The Yucca Mountain Project (YMP) 1998 Total System Performance Assessment Viability Assessment (TSPA-VA) analyzed the degradation of Zircaloy clad commercial fuel rods and the resulting exposure of the fuel in the event of a waste package failure. The cladding degradation mechanisms considered were damage before emplacement, mechanical failure from drift collapse, localized corrosion, general corrosion, delayed hydride cracking (DHC), hydride reorientation, creep rupture, and stress corrosion cracking (SCC). The potential for further cladding degradation due to cladding rupture as a result of fuel oxidation was also considered in the modeling effort. These models have been improved for use in future TSPAs.

The current cladding degradation model divides the analysis into two phases, cladding failure (perforation) and cladding unzipping (crack propagation caused by the expansion of UO2 fuel after reaction with water). Cladding failure occurs during reactor operation, from creep strain failure during high temperature periods in dry storage or in the early periods in the repository, or localized corrosion. After a Waste Package (WP) containing spent nuclear fuel in the repository fails, moisture is assumed to enter the waste package and the failed cladding starts to unzip (tear open) from the formation of secondary uranium phases. This slowly exposes the fuel. In addition, the inventory of fission products located in the gap between the cladding and fuel pellet is rapidly released. The cladding model limits the amount of fuel that is exposed to moisture and becomes available for dissolution. As a result, the doses to the affected population are reduced (factor of 20 to 50 in TSPA-VA) from the case where cladding is not considered.

In the framework of the 1991's French Law which defines the 3 major research lines to manage of nuclear wastes, CEA (Commissariat à l'Energie Atomique) is supporting a wide R&D program dealing with the long term behavior of spent fuel in various boundary conditions representative of interim storage and geological disposal. One major issue concerns the potential evolution with time of the thermo-mechanical properties of the irradiated cladding, the question of its integrity and whether the cladding can be considered as a first confinement barrier for radionuclides. The answer will strongly influence the design and the safety analyses of the interim storage and potentially the first stage of a geological disposal.

After irradiation, the thermo-mechanical properties of the spent fuel cladding are altered compared to those of the non irradiated material: presence of numerous irradiation defects due to the high irradiation, presence of external zirconia layers and hydrogen within the Zircaloy due to the external corrosion. Furthermore, the cladding is submitted to a relative high internal pressure field which is related to the production and release in the free volumes of fission gases and helium. Since the cladding is expected to undergo a relative high temperature field (∼300 - 400°C), long term creep is expected to become a relevant deformation mechanism which can potentially lead to a breaching of the cladding.In order to deal with this strategic issue, experimentation on irradiated samples as well as modelling work are in progress in CEA with the support of EDF and FRAMATOME. The ambitious objectives are to define and qualify the long term mechanical properties of the cladding, in particular a long term creep law and an adapted breaching criterion. Long term creep properties are studied through a stepwise approach from short term (few days) and high stress field experiments to long term (few years) and low stress field experiments. The validation of the extrapolation is ensured by complete metallurgical characterizations (HRTEM, XRD, H2 content) performed before and after deformation. The first step of this work, medium term (∼ 1 month) creep experiments on irradiated cladding will be presented.

This report summarizes results of an analysis of hydride-related degradation of commercial spent-nuclear-fuel cladding under repository conditions. Based on applicable laboratory data on critical stress intensity obtained under isothermal conditions, occurrence of delayed hydride cracking from the inner-diameter side of cladding is concluded to be extremely unlikely. The key process for potential initiation of delayed hydride cracking at the outer-diameter side is long-term microstructural evolution near the localized regions of concentrated hydrides, i.e., nucleation, growth, and cracking of hydride blisters. Such locally concentrated hydrides are, however, limited to some high-burnup cladding only, and the potential for crack initiation at the outer-diameter side is expected to be insignificant for the majority of spent fuels. Some degree of hydride reorientation could occur in highburnup spent-fuel cladding. However, even if hydride reorientation occurs, accompanying stress-rupture failure in spent fuel cladding is unlikely to occur.

