Uranophane [Ca(UO2)2(SiO3OH)2 · 5H2O] is a corrosion product of long-term leaching of spent fuel under oxidizing conditions and is a weathering product of uraninite in uranium ore deposits hosted by siliceous rocks. Incorporation of radionuclides into uranophane by coprecipitation may occur as a result of spent fuel alteration. Dissolution of uranophane leading to release of these radionuclides may therefore influence the longterm dissolved concentration and mobility of radionuclides at the proposed nuclear waste repository at Yucca Mountain, Nevada. In this study, the dissolution of uranophane in Ca- and Si-rich test solutions was investigated. Batch dissolution experiments were designed to approach uranophane equilibrium from undersaturated solutions at nearneutral pH (~6.0). Test solutions had initial U concentrations of 0.0 and 10-7 mol/L in matrices of ~10-2 mol/L CaCl2 and ~10-3 mol/L SiO2(aq). The test solutions were reacted with synthetic uranophane (confirmed by XRD and chemical analyses) and analyzed periodically over 10 weeks. Reaction quotients (Log Qs) derived from aqueous speciations of experimental solutions considered to be near equilibrium with uranophane ranged from 10.54 to 11.06 for the dissolution reaction: Ca(UO2)2(SiO3OH)2 · 5H2O + 6H+ ⇔ Ca2+ + 2UO22+ + 2SiO2(aq) + 9H2O.

Garnet phases have been considered as a durable crystalline waste form for hosting actinide. The garnet-structure phases with stoichiometries of Ca2,5Ce0,5Zr2Fe3O12, Ca2CeZrFeFe3O12, and Ca1,5GdCe0,5ZrFeFe3O12 were synthesized through cold pressing and sintering in air and oxygen to determine the optimum parameters for the formation of actinide waste forms. Cerium (Ce) was used as an imitator of plutonium due to its similarity in oxidation state and ionic radii. Gadolinium (Gd) plays a major role as an absorber of neutrons that prevents nuclear chain reaction. It also serves as imitators of the trivalent actinides. Ce-garnet or (Ce,Gd)-garnet is chemically analogous to the garnets with plutonium and/or trivalent actinides. The results of XRD and SEM-EDS examination of the products of experiments reveal that equilibrium state was reached at the temperatures of 1300 °C and 1200 °C for 1 and 5 hours heating, respectively.

An approach is presented for determining if the models used to calculate the release of radionuclides from defense high-level radioactive waste (HLW) glass for total system performance assessment (TSPA) calculations can be used to account for the release of radionuclides from waste forms other than standard borosilicate glasses. The fractional release rates of radionuclides due to waste form degradation, the available surface area, and the radionuclide inventory in an alternative waste form can be compared with the corresponding models used in TSPA for HLW glasses to determine if those models adequately represent the waste form. This approach is demonstrated for the ceramic and metallic waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel. Depending on the waste form, comparisons made with aspects of the HLW glass model may be based on similarities in degradation mechanisms or purely empirical.

Flowthrough dissolution tests using solutions with pH in the range 2 to 7 have been conducted on a moderate burnup Light Water Reactor spent fuel. Such low pH conditions have been modeled as possibly occurring in a failed waste package at the proposed repository at Yucca Mountain. The release oftotal uranium, 99Tc, 90Sr, 137Cs, and 239&240Pu were measured for up to 90% total reaction of the specimens. The reaction rates, determined both from the cumulative release and the release normalized to surface area, were found to decrease with increasing pH and with increasing extent of reaction. The implications to instantaneous release and long-term behavior ina geologic repository are discussed.

Chromium is of great concern in the vitrification of high-level nuclear waste sludges because it forms separate crystallites in the molten glass. Inadequate removal of chromium from sludges could result in the production of an unacceptably large volume of HLW glass. Alkaline oxidative leaching is considered one of the pretreatment strategies to remove chromium before the vitrification.

In this study, Cr(III) hydroxide solids were prepared under different conditions and characterized by EXAFS and IR. The rate of oxidation of the solids by hydrogen peroxide in alkaline solutions was studied by UV absorption spectroscopy. EXAFS and IR experiments indicate that the degree of oligomerization in the Cr(III) hydroxide solids increases with the increase in the concentration of NaOH in solution, the aging temperature and the aging time. The rate of oxidation of the solids follows the same order previously observed for Cr(III) oligomers in solution, i.e., species with higher degree of oligomerization are oxidized more slowly.

