Osmosis-induced swelling of Eurobitum bituminized radioactive waste in constant total stress conditions

In geological disposal conditions, contact of Eurobitum bituminized radioactive waste, which contains high amounts of the hygroscopic and highly soluble NaNO₃, with groundwater will result in water uptake and swelling of the waste, and in subsequent leaching of the embedded NaNO₃ and radionuclides. The swelling of and the NaNO₃ leaching from non-radioactive Eurobitum samples, comprised between two stainless steel filters and in contact with 0.1 M KOH, was studied in restricted (semi-confined) swelling conditions, i.e. under a constant total stress, or counterpressure, of 2.2, 3.3, or 4.4 MPa (i.e. oedometer conditions). Four tests were stopped after hydration times between 800 and 1500 days, and the samples were analyzed by micro-focus X-ray Computer Tomography (μCT) and by Environmental Scanning Electron Microscopy (ESEM). The complete set of data enabled a consistent interpretation of the observations and lead to an improved understanding of the phenomenology of the water uptake, swelling, and NaNO₃ leaching in restricted swelling conditions. Under the studied conditions, the bituminous matrix surrounding the NaNO₃ crystals and pores with NaNO₃ solution behaved as a highly efficient semi-permeable membrane, i.e. osmotic processes occurred. In the main part of the leached layers, a high average NaNO₃ concentration and related to this a high osmotic pressure prevailed, explaining why in the studied range the swelling was not measurably affected by the counterpressure. At the interface with the stainless steel filters, a low permeable re-compressed bitumen layer was formed, contributing to the slow release of NaNO₃ compared to the water uptake rate. A fully coupled Chemo-Hydro-Mechanical (CHM) constitutive model has been developed that integrates the key processes involved and that reproduces satisfactorily the results; this is presented in another work. Combination of the experimental and the modelling study allow to conclude that under semi-confined conditions the swelling of the bituminized waste, and its evolution with time, is the result of several transient processes (salts dissolution, diffusion of salts and water, advection, creep, involving a low permeability material with evolving thickness and properties) that moreover are non-linear and strongly coupled.     

双层孔隙介质中非饱和渗流的水势分布与变化

水势反映土壤水的能态,而水势梯度则是决定水流方向和速度的重要因素。本文主要介绍非饱和水通过以黄土(细颗粒)和石英砂(粗颗粒)组成的双层孔隙介质渗流的水势分布和变化的实验方法和主要实验结果,用以进一步分析解释以前工作中观察到的非饱和水在双层孔隙介质中的绕流现象。实验结果表明：在喷水强度一定情况下,石英砂层中的水势随石英砂层厚度的增加而减小;在石英砂层厚度一定情况下,石英砂层中的水势随喷水强度的增加而增大;石英砂层上方的水部分通过石英砂层向下流动,部分绕过石英砂层向下流动;绕过石英砂层的水部分地紧贴石英砂层下表面向石英砂层下方的黄土中流动。这是一些值得进一步探讨的现象。     

非饱和水通过双层孔隙介质渗流的定量实验

非饱和水通过双层孔隙介质的渗流对于放射性废物近地表处置库的顶盖设计具有重要意义。本文主要介绍非饱和水通过以黄土（细颗粒）和石英砂（粗颗粒）组成的双层孔隙介质渗流定量实验的装置、方法及其结果。从实验结果可见：黄土中的非饱和水在下渗过程中遇到石英砂层时，即使石英砂层很薄，且粒度很小，下渗水也发生绕流现象；相对绕流量的大小随水的喷淋强度的增大而减小，随石英砂层厚度的增加而增加。另外，实验发现，相当部分的绕流水是紧石英砂层的下表面运动的。由实验结果推断，在石英砂粒径0.2-0.45mm、埋深20cm情况下，不起绕流作用的石英砂层厚度≤1mm；对于≥2mm厚的石英砂层，相对绕流量达100％的喷淋强度需低于5mm/d。     

