The two-phase systems of pure diluent FS-13 – aqueous solution of 14 mol/L nitric acid and 0.02 mol/L solution of CMPO in diluent FS-13 – aqueous solution of 14 mol/L HNO3 were studied at autoclave temperatures 170 °C and 200 °C. The effect of pre-irradiation on the kinetics of thermolysis was determined. All samples were irradiated using an electron accelerator at a dose rate of 10 kGy/h up to absorbed doses of 0.1, 0.5 and 1 MGy. The parameters of heat and gas emissions were determined during thermolysis of the studied extraction system in a closed apparatus. It has been shown that the conditions required for the growth of autocatalytic oxidation are not created during heating of the two-phase systems in closed vessels even for the extraction systems contacted with 14 mol/l HNO3 for two weeks.

Plutonium oxide (PuO2) is one of the most highly radioactive components of nuclear fuel waste streams and its storage poses particular challenges due to the high temperatures produced by its decay and the production of gases (particularly H2 and steam). Its high radiotoxicity necessitates the use of analogues, such as CeO2, to allow the comprehensive study of its interaction with water under storage conditions.

We have developed a method which enables direct gravimetric measurement of water adsorption onto CeO2 thin films with masses in the microgram region. Porous CeO2 films were fabricated from a surfactant based precursor solution. The absorption of water onto the CeO2 coating at different relative humidities was studied in a closed reactor. Quartz Crystal Microbalance (QCM) gravimetry was used as a signal transducer, as changes in crystal resonant frequency due to absorbed mass are directly and linearly related to mass changes occurring at the crystal surface. Using this method, we have determined the enthalpy of absorption of water onto CeO2 to be 49.7 kJmol–1 at 75°C, 11 kJmol-1 greater than the enthalpy of evaporation. This enthalpy is within the range predicted for the absorption of water onto PuO2, indicating this method allows for investigation of water absorption using microgram samples.

Spent nuclear fuel from Finnish power plants is planned to be deposited deep in the crystalline bedrock. The bedrock needs to be well characterized to properly assess the risks inherent to the long term safety of the site. In the bedrock the possibly released radionuclides are assumed to be mainly transported by water conducting fractures and their transport is retarded by matrix diffusion and sorption. In this work transport properties of an 80 centimeters long veined gneiss drill core sample from Olkiluoto was studied using an advection-matrix diffusion experiment, which was developed to demonstrate the effect of rock matrix for transport of radionuclides in an advective fracture. The experiment was performed using H-3 and Cl-36 as tracers, and effective diffusion coefficients (De) of (1.7 ± 0.7) × 10-13 m2/s and (1.4 ± 1.0) × 10-14 m2/s and porosities of 1.1 ± 0.3 % and 0.23 ± 0.10 % were determined, respectively. A lower porosity and De for Cl-36 than for H-3 indicates an effect of anion exclusion and the results were found to be in agreement with previous laboratory experiments. However, the comparison to results from a similar in-situ experiment showed that the transport of H-3 and Cl-36 is retarded more in laboratory than in in-situ conditions by matrix diffusion.

The International Simple Glass (ISG) is considered as reference benchmark glass and is used in the frame of an international collaboration for the study of the dissolution mechanisms of high-level vitrified nuclear waste.

In this work the forward dissolution rate of the ISG was determined in different alkaline solutions, as a simulation of the disposal conditions foreseen by the Belgian concept for geological disposal of vitrified waste. The determination of the forward dissolution rate was done by measuring the Si released from the glass in solution in tests performed at 30 °C in four different KOH solutions with pH varying from 9 to 14 and in artificial cementitious water at pH 13.5.

The forward dissolution rates determined for the ISG in high pH solutions are in good agreement with the results obtained for a lower pH range.

The rates obtained in this study, moreover, were compared with the rates measured in the same conditions for SON68 glass from a previous work. The values obtained for the two glasses are comparable in artificial cementitious water and in KOH at moderately alkaline pH. At higher pH, ISG glass shows a lower forward dissolution rate with respect to SON68 (0.20 g·m-2·d for ISG and 0.35 g·m-2 d for SON68 at pH 14).

This study details the characterization of a glass sample exposed to hyperalkaline water and calcium-rich sediment for an extended time period (estimated as 2 - 70 years) at a lime (CaO) waste site in the UK. We introduce this site, known as Peak Dale, in reference to its use as a natural analogue for nuclear waste glass dissolution in the high pH environment of a cementitious engineered barrier of a geological disposal facility. In particular, a preliminary assessment of alteration layer chemistry and morphology is described and the initiation of a long-term durability assessment is outlined.