AECL Whiteshell Laboratories (WL), near Winnipeg, Canada has been in operation since the early 1960s. R&D programs carried out at WL include a 60 MW organic-cooled research reactor, which operated from 1965 to 1985, reactor safety research, small reactor development, materials science, post irradiation examinations, chemistry, biophysics and radiation applications. The Canadian Nuclear Fuel Waste Management Program was conducted and continues to operate at WL and also at the nearby Underground Research Laboratory.

In the late-1990s, AECL began to consolidate research and development activities at its Chalk River Laboratories (CRL) and began preparations for decommissioning WL. Preparations for decommissioning included a staged shutdown of operations, planning documentation and licensing for decommissioning. As a prerequisite to AECL's application for a decommissioning licence, an environmental assessment (EA) was carried out according to Canadian environmental assessment legislation. The EA concluded in 2002 April when the Federal Environment Minister published his decision that WL decommissioning was not likely to cause significant adverse environmental effects and that no further assessment by a review panel or mediation would be required

In 2002 December, the Canadian Nuclear Safety Commission issued a decommissioning licence for WL, valid until December 31, 2008. The licence authorized the first planned phase of site decommissioning as well as the continuation of selected research programs. The six-year licence for Whiteshell Laboratories was the first overall decommissioning license issued for a Canadian Nuclear Research and Test Establishment and was the longest licence term ever granted for a nuclear installation of this complexity in Canada.

The first phase of decommissioning is now underway and focuses on decontamination and modifications to nuclear facilities, such as the shielded facilities, the main R&D laboratories and the associated service systems, to achieve a safe state of storage-with-surveillance. Later phases have planned waste management improvements for selected wastes already in storage, eventually followed by final decommissioning of facilities and infrastructure and removal of most wastes from the site.

This paper provides an overview of the planning, environmental assessment, licensing, and organizational processes for decommissioning and selected descriptions of decommissioning activities currently underway at AECL Whiteshell Laboratories.

The properties of actinide glasses are studied in the context of high-level waste management programs. Reprocessing high burnup fuels in particular will increase the minor actinide content in the glass package, resulting in higher cumulative alpha decay doses in the glass, and raising the question of the glass matrix behavior and especially its containment properties. The effect of alpha self-irradiation on the glass behavior is evaluated by doping the glass with a short-lived actinide (244Cm) to reach in several years the alpha dose received by the future glass packages over several thousand years. “R7T7” borosilicate glasses were doped with 3 different curium contents (0.04 wt%, 0.4 wt% and 1.2 wt% 244CmO2). The glass homogeneity was characterized by SEM and XRD. Chemical analysis after dissolution of the glass confirmed the chemical composition. The curium content and its distribution in the glass specimens were checked by calorimetry measurements and gamma spectrometry on 243Cm and 245Cm isotopes. The density and mechanical properties of the curium-doped glasses were characterized up to 2 × 1018 α/g, revealing only a slight evolution of the macroscopic behavior of R7T7 glass in this range. The leaching behavior of curium-doped glass was also studied by Soxhlet tests. The results do not show any significant evolution of the initial alteration rate with the alpha dose.

For the performance assessment of the radioactive waste repository, it is important to clarify the dynamic behavior of silica (silicic acid and hydrous or unhydrous silicon dioxides). The behavior of silica would be complicated when cement is used for the construction of the repository, since silica takes various forms due to polymerization, precipitation and dissolution with change of pH or temperature. In order to know the fundamental kinetic property of silica, this study has examined the dissolution rate of colloidal-silica.