Magnesium potassium phosphate ceramic for Tc immobilization

Technetium-99, present in the US Department of Energy’s (DOE) high-level waste (HLW) as a by-product of fission reactions, poses a serious environmental threat because it has a long half-life, is highly mobile in its soluble Tc⁷⁺ oxidation state and is volatile at high temperatures. Magnesium potassium phosphate (MKP) ceramics have been developed to treat ⁹⁹Tc that has been partitioned and eluted from simulated high-level tank wastes by means of sorption processes. Waste forms were fabricated by adding MKP binder and a reducing agent (SnCl₂) to the ⁹⁹Tc-containing aqueous waste. In addition, waste forms were fabricated by first precipitating ⁹⁹Tc from the waste and subsequently solidifying it in MKP. ⁹⁹Tc loadings in the waste forms were as high as 900 ppm by weight. Waste form performance was established through various strength, leaching, and durability tests. Long-term leaching studies, as per the ANS 16.1 procedure, showed leachability indices between 11 and 14 for ⁹⁹Tc under ambient conditions. The normalized leach rate for ⁹⁹Tc, according to the product consistency test, was as low as 1.1 × 10⁻³ g/m² d. The waste forms exhibited a compressive strength of ≈30 MPa and were durable in an aqueous environment. Containment of ⁹⁹Tc in MKP ceramics is believed to be due to a combination of appropriate reducing environment (determined from Eh-pH measurements) and microencapsulation in a dense matrix.     

Raman and X-ray absorption spectroscopic studies of hydrothermally altered alkali-borosilicate nuclear waste glass

Raman spectroscopy and X-ray absorption spectroscopy (XAS) are used to characterize structural changes that took place in hydrothermally altered (Na,K)-alumina-borosilicate glasses with different Na/K ratios, formulated as part of a durability study to investigate the behavior of glasses for nuclear waste storage. The hydrothermal experiments, or vapor hydration tests (VHT), were performed on each glass for 3 and 20 days at 200 °C to accelerate and approximate long-term alteration processes that may occur in a nuclear waste repository. Results found for both glasses and their VHT altered counterparts show little, if any, structural influence from the different starting Na/K ratios. X-ray diffraction, differential scanning calorimetry, scanning electron microscopy, and Raman spectroscopy indicate that the altered samples are mostly amorphous with small amounts of analcime-like and leucite-like crystals within 200 μm of the sample surface and contain up to 9.7 wt.% water or OH. The Raman data are nearly identical for the amorphous portions of all altered VHT samples investigated, and indicate that two glass structural changes took place during alteration: one, partial depolymerization of the alumina-borosilicate network, and two, introduction of water or OH. Al and Si XAS data indicate tetrahedral AlO₄ and SiO₄ environments in the original glasses as well as in the altered samples. Small energy shifts of the Si K-edge also show that the altered VHT samples have less polymerized networks than the original glass. Na XAS data indicate expanded Na environments in the VHT samples with longer Na-O distances and more nearest-neighbor oxygen atoms, compared with the original glasses, which may be due to hydrous species introduced into the expanding Na-sites.     

Distribution of trace impurities of fluorocarbons in the condensed phase – vapor system based on uranium hexafluoride in equilibrium and during evaporation

The results of investigations of the distribution of fluorocarbons in two-phase solid phase–gas and liquid–vapor systems based on uranium hexafluoride are presented. The systems are studied under equilibrium conditions, under which a constant distribution of fluorocarbons is established between the phases in a closed volume, and under nonequilibrium conditions created with stationary removal of the gas phase from the volume with liquid or solid uranium hexafluoride. The investigations were performed on specially prepared standard mixtures of uranium hexafluoride with fluorocarbon fraction from 10^–4 to 10^–1%. The special features of the evaporation processes in the two-phase systems studied are shown, the distribution laws of the fluorocarbons between phases of the main substance are established, and the regimes of evaporation of the mixtures with the smallest disruptions in the uniformity of the fluorocarbon inflow with the flow of the evaporated uranium hexafluoride are determined.     