Safe interim dry storage of spent nuclear fuel (SNF) must be maintained for a minimum of twenty years according to the Code of Federal Regulations. The most important variable that must be regulated by dry storage licensees in order to meet current safety standards is the temperature of the SNF. The two currently accepted models for defining the maximum allowable initial storage temperature for SNF are based on the diffusion controlled cavity growth (DCCG) failure mechanism proposed by Raj and Ashby. These models may not give conservative temperature limits. Some have suggested using a strain-based failure model to predict the maximum allowable temperatures, but we have shown that this is not applicable to SNF as long as DCCG is the assumed failure mechanism. Although the two accepted models are based on the same fundamental failure theory (DCCG), the researchers who developed the models made different assumptions, including selection of some of the most critical variables in the DCCG failure equation. These inconsistencies are discussed together with recommended modifications to the failure models based on more recent data.

When spent fuel eventually comes into contact with groundwater, fuel matrix dissolution will be strongly influenced by redox conditions in the near field. The most significant factors influencing redox conditions at the fuel surface in a reducing environment are alpha-radiolysis of water and the presence of reductants such as Fe(II) and H2 arising principally from canister corrosion. The radiolytic yield (G) of molecular oxidants, generally considered to be ∼1 molecule of hydrogen peroxide per 100 eV for alpha-radiolysis, is expected to be considerably lower in the presence of reductants but the overall effect on the rate of matrix dissolution cannot yet be reliably quantified. We have attempted to estimate the effective yield of oxidants by examining the results from various studies of spent fuel and UO2 dissolution, including alpharadiolysis experiments. The analysis suggests that the effective yield (Geff) is likely to be no greater than 0.01 in the repository environment. The implications of low Geff values are discussed in relation to fuel dissolution rates. Some other aspects of radiolysis relevant to nearfield redox chemistry are also examined, including the potential significance of alpha emitters sorbed in the repository near field in producing radiolytic oxidants.

This work uses an external alpha beam to irradiate an uranium oxide/water interface and investigates the release of uranium in aerated deionized water under alpha irradiation. A high energy alpha beam delivered by a cyclotron (CERI-CNRS) goes through the oxide and emerges in the water with a 20 MeV energy. First results are reported here showing that the uranium mass loss rate increases by three orders of magnitude in aerated deionized water under high flux (≥3.3×1010 α.cm−2s−1).

The UO2 in spent nuclear fuel is unstable under moist oxidizing conditions and will be altered to uranyl oxide hydrate phases. The transuranics released during the corrosion of spent fuel may also be incorporated into the structures of secondary U6+ phases. The incorporation of radionuclides into alteration products will affect their mobility. A series of precipitation tests were conducted at either 150 or 90°C for seven days to determine the potential incorporation of Ce4+ and Nd3+ (surrogates for Pu4+ and Am3+, respectively) into uranium phases. lanthinite ([U24+(UO2)4O6(OH)4(H2O)4](H2O)5) was produced by dissolying uranium oxyacetate in a solution containing copper acetate monohydrate as a reductant. The leachant used in these tests were doped with either 2.1 ppm cerium or 399 ppm neodymium. Inductively coupled plasmamass spectrometer (ICP-MS) analysis of the solid phase reaction products which were dissolved in a HNO3 solution indicates that about 306 ppm Ce (Kd = 1020) was incorporated into ianthinite, while neodymium contents were much higher, being approximately 24,800 ppm (Kd 115). Solid phase examinations using an analytical transmission electron microscope/electron energy-loss spectrometer (AEM/EELS) indicate a uniform distribution of Nd, while Ce contents were below detection. Becquerelite (Ca[(UO2)6O4(OH)6]·8H2O) was produced by dissolving uranium oxyacetate in a solution containing calcium acetate. The leachant in these tests was doped with either 2.1 ppm cerium or 277 ppm neodymium. ICP-MS results indicate that about 33 ppm Ce (Kd = 17) was incorporated into becquerelite, while neodymium contents were higher, being approximately 1,300 ppm (Kd = 5). Homogeneous distribution of Nd in the solid phase was noted during AEM/EELS examination, and Ce contents were also below detection.