In the experiment, the concentration of silica in a soluble form was determined by the silicomolybdenum-yellow method. In this study, soluble-silica was defined as silica reacting with molybdate reagent and coloring yellow, and colloidal-silica was defined as silica in liquid phase except for soluble-silica. Colloidal-silica was obtained through the polymerization process, where the pH value of silica solution was brought down from over 10 by HNO3 solution. This study examined dissolution rate of colloidal-silica again by setting to 10 or 13 in pH-value and 288 K, 298 K or 313 K in temperature. In the experimental results, the dissolution reaction of colloidal-silica proceeded linearly with time, when the dissolution of colloidal silica was not restricted by the solubility of silica. To estimate the dissolution rate, we assumed df /dt = k*, where f is the soluble-silica fraction defined as the amount of soluble-silica divided by the silica amount introduced into the sample solution, t the time (s) and k* the rate constant (s-1). The activation energy for the dissolution of the colloidal-silica at pH 13 was estimated to be approximately 80 kJ·mol-1which was similar to that for amorphous silica (solid phase) at pH 13. This suggests the same reaction mechanism for the dissolution of colloidal-silica and amorphous silica in highly alkaline solution.

So-called alpha-doped UO2 (i.e. UO2 containing short-lived alpha-emitters) can reproduce the activity levels of spent fuel at different ages. Previous experiments showed significant effects of alpha-radiolysis at relatively high alpha-activity levels (e.g., in the range 108 – 1010 Bq/g). This study shows the results of leaching tests done on low alpha-activity material (106 – 107 Bq/g). UO2 containing ∼10 and ∼1 wt. % 233U was leached at room temperature in deionized and carbonated water under deaerated conditions. Higher release in carbonated water was observed. A clear radiolytic enhancement of the concentration of uranium in solution was observed for the material containing 10 wt. % 233U. The dissolved U concentration was very close to the values previously observed for higher activity alpha-doped UO2 and confirmed the finding that under relatively high surface/volume condition the measured amount of U in solution is essentially independent of the alpha-activity level. No significant radiolysis effect was detected for the material doped with 1% 233U compared to undoped UO2 after ∼4 months of leaching. The post- leaching SEM examination revealed no pronounced etching of the surfaces.

One of the high levels of actinide, and in particular Cm, waste streams at the Russian radiochemical Production Association (PA) Mayak was generated during spent fuel reprocessing. Using oxalate precipitation, the rare earth elements (REE) and transuranic elements (TRU) settled out in the form of oxalate residues. Due to in high REE contents in this residue, the mineral-like matrix based on (REE)PO4 solid solution, with monlclinic monazite structure have been proposed to use as a suitable ceramics form for final actinide immobilization. For this purpose the synthetic REE oxalates were first transformed into REE orthophosphates in a thin-film evaporator (TFE). Then the (REE)PO4 powder was compacted both by either hot uniaxial pressing (HUP) or cold uniaxial pressing followed by sintering (CUP). This ceramic with the monazite structure has a high density and exhibits chemical durability by leaching.

The solubility values usually used in Belgium for Performance Assessment of the disposal of HLW in deep geological formations are based on migration experiments in the far-field and, thus, can be considered as "far-field solubilities". The particular chemistry in the near-field of the repository will also influence the solubility of the released radionuclides. Therefore, the determination of maximum concentrations of the radionuclides in the near-field is also useful. This was approached by leach tests with doped glass in near-field conditions. The experiment consisted of static leach tests with the french R7T7 reference glass doped with 237Np (0.34 wt%) and 99Tc (0.12 wt%), two key radionuclides in the performance assessment of the disposal system. The glass was put in contact with FoCa7-clay together with pyrite, metallic iron and Boom clay water to simulate the near-field conditions under reducing conditions. The interpretation of the results is based on the measurement of the steady-state Np and Tc concentrations in the leachates after ultrafiltration through 100,000 MWCO membranes and on the activity measured on the clay after different test durations. Most of the Np and Tc activity (99.5%) was found associated to the clay and increasing as a function of time while the Np and Tc concentration in the leachates were stable and extremely low with a mean value of 1.5×10−10M for Np and 5.6×10−10M for Tc. This is due at least partially to sorption on the clay, but there are indications for precipitation as well.