A numerical and experimental study of water ingression phenomena in melt pool coolability

During a postulated severe accident, the core can melt and the melt can fail the reactor vessel. Subsequently, the molten corium can be relocated in the containment cavity forming a melt pool. The melt pool can be flooded with water at the top for quenching it. However, the question that arises is to what extent the water can ingress in the corium melt pool to cool and quench it. To reveal that, a numerical study has been carried out using the computer code MELCOOL. The code considers the heat transfer behaviour in axial and radial directions from the molten pool to the overlaying water, crust generation and growth, thermal stresses built-in the crust, disintegration of crust into debris, natural convection heat transfer in debris and water ingression into the debris bed. To validate the computer code, experiments were conducted in a facility named as core melt coolability (COMECO). The facility consists of a test section (200 mm × 200 mm square cross-section) and with a height of 300 mm. About 14 L of melt comprising of 30% CaO + 70% B₂O₃ (by wt.) was poured into the test section. The melt was heated by four heaters from outside the test section to simulate the decay heat of corium. The melt was water flooded from the top, and the depth of water pool was kept constant at around 700 mm throughout the experiment. The transient temperature behaviour in the melt pool at different axial and radial locations was measured with 24 K-type thermocouples and the steam flow rate was measured using a vortex flow meter. The melt temperature measurements indicated that water could ingress only up to a certain depth into the melt pool. The MELCOOL predictions were compared with the test data for the temperature distribution inside the molten pool. The code was found to simulate the quenching behaviour and depth of water ingression quite well.     

Experimental investigation on critical heat flux in horizontal channel for low pressure low flow (LPLF) condition

Studies reported in the past on critical heat flux (CHF) are mostly limited to vertical flow, large channel diameter, high pressure and high mass flux. Only few investigations are reported in the literature for horizontal flow CHF especially under low pressure and low flow conditions. Hence, predictive methods of CHF for horizontal flow are scarce. There is a need for understanding CHF in horizontal flow under low pressure and low flow conditions because they are commonly encountered in nuclear reactor fuel channels of pressurized heavy water reactor (PHWR) under loss of coolant accidental (LOCA) conditions. The present work investigates CHF of horizontal flow for low flow rates (mass flux of 100–400 kg/m² s) at nearly atmospheric pressure conditions. Parameters covered in this study are diameter (5.5 mm, 7.5 mm and 9.5 mm), length (0.45 m and 0.8 m) and a inlet temperature of 32 °C. The first occurrence of ‘red hot’ spot on the test section is considered as the onset of critical heat flux condition in the present work. Experimental results obtained are compared with Groeneveld et al. (2007) look up table data for vertical flow after applying correction factor given by Wong et al. (1990). The deviation of experimental CHF data from those predicted using Groeneveld et al. (2007) look up table and Wong et al. (1990) correction factor is more than 50%.     

De-agglomeration mechanisms of TiO2 aerosol agglomerates in PWR steam generator tube rupture conditions