A good understanding of the spent fuel matrix (UO2) behavior under predisposal activities conditions is required for the proper performance assessment of a final repository. Hence, the oxidation evolution of UO2 under dry interim storage conditions, as a main predisposal action within the Spanish strategy, needs to be addressed. For this aim, in this work a detailed in situ Raman spectroscopy study of the surface oxidation of a UO2.00 disk heated in the presence of synthetic air at 573 K is presented. The spectra analysis required two previous studies. In the first one, UO2+x powder samples with controlled degree of non-stoichiometry were identified by thermogravimetric analysis and subsequently characterized by Raman spectroscopy. The equations obtained from this study enable estimating the oxidation degree of any UO2+x sample (for x < 0.20) at atmospheric conditions. The second one was performed in order to use these equations for the in situ experiments (at 573 K), since the shift of the bands due to temperature needs to be taken into account. Thus, the behavior of the Raman spectra as a function of temperature was analyzed and a correction term thereafter introduced in the initial equations.

The Swedish spent nuclear fuel canister design KBS-3 consists of a cylindrical copper shell surrounding an iron insert that holds the spent fuel. Like in most other canister designs the mass of iron constitutes the majority of the canister weight. In order for groundwater to access the spent fuel in a future repository the copper shell must fail and iron corrosion occur. Spent nuclear fuel dissolution will therefor likely proceed under conditions of simultaneous anoxic iron corrosion. The iron corrosion can likely suppress the spent fuel release by creation of strongly reducing conditions from Fe(II) formation and the generation of large quantities of hydrogen. Redox sensitive radionuclides may either be reductively precipitated by dissolved Fe(II) or from interaction with iron corrosion products such a magnetite or green rusts. The generated hydrogen (up to several MPa) may also inhibit the spent nuclear fuel dissolution at the surface of the fuel via the so called hydrogen effect. In order to probe these effects an autoclave experiment was performed in which a basket with PWR spent nuclear fuel (burnup ∼43 MWd/kgU) was suspended in an autoclave containing a simplified groundwater (10 mM NaCl, 2 mM NaHCO3) with iron powder. The autoclave was sparged and pressurized with argon. Following an initial rise in radionuclide concentrations from dissolution of pre-oxidised phases the U concentration dropped to 3x10-9 M within 76 days, in-line with the solubility of amorphous UO2, expected to form under reducing conditions. Any Cs and Sr release also ceased within 223 days indicating complete transition from dissolution of pre-oxidized phases and instant release fractions to conditions with inhibition of the dissolution of the fuel matrix. Gas phase analysis and pressure monitoring showed a steady build-up of hydrogen at a rate higher than what could be attributed to radiolysis, reaching hydrogen partial pressures of several hundred kPa. The results indicate continuous corrosion of iron, with magnetite as the dominating iron corrosion product.

A disposal measure for fuel debris generated at the accident in the Fukushima Daiichi Nuclear Power Station has been studied so far. However, physical and chemical properties of the fuel debris located in reactor containment vessels have not yet been obtained. In order to investigate the safety function of barriers required for disposal of fuel debris, sensitivity analysis for radionuclide migration was carried out, considering with uncertainty of the properties. As a result, it is indicated that it was important for evaluation of fuel debris disposal to obtain the physical and chemical properties of 14C and 129I during release to groundwater, in addition to 238U.

In disposal of high-level radioactive waste, carbon steel overpack will be corroded after closure of the repository, creating a reducing, low-pH environment around the repository. A plutonium diffusion experiment was performed over 15 years with Kunigel V1, which is a typical Japanese bentonite that contains about 50% montmorillonite, in contact with an iron coupon. A tracer solution (10 µL) containing 1 kBq of 238Pu was applied at the interface between the iron coupon and compacted bentonite that was saturated with deionized water. After the diffusion period, the plutonium distribution in the bentonite specimen was measured with an alpha scintillation counter, and the iron and sodium distributions were obtained by inductively coupled plasma-mass spectrometry. Plutonium penetrated into the bentonite to a depth of 2 mm, and more than 90% of plutonium remained in corrosion product at the interface. The bentonite around the interface was dark green like green rust or magnetite according to visual observation. Iron was detected throughout the bentonite and there was a particularly high iron concentration from the interface to a depth of 2 mm, whereas the sodium concentration decreased slightly from the interface to 2 mm. We proposed that ferrous ions diffused into bentonite as the iron coupon corroded and precipitated such as magnetite. The magnetite precipitation would decrease the bentonite pH, resulting in the dissolution and migration of plutonium. Small cracks were also observed 2 to 6 mm from the interface and could have been formed by the migration of hydrogen generated by corrosion.