polycrystalline sample of Ti-pyrochlore, (Ca,Gd,Hf,U,Pu)2Ti2O7, doped with approximately 8.7 wt.% 238Pu and 20.0 wt.% 238U, has been studied at different cumulative doses of alpha induced radiation damage using scanning electron microscopy (SEM) and quantitative electron probe microanalysis (EMPA). All analyses of 238Pu-doped Ti-pyrochlore were done for comparison with 239Pu-doped pyrochlore synthesized under similar conditions. Our results suggest that accelerated alpha-induced radiation damage affects not only the crystalline structure of 238Pu-doped pyrochlore, but also causes changes in the chemical composition of the pyrochlore, and as proposed, destruction of pyrochlore-U-Pu solid solution. Numerous inclusions of a separate U-oxide phase were found in the 238Pu-doped pyrochlore crystalline matrix. The inclusions were observed visually using SEM and increased as a function of the cumulative dose. Also, after a cumulative dose of 1.1 × 1025 alpha decays/m3 the pyrochlore became nearly amorphous and inclusions of Hf-Ti-Ca-O phase were found in the crystalline matrix. In all cases separate phases of U-oxide and Hf-Ti-Ca-O were localized in chemically inhomogeneous areas. The results obtained allow us to assume that under self-irradiation the amorphization of Ti-pyrochlore might be accompanied with destruction of single solid solution, (Ca,Gd,Hf,U,Pu)2Ti2O7, into several pyrochlore phases of different chemical composition and precipitation of some amount of Pu and U into separate oxide phases.

Two Bax CsyMnzTi8-zO16 hollandite samples, containing excess rutile, have been prepared, targeting the Mn valence as 3+ by sintering at 1300oC in air (z = 2x +y) and as 2+ (z = x + y/2) in argon. SEM showed the sample sintered in air to contain major hollandite and minor Mn titanate and rutile. The sample sintered in argon contain major hollandite, and minor Ba titanate and rutile. X-ray absorption near edge spectroscopy (XANES), electron energy loss spectroscopy (EELS) and X-ray photoelectron spectroscopy (XPS) were used to determine the oxidation state of the Mn in the samples. Transmission manganese K-edge X-ray absorption spectra of the samples were similar but the precise ratio of oxidation states could not be calculated from these spectra.

The electrical conductance or conductivity of three salt mixtures, Na-K-Cl-NO3, Ca-K-Cl and Ca-Na-Cl, were measured at 25, 50 and 70°C [77, 122, and 158 °F] as a function of relative humidity (RH). Mutual deliquescence and efflorescence RH (MDRH and MERH) values were determined based on the conductivity measurements. It was found that the conductivity of the three salt mixtures started to increase at RH values that are approximately 40 % of their MDRH and increased by 1to 2 orders of magnitude just before reaching the MDRH. At the MDRH, a significant increase in conductivity was observed. The MDRH and MERH for the Ca-K-Cl and Ca-Na-Cl mixtures were found to be approximately 15 % in the temperature range of 50 to 70 °C [122 to 158 °F]. The MDRH and MERH for the Na-K-Cl-NO3system were found to be approximately 54 % at 50 °C [122 °F] and decreased significantly with an increase in temperature.

Immobilization of radioactive wastes arising from the pyrochemical reprocessing of Pu metal with calcium phosphate as the host material has been investigated. Initial studies in which non-active elements are substituted for the actinide elements present in the wastes demonstrated that effective immobilization of both the cations and chloride ions was achieved within the chlorapatite Ca5(PO4)3Cl, and spodiosite, Ca2(PO4)Cl, mineral phases produced during the solid state reaction between host and waste. Follow-up studies with radioactive simulated waste composed of CaCl2, PuCl3 and AmCl3have confirmed the viability of the process and the durability of the product in aqueous solution. More recently a study of the long-term radiological stability and chemical durability of the ceramic product has commenced in which accelerated ageing of the ceramic phases is being achieved through the replacement of the 239Pu (half life 24,360y) that is contained in the actual waste by the short lived isotope 238Pu (half life 86.4y). The preparation of non-active and active samples and the crystal phases formed in each system as determined with XRD are herein reported. Data are presented which show that there is no detectable change in the XRD pattern of the samples containing 238Pu after 532 days of ageing, corresponding to around 400 years of real time ageing employing 239Pu. The detailed effects of ageing on the crystal structure and on the chemical durability of samples in aqueous solution are reported.

