is little known. In this study, the thermal decomposition behavior of particles containing C-14 and the mass spectra of gases released through thermal decomposition were analyzed in order to examine the chemical form of C-14 generated in PWRs. In this study, the gases released during the thermogravimetric (TG) analysis were partially oxidized to CO2, trapped in an alkaline solution and analysed for C-14. Another part of the gas was analysed directly by mass spectrometry (MS). The residues obtained after TG analysis were also analysed for C-14 by oxidizing the residue to CO2, trapping the CO2 in alkaline solutions and analysing C-14 by LSC. During TG-analysis in inert gas (He) atmosphere, about 90% of C-14 was found in residue, while when air was used during TG-analysis, no C-14 could be detected in residue. From the MS-analysis of species released during TG-analysis in inert gas, fragments regarded as originating from ion-exchange resins were detected in released gases. Based on this result, it was found that while substances originating from ion-exchange resin were present in radioactive particles generated in a PWR, the main part of C-14 was contained in the residue after heating in the form of thermally stable substances, easy to be oxidized by air at high temperature. It was not possible to determine their exact chemical composition in this work but also, In addition, sorption of insoluble C-14 to cementitious materials was preliminarily examined. As a results, the concentration of insoluble C-14 decreased greatly in 7 days. That means insoluble C-14 tended not to stay in water. Elucidation of the sorption mechanism in the disposal environment of these C-14 is also a future task.