Studies on weldability of Ti-5Ta-1.8Nb alloy

The welding, qualification and characterization of welds of Ti-5Ta-1.8Nb alloy, which is being developed for high corrosion resistant performance, are reported. Based on the studies performed as per the ASME Section IX standards, welding procedure specification and procedure qualification record have been formulated. The heterogeneous microstructures of the weldment are rationalized, based on phase transformation in the alloy system and differences in the thermal cycles of various microscopic regions.     

Behavior of molybdenum in pyrochemical reprocessing: A spectroscopic study of the chlorination of molybdenum and its oxides in chloride melts

The high temperature reactions of molybdenum and its oxides with chlorine and hydrogen chloride in molten alkali metal chlorides were investigated between 400 and 700 °C. The melts studied were LiCl-KCl, NaCl-CsCl and NaCl-KCl and the reactions were followed by in situ electronic absorption spectroscopy measurements. In these melts Mo reacts with Cl₂ and initially produces MoCl₆²⁻ and then a mixture of Mo(III) and Mo(V) chlorocomplexes, the final proportion depending on the reaction conditions. The Mo(V) content can be removed as MoCl₅ from the melt under vacuum or be reduced to Mo(III) by Mo metal. The reaction of Mo when HCl gas is bubbled into alkali chloride melts yields only MoCl₆³⁻. MoO₂ reacts in these melts with chlorine to form soluble MoOCl₅²⁻ and volatile MoO₂Cl₂. MoO₃ is soluble in chloride melts and then decomposes into the oxychloride MoO₂Cl₂, which sublimes or can be sparged from the melt, and molybdate. Pyrochemical reprocessing can thus be employed for molybdenum since, after various intermediates, the end-products are chloride melts containing chloro and oxychloro anions of molybdenum plus molybdate, and volatile chlorides and oxychlorides that can be readily separated off. The reactions were fastest in the NaCl-KCl melt. The X-ray diffraction pattern of MoO₂Cl₂ is reported for the first time.     

A creep rupture criterion for Zircaloy-4 fuel cladding under internal pressure

The usual criterion which limits the cladding strain to 0.01 to prevent the creep rupture under internal pressure seems too conservative for application to transport and interim storage. So we have analysed CEA’s data on this subject for CWSR Zircaloy-4 in order to find a less conservative criterion. Temperatures between 350 and 470 °C were studied for stresses between 100 and 550 MPa, according to the irradiation level from 0 to 9.5 × 10²⁵ n m⁻². Except for high stressed irradiated material (because of low ductility), the plastic instability appears as the major mechanism of rupture. For the unirradiated material, it is essentially due to the stress increase with strain. This instability is accelerated by annealing for the irradiated one at moderate or low stress. From these considerations, we propose a new rupture criterion for CWSR Zircaloy-4 cladding submitted to internal pressure, for both unirradiated and irradiated materials.     

Separation of uranium from (U, Th)O2 and (U, Pu)O2 by solid state reactions route

Solid state reactions of UO₂, ThO₂, PuO₂ and their mixed oxides (U, Th)O₂ and (U, Pu)O₂ were carried out with sodium nitrate upto 900 °C, to study the formation of various phases at different temperatures, which are amenable for easy dissolution and separation of the actinide elements in dilute acid. Products formed by reacting unsintered as well as sintered UO₂ with NaNO₃ above 500 °C were readily soluble in 2 M HNO₃, whereas ThO₂ and PuO₂ did not react with NaNO₃ to form any soluble products. Thus reactions of mixed oxides (U, Th)O₂ and (U, Pu)O₂ with NaNO₃ were carried out to study the quantitative separation of U from (U, Th)O₂ and (U, Pu)O₂. X-ray diffraction, X-ray fluorescence, thermal analysis and chemical analysis techniques were used for the characterization of the products formed during the reactions.     

Structure of zirconium alloy oxides formed in pure water studied with synchrotron radiation and optical microscopy: relation to corrosion rate

A detailed study was undertaken of oxides formed in 360 °C water on four Zr-based alloys (Zircaloy-4, ZIRLO™,¹ Zr-2.5%Nb and Zr-2.5%Nb-0.5%Cu) in an effort to relate oxide structure to corrosion performance. Micro-beam X-ray diffraction was used along with transmitted light optical microscopy to obtain information about the structure of these oxides as a function of distance from the oxide-metal interface. Optical microscopy revealed a layered oxide structure in which the average layer thickness was inversely proportional to the post-transition corrosion rate. The detailed diffraction studies showed an oxide that contained both tetragonal and monoclinic ZrO₂, with a higher fraction of tetragonal oxide near the oxide-metal interface, in a region roughly corresponding to one oxide layer. Evidence was seen also of a cyclic variation of the tetragonal and monoclinic oxide across the oxide thickness with a period of the layer thickness. The results also indicate that the final grain size of the tetragonal phase is smaller than that of the monoclinic phase and the monoclinic grain size is smaller in Zircaloy-4 and ZIRLO than in the other two alloys. These results are discussed in terms of a model of oxide growth based on the periodic breakdown and reconstitution of a protective layer.     