A steam generator tube rupture (SGTR) in a pressurized water reactor (PWR) might be a major source of accidental release of radioactive aerosols into the environment during severe accident due to its potential to by-pass the reactor containment. In the ARTIST program, tests were carried out at flow conditions typical to SGTR events to determine the retention of dry aerosol particles inside a steam generator tube. The experiments with TiO₂ agglomerates showed that for high velocities in the range of 100-350 m/s, the average particle size at the outlet of the tube was significantly smaller than at the inlet due to particle de-agglomeration. Earlier, particle de-agglomeration has not been considered significant in nuclear reactor severe accidents. However, the tests in ARTIST program have shown that there is a possibility that TiO₂ aerosol particles de-agglomerate inside a tube and in the expansion zone after the tube exit under SGTR conditions.In this investigation, we measured TiO₂ aerosol de-agglomeration in the tube with very high flow velocities with two different TiO₂ aerosols. The de-agglomeration was determined by measuring the size of the agglomerates at the inlet and outlet of the test section. The test section was composed of tubes with three different lengths, 0.20, 2.0 and 4.0 m, followed by an expansion zone.The main results were: (i) the de-agglomerate process was relatively insensitive to the initial particle size distribution, (ii) the agglomerates were observed to de-agglomerate in all the tubes, and the resulting particle size distributions were similar for both TiO₂ aerosols, (iii) at high flow rates, increasing the gas mass flow rate did not produce further de-agglomeration, and (iv) the agglomerates did not de-agglomerate to primary particles. Instead, after de-agglomeration the particles had a median outer diameter D_c = 1.1 μm. Based on analysis using computational fluid dynamics (CFDs), the de-agglomeration was caused by the turbulent shear stresses due to the fluid velocity difference across the agglomerates in the viscous subrange of turbulence.It has to be noted that the particles used in this investigation were TiO₂ agglomerates, and not prototypical nuclear aerosols with significantly different characteristics. Therefore, the results of this investigation cannot be directly used to determine whether the nuclear aerosol particles may de-agglomerate in SGTR sequences. However, this investigation highlights the possibility of particle de-agglomeration under SGTR conditions, and identifies the mechanism of the de-agglomeration inside the broken tube and when the aerosol is discharged to an open space.     

Low-temperature immobilization of actinides and other components of high-level waste in magnesium potassium phosphate matrices

The magnesium potassium phosphate (MPP) matrices formed by the room-temperature reaction of MgO and KH₂PO₄ solution were used for immobilization of simulated liquid alkaline high-level waste (HLW) containing actinides, fission and corrosion products. Novel procedures of solidification of HLW simulants were developed to increase stability of the MPP matrices to leaching radionuclides (Pu, Np, Am, Cs, Sr, Tc, I and Se), matrix-forming (K, Mg and PO₄) and admixture components (NO₃, NO₂, Na and others) as well as hazardous elements (Pb, Cr, Zn and others) according to the ANS, PCT, TCLP standards. Density (∼1.7 g/cm³), compressive strength (>20 MPa), radiation resistance of the matrices and chemical yield of radiolytic hydrogen (0.004 molecule H₂/100 eV) were determined. The phase composition of the matrices and distribution of radionuclides were studied by XRD and autoradiography correspondingly.     

Sensitivity analysis of numerically determined linear stability boundaries of a supercritical heated channel

The large change in density which occurs when supercritical water is heated above or near to the pseudocritical temperature in a vertical channel can result in the onset of flow instabilities (density wave oscillations). Near to the critical point, substance properties such as enthalpy, density, viscosity, etc. all have larger relative uncertainties compared to subcritical conditions. The goal of this study is to quantify the effect of these property uncertainties and system uncertainties on numerically determined stability boundaries. These boundaries were determined through an eigenvalue analysis of the linearised set of equations. The sensitivity analysis is performed in a forward way. The results show that the impact of the density and viscosity tolerance individually as well as that of the uncertainty of the imposed pressure drop are negligible. The tolerance on the derivative of the density with regard to the enthalpy propagates only noticeably at low N_SUB numbers (T_in > 370 °C). The friction factor and the heat flux distribution uncertainties have a comparable effect, being more pronounced near the bend in the stability curve. The most significant uncertainty was found to be that of the geometry, even a ±25 μm uncertainty on length scales results in a large uncertainty. The results also showed that the stability boundary is linked to the friction distribution rather than its average value, and that different correlations result in strong changes of the predicted boundary. This emphasizes the need for an accurate friction correlation for supercritical fluids. These findings are important to assess the design of experimental facilities which use scaling fluids.     