In order to get a better understanding for the spent fuel corrosion process, the variations of important intensive parameters such as pH and the redox potential (Eh) of the bulk solution have been continuously measured during long term sequential leaching experiments. These data may be used together with standard chemical analytical data for modeling spent fuel corrosion, especially in anoxic or reducing conditions. In order to overcome difficulties caused by the strong radiation field and the long experiment times, a method for in-situ measurements of pH and Eh using a computer controlled system was worked out. The stability of the measuring system over long time periods was then tested; pH values stable within 0.05 pH units/year in buffered systems were measured. The variations of these parameters in a variety of conditions and solution compositions were followed continuously during the spent fuel leaching process.

A discussion of the results of spent fuel leaching in anoxic conditions is presented, pointing out the difficulties to realize in laboratory near field conditions. An interesting case of calcite co-precipitation/co-dissolution is also presented. Dissolution experiments show that the calcite precipitated previously during spent fuel leaching experiments in synthetic groundwater contained considerable amounts of actinides and fission products.

One of the factors considered within the studies of performance assessment on spent fuel under final repository conditions is the effect of the radiation on its leaching behavior. Radiation from spent fuel can modify some properties of both solid phase and leachant and therefore it would alter the chemical behavior of the near field.

Particularizing in the effect of the radiation on the leachant, it will cause generation of radiolytic species that could change the redox potential of the environment and therefore may bring on variations in the leaching process.

In this work, the chemical analogue utilized was SIMFUEL (natural UO2 doped with non-radioactive elements simulating fission products) and the leachants selected were saline and granite bentonite waters both under initial anoxic conditions. To emulate γ radiation field of a spent fuel, leaching experiments with external 60Co sources in a irradiation facility (Nayade) were performed. Initial dose rate used was 0.014 Gy/s.

Preliminary results indicate that radiation produces an increase of the uranium dissolution rate, being the concentrations measured close to those obtained in oxic atmosphere without radiation field. In addition the solubility solid phases from experimental conditions were calculated, for both granite bentonite water and 5 m NaCl media.

On the other hand, a tentative approach to model the role of γ radiolysis in these SIMFUEL tests has been carried out as well.

Correctly identifying the possible alteration products and accurately predicting their occurrence in a repository-relevant environment are the key for source-term calculations in a repository performance assessment. Uraninite in uranium deposits has long been used as a natural analog to spent fuel in a repository because of their chemical and structural similarity. In this paper, a SEM/AEM investigation has been conducted on a partially alterated uraninite sample from a uranium ore deposit of Shinkolobwe of Congo. The mineral formation sequences were identified: uraninite → uranyl hydrates → uranyl silicates → Ca-uranyl silicates or uraninite → uranyl silicates → Ca-uranyl silicates. Reaction-path calculations were conducted for the oxidative dissolution of spent fuel in a representative Yucca Mountain groundwater. The predicted sequence is in general consistent with the SEM observations. The calculations also show that uranium carbonate minerals are unlikely to become major solubility-controlling mineral phases in a Yucca Mountain environment. Some discrepancies between model predictions and field observations are observed. Those discrepancies may result from poorly constrained thermodynamic data for uranyl silicate minerals.

In 1993 SKB launched its Encapsulation Plant Project. Within this project, SKB has:

a) designed a facility for encapsulation of nuclear fuel

b) laid down the design premises for a canister for disposal of nuclear fuel

c) tested and developed fabrication methods for copper canisters

d) evaluated the long term chemical and mechanical behaviour of the canister

e) made preliminary plans for a factory for the production of copper canister

f) constructed a canister laboratory for full scale testing of the key operations in an encapsulation plant.