Investigation of electrorefining of metallic alloy fuel onto solid Al cathodes

This work concerned the electrorefining of UZr and UPuZr alloys on a solid aluminium cathode, in the LiCl-KCl eutectic melt containing U³⁺, Pu³⁺, Np³⁺, Zr²⁺ or Zr⁴⁺, Am³⁺, Nd³⁺, Y³⁺, Ce³⁺ and Gd³⁺ chlorides. During constant current electrolyses, the use of a cathodic cut-off potential (−1.25 V versus Ag/AgCl) allowed to selectively deposit actinides (mainly U), while lanthanides remained in the salt. The aim was to determine the maximal load achievable on a single aluminium electrode. The total exchange charge was 4300 C, which represents the deposition of 3.72 g of actinides in 4.17 g Al, yielding a composition of 44.6 wt% An in Al. It was shown that the melting of the cathode contributed to increase the total amount of actinides deposited on the aluminium.     

Fuel oxidation and thermal conductivity model for operating defective fuel rods

A model has been developed to describe the fuel oxidation behaviour, and its influence on the fuel thermal conductivity, in operating defective nuclear fuel rods. The fuel-oxidation model is derived from adsorption theory and considers the influence of the high-pressure environment that results from coolant entry into the fuel-to-clad gap. This model is in agreement with the fuel-oxidation kinetics observed in high-temperature annealing experiments conducted at 1473-1623 K in steam over a range of pressure from 0.001 to 0.1 MPa. Using a Freundlich adsorption isotherm, the current model is also consistent with recent experiments conducted at a higher pressure of 7 MPa. The model also considers radiolytic effects as a consequence of fission fragment bombardment in the fuel-to-clad gap. This treatment suggests that radiolysis-assisted oxidation is insignificant in operating defective rods (as compared to thermal effects), as supported by limited in-reactor data. The effects of diffusion of the interstitial oxygen ions in the solid in the operating rod is further discussed.     

Spent fuel radionuclide source term model for assessing spent fuel performance in geological disposal. Part II: Matrix alteration model and global performance

In the framework of the research conducted on the long term evolution of spent nuclear fuel under geological disposal conditions, a source term model has been developed to evaluate the instantaneous release of radionuclides (RN) (instant release fraction, IRF) and the delayed release of the RN which are embedded within the matrix. This model takes into account most of the scientific results currently available except the effect of hydrogen and the current knowledge of the uncertainties. IRF was assessed by considering the evolution with time of the RN inventories located within the fuel microstructure to which no confinement properties can be allocated over the long term (gap, rim, grain boundaries). This allows for bounding values for the IRF as a function of time of canister breach and burnup. The matrix radiolytic dissolution was modeled by a simple kinetic model neglecting the recombination of radiolytic species and the influence of aqueous ligands. The oxidation of the UO₂ matrix was assumed not to be kinetically controlled. Spent fuel performance was therefore demonstrated to mainly depend on the reactive surface area.     

Measurement of the specific heat of Zr–40 wt%U metallic fuel

The specific heat capacities of un-irradiated and irradiated metallic Zr–40 wt%U fuel have been measured between 50 °C and 1000 °C with a differential scanning calorimetry. The irradiated fuels have three different burnup levels of 0.38, 0.70 and 0.92 g-fission product (FP)/cm³. The measured specific heat for the un-irradiated fuel is representative and consistent with the values estimated from the Neumann–Kopp rule. The irradiated fuels exhibited a complicated behavior of the heat capacities. The unique characteristics of the specific heat capacities can be explained by the recovery of radiation damage, the formation of fission gas bubbles and fission gas release, and a phase transition in the irradiated fuels. An examination of the microstructure revealed that multiple large bubbles were formed in the irradiated fuel during specific heat measurement. The measured specific heat is expected to enable us to estimate the stored energy in the metallic fuel during certain accident scenarios and to determine the thermal conductivity of zirconium–uranium metallic fuel.     

Radionuclide release from high burnup spent fuel during corrosion in salt brine in the presence of hydrogen overpressure

In the case of a contact between groundwater and Fe-based spent fuel disposal containers in a repository large amounts of hydrogen will be produced by the corrosion of iron, which may result in significant hydrogen pressures. To quantify to what extent the hydrogen overpressure may counteract radiolysis enhanced matrix dissolution, related experimental work has been performed. High burnup spent fuel was corroded in 5.6 mol (kg H₂O)⁻¹ NaCl solution applying H₂ overpressures (experimental set 1) <0.17 bar by radiolysis, (experimental set 2) 2.8 bar by Fe corrosion, (experimental set 3) 3.2 bar by external H₂ gas. In the absence of Fe (experimental set 3) the UO₂ matrix dissolution rate decreased by a factor of about 10. In this test the concentrations of U, Np, Tc in solution were found to be decreasing by at least two orders of magnitude, and ranging within the same level as in the presence of Fe powder (experimental set 2). However, Pu and Am concentrations (experimental set 3) were less affected, due to the high sorption capacity for these radioelements onto Fe corrosion products.     