In-vessel melt pool coolibility test—Description and results of LIVE experiments

The LIVE test program investigates in-vessel melt pool behaviour and cooling strategies for in-vessel corium retention during severe accidents in LWRs. The main part of the LIVE facility is a 1:5 scaled semi-spherical lower head of a typical pressurized water reactor. Up to now, LIVE experiments have been performed in different external cooling conditions, melt volumes and heat generation rates. At present the well-known simulant material KNO₃-NaNO₃ in non-eutectic composition (80 mole% KNO₃-20 mole% NaNO₃) and in eutectic composition (50 mole% KNO₃-50 mole% NaNO₃) is used. This work presents the behaviour of a homogenous melt pool regarding the 3D heat flux distribution through vessel wall, melt pool temperature, crust thickness and the pool melt composition in transient and in steady state conditions.     

Critical heat fluxes of subcooled water flow boiling in a short vertical tube at high liquid Reynolds number

The steady state critical heat fluxes (CHFs) and the heat transfer of the subcooled water flow boiling for the flow velocities (u = 17.2-42.4 m/s), the inlet subcoolings (ΔT_sub,in = 80.9-147.6 K), the inlet pressures (P_in = 812.1-1181.5 kPa) and the exponentially increasing heat input (Q₀ exp(t/τ), τ = 8.5 s) are systematically measured by the experimental water loop comprised of a new multi-stage canned-type circulation pump with high pump head. The SUS304 test tube of inner diameter (d = 6 mm), heated length (L = 59.5 mm), L/d = 9.92 and wall thickness (δ = 0.5 mm) with surface roughness (Ra = 3.18 μm) is used in this work. The steady state CHFs of the subcooled water flow boiling for the flow velocities ranging from 17.2 to 42.4 m/s are clarified. The steady state CHFs are compared with the values calculated by our transient CHF correlations against outlet and inlet subcoolings based on the experimental data for the flow velocities ranging from 4.0 to 13.3 m/s. The influence of flow velocity at high liquid Reynolds number on the subcooled flow boiling CHF is investigated in detail and the widely and precisely predictable correlations of the transient CHF correlations against outlet and inlet subcoolings in a short vertical tube are derived based on the experimental data at high liquid Reynolds number. The transient CHF correlations can describe the subcooled flow boiling CHFs for the wide range of flow velocities at high liquid Reynolds number obtained in this work within ±15% difference.     

Monte Carlo criticality analysis of simple geometries containing tungsten-rhenium alloys engrained with uranium dioxide and uranium mononitride

The critical mass and dimensions of simple geometries containing highly enriched uranium dioxide (UO₂) and uranium mononitride (UN) encapsulated in tungsten-rhenium alloys are determined using MCNP5 criticality calculations. Spheres as well as cylinders with length to radius ratios of 1.82 are computationally built to consist of 60 vol.% fuel and 40 vol.% metal matrix. Within the geometries, the uranium is enriched to 93 wt.% uranium-235 and the rhenium content within the metal alloy was modeled over the range of 0-30 at.%. The spheres containing UO₂ were determined to have a critical radius of 18.29-19.11 cm and a critical mass ranging from 366 kg to 424 kg. The cylinders containing UO₂ were found to have a critical radius ranging from 17.07 cm to 17.84 cm with a corresponding critical mass of 406-471 kg. Spheres engrained with UN were determined to have a critical radius ranging from 14.82 cm to 15.19 cm and a critical mass between 222 kg and 242 kg. Cylinders which were engrained with UN were determined to have a critical radius ranging from 13.81 cm to 14.15 cm and a corresponding critical mass of 245-267 kg. The critical geometries were also computationally submerged in a neutronically infinite medium of fresh water to determine the effects of rhenium addition on criticality accidents due to water submersion. The Monte Carlo analysis demonstrated that rhenium addition of up to 30 at.% can reduce the excess reactivity due to water submersion by up to $5.07 for UO₂ fueled cylinders, $3.87 for UO₂ fueled spheres and approximately $3.00 for UN fueled spheres and cylinders.     