The corrosion of the Fe-based spent fuel containing canister produces large amounts of hydrogen, which dominate the composition of the gas phase. To quantify to what extent the hydrogen over- pressure may counteract radiolysis enhanced matrix dissolution, related experimental work has been performed. High burnup spent fuel was corroded in 5 M NaCl solution in different tests: (1) Fe ab-sence (<0.16 bar H2 overpressure), (2) Fe presence (2.8 bar H2 overpressure), (3) Fe absence, external applied H2 overpressure of 3.2 bar. In the absence of Fe, after the application of H2 overpressure the UO2 matrix dissolution rate decreased by a factor of about 10. The concentrations of U, Np, Tc in so- lution were found to be decreasing at least two orders, and ranging within the same level as in the presence of Fe powder. However, Pu and Am concentrations were less affected, due to the absence of Fe powder and the associated high sorption capacity for these radioelements.

The evaporation of a range of synthetic pore water solutions representative of the potential high-level-nuclear-waste repository at Yucca Mountain, NV is being investigated. The motivation of this work is to understand and predict the range of brine compositions that may contact the wastecontainers from evaporation of pore waters, because these brines could form corrosive thin films on the containers and impact their long-term integrity. A relatively complex synthetic Topopah Spring Tuff pore water was progressively concentrated by evaporation in a closed vessel, heated to 95°C in a series of sequential experiments. Periodic samples of the evaporating solution were taken to determine the evolving water chemistry. According to chemical divide theory at 25°C and 95°C our starting solution should evolve towards a high pH carbonate brine. Results at 95°C show that this solution evolves towardsa complex brinethat contains about 99 mol% Na+for the cations, and 71 mol% Cl-, 18 mol% ΣCO2(aq), 9 mol% SO42- for the anions. Initial modeling ofthe evaporating solution indicates precipitation of aragonite, halite, silica, sulfate and fluoride phases. The experiments have been used to benchmark the use of the EQ3/6 geochemical code in predicting the evolution of carbonate-rich brines during evaporation.

A probabilistic model to predict the lifetimes of the engineered barrier system proposed for the Yucca Mountain repository is described. The model assumes that the titanium Grade-7 drip shield will fail by hydrogen-induced cracking and the Alloy-22 waste package by a combination of passive and crevice corrosion. The model predicts that crevice corrosion of the waste package can be completely avoided ifthe drip shield deflects seepage drips for between 2000 (realistic behaviour) and 6000 years (conservative behaviour). Sensitivity calculations on the crevice corrosion model suggest that early waste package failure is extremely unlikely providing the drip shield performs its function for a minimum of ~ 300 years.

Systematic studies of a series of RE-titanate pyrochlore single crystals determined the formation enthalpies, ΔH0f-ox, and the critical amorphization temperature, Tc, the temperature above which the crystal can no longer be amorphized. A negative linear correlation was observed between ΔH0f-ox and Tc. In general as the cation radius ratio, RA/RB, decreases the formation enthalpy becomes less exothermic, i.e. the pyrochlore structure becomes less thermodynamically stable while the susceptibilty to ion beam-induced amorphization decreases. The opposite relation, i.e. a positive linear correlation between ΔH0f-ox and Tc was observed for a series of REPO4 single crystals with both the monazite and xenotime structures. Both relations can be understood in terms of pyrochlore and orthophosphate crystal chemistry, their formation of point defects, and the role of ion-irradiation induced phase transformations.

The oxidation of UO2–type spent nuclear fuel (SNF) in gaseousmedia was studied at different temperatures and oxygen contents using gravimetric and powder X-ray diffraction (XRD) techniques. The aim of the study was to determine the mechanism(s) of thermal-oxidation alteration of SNF during long-term dry storage. The samples used in the experiments were chips of RBMK-1000 fuel rods.Oxidation of UO2with a mean burn-up of 10.7 and 19.73 MW d/kg in humid air was observed at a temperature as low as 150°C. At 200°C nearly all of the UO2was transformed into U3O8 between 3500-4000 hours. In a humid nitrogen environment containing of 0.05-1.3 vol. % oxygen at 300°C, the UO2 completely transformed to U3O8 between 2500-3000 hours. Oxidation of UO2in samples with small amounts of jacket damage (e.g., <0.04 MM2)ll progresses more slowly and after â3000 hours the oxygen-to-uranium ratio was 2.56.Stabilization of the oxidation process was not observed in the fuel samples upto an O/U ratio of 2.4, which may be attributed to the smallburn-up of the fuel under investigation.