The conclusions of this project were that a canister consisting of an outer layer of 50 mm copper over an insert of cast nodular iron would provide sufficient corrosion protection and would have sufficient mechanical strength. This canister can be produced by several methods such as forming from rolled plates, hot extrusion, and pierce and draw. These methods have been tested at full scale and the results from these manufacturing tests will be presented and discussed. The canister will be sealed by electron beam welding in the encapsulation plant and the integrity of the weld will be verified by ultrasonic testing and high-energy radiography. The final development work in this area will be performed in the Canister Laboratory in Oskarshamn, Sweden. The laboratory is now in operation and is equipped with a 100 kW EB-welding unit, a 9 MV digital X-ray unit and a state-of-the-art phased array ultrasonic testing unit.

Localized corrosion of high level nuclear waste containers is considered an important factor that will have a strong influence on the overall performance of the proposed repository at Yucca Mountain, NV. The present candidate container material, Alloy 22 [UNS N06022 (57Ni-22Cr-13.5Mo-3W-3Fe)], is highly resistant to localized corrosion. Assessing the performance of the HLW containers also requires an evaluation of localized corrosion resistance and determination of the passive corrosion rate of Alloy 22 weldments. The localized corrosion resistance of welded Alloy 22 specimens was evaluated by measuring the repassivation potential in chloride containing solutions whereas potentiostatic tests were used to determine the passive corrosion rate of the welded material. The results for the welded material are compared to those for the base metal.

Stress corrosion crack growth in pure copper has been studied with the aim of determining data which may be used in extrapolations to conditions of interest for use of copper as a canister material for long term storage of spent nuclear fuel. The canister should retain its integrity for a time of 100 000 years and the ultimate aim of the project is to show that stress corrosion cracking is not a threat to the canister integrity. The crack growth is studied as a function of applied stress intensity factor and environmental factors such as solute concentration and electrochemical potential. The approach used is to determine crack growth rates under conditions which give low but measurable rates which can then be extrapolated to times and conditions of relevance to canister integrity. The most accurate method of on-line crack growth monitoring, DC potential drop, is unfortunately in conflict with the low resistivity of copper. The results so far indicate that there is a stres.intensity threshold for SCC growth in copper which is not appreciably lower than 30 MPa√m. Additional data are needed to improve the credibility of the threshold value. It is also proposed that crack growth experiments are performed in which the environment is gradually made less aggressive in order to prove that growth stops before environmental conditions of relevance to canister integrity are reached.

Predictions from a mathematical model of the propagation of a corrosion pit in copper are reported. The model uses equilibrium data for solid and aqueous species to calculate local chemical and electrochemical equilibria in small volume elements. Mass transport between elements under local internal equilibrium is calculated using aqueous diffusion coefficients with the constraint of electrical neutrality. Propagation of a corrosion pit is deemed possible when the fraction of the oxidised copper that forms solid corrosion products, at the copper metal, is insufficient to completely cover the underlying metal. The effect of pH and salt concentrations in the bulk water was studied by varying the composition of the water.

Results are presented in the form of E-log [Cτ] diagrams where E is the applied potential and [Cτ] is the total bulk concentration of chloride. The E-log [Cτ] diagram shows two separate areas where pitting is found to be possible. One region at low chloride concentration and high potential and one region at high chloride concentration and low potential. Increased sulphate concentration is found to be detrimental with respect to pitting corrosion, particularly in the high potential region. Increased carbonate concentration is found to be beneficial, particularly in the low potential region. Pitting corrosion of copper can be described as a case of galvanic corrosion where cuprous oxide at a pH similar to that of the bulk is the cathode material for oxygen reduction and copper metal at the local, lower pH in a corrosion pit may behave as the anode.

Changes in the microstructure, mechanical properties and corrosion resistance of C-22 alloy were studied systematically as a function of aging temperature and aging time. Aging was performed in the temperature range 260°C to 800°C for times between 0.5 h and 40,000 h. For aging temperatures of 600°C and higher, precipitation of tetrahedral close packed (TCP) phases in C-22 alloy induce a decrease in its mechanical properties and corrosion resistance in aggressive acidic solutions. At the lower aging temperatures, long range ordering (LRO) was observed, which did not produce changes in the chemical resistance of the alloy. Arrhenius extrapolations of the high temperature data predict that C-22 alloy will be thermally stable when exposed to temperatures in the order of 300°C for times higher than 10,000 years.