Analysis of the processes in spent fuel pools of Ignalina NPP in case of loss of heat removal

Ignalina NPP is the only nuclear power plant in Lithuania consisting of two units, commissioned in 1983 and 1987. Unit 1 of Ignalina NPP was shutdown for decommissioning at the end of 2004 and Unit 2 is to be operated until the end of 2009. Both units are equipped with channel-type graphite-moderated boiling water reactors RBMK-1500. According to the design, the spent fuel should be returned for reprocessing to Russia. However actually any fuel assembly has not been taken out from territory of the Ignalina NPP and all assemblies of spent fuel are stored in the spent fuel pools and dry on-site storage facility. Thus, the safety justification of facilities for intermediate spent fuel assemblies’ storage in Ignalina NPP is very important. This paper presents the results of loss of heat removal accidents (the most probable beyond design basis accident) in spent fuel pools of Ignalina NPP. The analysis was performed by employing best-estimate system thermal hydraulic code RELAP5 and codes for severe accidents ATHLET-CD and ASTEC. The best-estimate analysis, performed using RELAP5, allowed to investigate in the details the water evaporation, uncovering and fuel assemblies heat-up processes, when heat removal from the structures of buildings and pools are evaluated. The processes of spent fuel assemblies’ degradation due to loss of long-term heat removal were analyzed using ATHLET-CD and ASTEC codes. The results of calculations showed that the increase in water temperature in the pools from 50 °C up to 100 °C takes approximately 80-110 h, the evaporation of water volume down to uncovering of fuel assemblies takes approximately 220-260 additional hours. Later, after 200-300 h, the temperature of fuel claddings exceeds 800-1000 °C and the failures of fuel claddings occur due to cladding ballooning. The total amount of hydrogen generated up to time of complete water evaporation from spent fuel pools is about 7500-16,000 kg. These results of performed analysis were used for development of accident management guidelines for spent fuel pools of RBMK-1500.     

Absolute molecular flux and angular distribution measurements to characterize DNA/RNA vapor jets

Vapor jets of DNA and RNA bases (adenine, cytosine, thymine, and uracil) from an oven with a capillary exit have been studied in the intermediate regime between molecular and viscous flow corresponding to Knudsen numbers in the range 0.1 < K_n < 10. The temperature control method ensured stationary flow. Assuming the Knudsen hypothesis, the pressure of sublimated molecules in the oven was determined as a function of temperature and the transmission probability of the capillary (Clausing factor). Thus it was possible to relate the oven temperature and pressure to the total flux through the capillary, determined by measuring the total mass of DNA/RNA base molecules condensed on a cold surface intersecting the jet. The angular distribution of molecules in the jet has been also studied experimentally using an optical interference method. The measured profiles are in good agreement with Troïtskii’s [Sov. Phys. JETP 7 (1962) 353] analytical law for (cos θ)^3/2 angular dependence in the intermediate regime with error functions associated with the mean free path between intermolecular collisions.     

Titanate ceramics for immobilisation of uranium-rich radioactive wastes arising from Mo production

Uranium-rich liquid wastes arising from UO₂ targets which have been neutron-irradiated to generate medical radioisotopes such as ^99mTc require immobilisation. A pyrochlore-rich hot isostatically pressed titanate ceramic can accommodate at least 40 wt% of such waste expressed on an oxide basis. In this paper, the baseline waste form composition (containing 40 wt% UO₂) was adjusted in two ways: (a) varying the UO₂ loading with constant precursor oxide materials, (b) varying the precursor composition with constant waste loading of UO₂. This resulted in the samples having a similar phase assemblage but the amounts of each phase varied. The oxidation states of U in selected samples were determined using diffuse reflection spectroscopy (DRS) and electron energy loss spectroscopy (EELS). Leaching studies showed that there was no significant difference in the normalised elemental release rates and the normalised release rates are comparable with those from synroc-C. This demonstrates that waste forms based on titanate ceramics are robust and flexible for the immobilisation of U-rich waste streams from radioisotope processing.     

Desorption of water adsorbed on iron oxide by laser irradiation

Desorption of water adsorbed on iron oxide by laser irradiation was studied by means of a time-of-flight (TOF) technique. The wavelength of the laser for desorption was varied from 355 to 600 nm. The energy threshold of the water desorption ranged around 2.0-2.3 eV. Based on the fact that this energy threshold approximately corresponds to the bandgap of Fe₂O₃, the initial process of water desorption is considered to be the electronic excitation of the iron oxide from the valence band to the conduction band. Analysis of the velocity distribution of the desorbed water suggests that following the electronic excitation of the iron oxide the desorption is caused by both thermal and nonthermal processes. The thermal process is caused by the rise of the surface temperature that occurs after the scattering and de-excitation of the excited electron in the iron oxide. In the case of the laser at λ = 355 nm, the desorption was mainly caused by the thermal process. On the other hand, in the case of the laser at λ = 430 nm, the desorption was mainly caused by the nonthermal process. The desorption caused by the nonthermal process is attributed to the transfer of the electron excited in the iron oxide to the adsorbed hydroxyl.     

Monte Carlo simulation of nonlinear reactive contaminant transport in unsaturated porous media

In the current proposed solutions of radioactive waste repositories, the protective function against the radionuclide water-driven transport back to the biosphere is to be provided by an integrated system of engineered and natural geologic barriers. The occurrence of several nonlinear interactions during the radionuclide migration process may render burdensome the classical analytical–numerical approaches. Moreover, the heterogeneity of the barriers’ media forces approximations to the classical analytical–numerical models, thus reducing their fidelity to reality. In an attempt to overcome these difficulties, in the present paper we adopt a Monte Carlo simulation approach, previously developed on the basis of the Kolmogorov–Dmitriev theory of branching stochastic processes. The approach is here extended for describing transport through unsaturated porous media under transient flow conditions and in presence of nonlinear interchange phenomena between the liquid and solid phases. This generalization entails the determination of the functional dependence of the parameters of the proposed transport model from the water content and from the contaminant concentration, which change in space and time during the water infiltration process. The corresponding Monte Carlo simulation approach is verified with respect to a case of nonreactive transport under transient unsaturated flow and to a case of nonlinear reactive transport under stationary saturated flow. Numerical applications regarding linear and nonlinear reactive transport under transient unsaturated flow are reported.     

A critical examination of the thermodynamics of water adsorption on actinide oxide surfaces

The reversible adsorption of water from actinide oxide surfaces is examined from several viewpoints in this article. A reinterpretation and critical look at the previously published thermodynamic values for desorption of water from PuO₂ [J. Phys. Chem. 77 (1973) 581] are reexamined in light of more recent mathematical treatments of thermal desorption data from high surface area materials. In addition, the time and temperature dependent process of water adsorption/desorption in closed system experiments is examined using chemical kinetics modeling. A simple experimental method and mathematical treatment of determining adsorption enthalpies based upon a closed system is also described. The desorption enthalpy for reversibly adsorbed water from PuO₂ is determined to be a function of adsorbate coverage with values ranging from 51 to 44 kJ mol⁻¹ for coverages of one to several monolayers (MLs). Consistent desorption enthalpy values are obtained using either approach thus highlighting the importance of proper interpretation of adsorption parameters determined from high surface area powders. Reversible adsorption/desorption equilibrium of water with actinide oxide materials is discussed from the practical standpoint of storage and subsequent pressurization of containers. These results obtained from PuO₂ surfaces are consistent with desorption enthalpies of water from a low surface area UO₂ that has been measured using ultra-high vacuum thermal desorption mass spectroscopy to be 42.2 kJ mol⁻¹.