In connection with the problems of mid-temperature embrittlement and sulfide corrosion of copper, we perform an ab initio study of intrinsic properties of copper(I) sulfide in the anti-fluorite crystal structure (digenite). The energies of the (111) and (110) non-polar surfaces of Cu2S are calculated using the interface Green's function technique. The (111) surface is found to have the lowest energy, in agreement with the cleavage pattern of digenite mineral. The locally self-consistent Green's function method is used to obtain the formation and interaction energies of native point defects in bulk digenite. The results show that digenite exists as a non-stoichiometric compound Cu2–δS with stable (constitutional) cation vacancies, in agreement with experiment. This natural presence of constitutional cation vacancies combined with the calculated low formation energy of Frenkel defects implies a high cation mobility in Cu2–δS, which is consistent with the superionic behavior of digenite.

The near-field chemistry of the repository for spent nuclear fuel arising from interactions between the groundwater, compacted bentonite clay, canister and the spent fuel was calculated using a three-successive-closed-systems approach. The calculations were performed for fresh granitic and saline groundwaters using the thermodynamic computer codes, HYDRAQL/CE and EQ3/6. The effects of water chemistry inside the canister as well as water radiolysis on fuel dissolution were taken into consideration. The groundwater and the three barriers of the near-field were accounted for by this approach, with particular emphasis given to the pH, Eh and actinide solubilities.

A radionuclide source term model has been developed which simulates the biogeochemical evolution of the Drigg low level waste (LLW) disposal site. The DRINK (DRIgg Near field Kinetic) model provides data regarding radionuclide concentrations in groundwater over a period of 100,000 years, which are used as inputs to safety assessment calculations. The DRINK model considers the coupled interaction of the effects of fluid flow, microbiology, corrosion, chemical reaction, sorption and radioactive decay. The model simulates the development of a period of reducing conditions resulting from degradation of cellulose and steel wastes. Under these conditions U and Th remain as solubility controlling solids for periods over 30,000 years and provide an important source of daughter nuclides such as Ra. The fraction of 14C is followed through all reactions involving carbon. Less than 5% of 14C is present as mobile aqueous species.

The Nuclear Regulatory Commission (NRC) Total-system Performance Assessment (TPA) code is a tool to independently evaluate the long-term performance of the proposed Yucca Mountain (YM) high-level radioactive waste (HLW) repository by modeling processes such as the dissolution of the spentnuclear fuel (SNF) and the subsequent transport through unsaturated and saturated highly heterogeneous fractured porous media to a hypothetical release into the biosphere. The release rates of radionuclides into the biosphere corresponding to some of the SNF dissolution models show a sinusoidal trend with an overall decrease in the rate with time; for other models the sinusoidal behavior is nonexistent. This study identifies two key mechanisms contributing to these trends: the spatial discretization of the repository into subareas and radionuclide-specific sorption properties in the saturated zone alluvium. Other mechanisms which might affect the release rates, such as solubility limits, radioactive decay and ingrowth, inventory depletion, and transport properties of the unsaturated zone (UZ), do not significantly contribute to these trends.

Bore-hole repositories (BHR) are considered to be promising for disposal of HLW and spent sealed radiation sources (SRS). A safety assessment of BHR disposal of SRS was performed using geologic environmental analysis, available parameters for BHR and SRS design and radionuclide inventory. The probabilistic calculations take into account some data uncertainties and variability. The results showed that there is practically no release of short-lived radionuclides into the environment for about 1000 years. This is completely due to the very low corrosion rate of the lead matrix in which the SRS are encapsulated. Various models were applied for more detailed numeric simulation of the repository temperature, radiation fields, and transport of released radionuclides in the geosphere. Ultra-conservative scenarios were chosen for these models. The worst case comprises both breaching of all engineered barriers and flooding of the disposal site plus eventual failure of an imperfect SRS immobilization matrix with some sources partly exposed by the breached matrix. For this extreme case, the maximum dose was found to be not higher than 55 – 75 µSv/y.