On stability of spatial distributions of crystal structure defects in irradiated high burnup UO2 fuel

Conditions of Kinoshita instability development of point defects and dislocation spatial distributions in the crystal structure of UO₂ fuel are studied. As a result of the instability development, spatially non-uniform regions with increased dislocation density are formed. Closed-form expressions of instability increment and spatial scale are derived. Parameters of the instability for irradiation conditions of high burnup UO₂ fuel are obtained by means of numerical simulation. Instability development time is shown to be inversely proportional to fission rate and it increases as dislocation density decreases. Calculated values of instability spatial scale and increment are in accordance with the size of fine grains and their formation rate in the peripheral zones of high burnup LWR fuel pellets.     

Estimation of maximum coated particle fuel compact packing fraction

This work develops an analytic fuel fraction packing model for a high temperature gas cooled reactor fuel compact fabricated from overcoated particles of a single size. The model includes the effects of one dimensional compression and finite matrix grain size. One dimensional compression limits the maximum fuel packing fraction to about 48% for the pressed compact in this single sized particle system. This limit is due to two effects. The first is that the process of die loading limits the pre-compression packing configuration to one that is stable under gravity, which is not the most space efficient one. The second effect is due to the one dimensional compression which reduces only the axial dimension of the particle lattice rather than uniformly compressing the lattice. The die wall can also limit the maximum packing fraction by preventing the nearby particles from moving into a more space efficient configuration.     

Statistical approach and benchmarking for modeling of multi-dimensional behavior in TRISO-coated fuel particles

The fundamental design for a gas-cooled reactor relies on the behavior of the coated particle fuel. The coating layers, termed the TRISO coating, act as a mini-pressure vessel that retains fission products. Results of US irradiation experiments show that many more fuel particles have failed than can be attributed to one-dimensional pressure vessel failures alone. Post-irradiation examinations indicate that multi-dimensional effects, such as the presence of irradiation-induced shrinkage cracks in the inner pyrolytic carbon layer, contribute to these failures. To address these effects, the methods of prior one-dimensional models are expanded to capture the stress intensification associated with multi-dimensional behavior. An approximation of the stress levels enables the treatment of statistical variations in numerous design parameters and Monte Carlo sampling over a large number of particles. The approach is shown to make reasonable predictions when used to calculate failure probabilities for irradiation experiments of the New Production - Modular High Temperature Gas Cooled Reactor Program.     

Behavior of thorium-uranium (IV) phosphate-diphosphate sintered samples during leaching tests. Part I - Kinetic study

The dissolution of β-TUPD sintered samples was examined in various conditions of pH, temperature, concentrations of anions in the leachate and leaching flow rates. All the normalized dissolution rates were in the range 10⁻⁷ to 10⁻⁴ g m⁻² day⁻¹ even in very aggressive media, showing the good resistance of these ceramics to aqueous alteration. The first part of this paper describes several parameters exhibiting a significant influence on the normalized dissolution rate of the pellets prepared. Both the partial order relative to the proton concentration (n = 0.39-0.41) and the apparent activation energy (E_app = 49 kJ mol⁻¹) were found in good agreement with the data reported for powdered samples showing that the sintering process does not degrade the chemical durability of the ceramics. Moreover, due to the high thermodynamical constant of complexation of phosphate species for tetravalent uranium and thorium, the influence of other ligands such as nitrate, chloride or sulphate on the normalized dissolution rates was limited. Near the equilibrium, the increasing of the leaching time, the temperature or the leachate acidity led to the thorium precipitation at the surface of the pellets either in static or in dynamic conditions. Consequently, the dissolution became clearly incongruent and controlled by saturation processes which are described in the second part of this paper.     

Influence of ageing heat treatment on alloy A-286 microstructure and stress corrosion cracking behaviour in PWR primary water

The influence of ageing heat treatment on alloy A-286 microstructure and stress corrosion cracking behaviour in simulated Pressurized Water Reactor (PWR) primary water has been investigated. A-286 microstructure was characterized by transmission electron microscopy for ageing heat treatments at 670 °C and 720 °C for durations ranging from 5 h to 100 h. Spherical γ′ phase with mean diameters ranging from 4.6 to 9.6 nm and densities ranging from 8.5 × 10²² m⁻³ to 2 × 10²³ m⁻³ were measured. Results suggest that both the γ′ phase mean diameter and density quickly saturate with time for ageing heat treatment at 720 °C while the γ′ mean diameter increases significantly up to 100 h for ageing heat treatment at 670 °C. Grain boundary η phase precipitates were systematically observed for ageing heat treatment at 720 °C even for short ageing periods. In contrast, no grain boundary η phase precipitates were observed for ageing heat treatments at 670 °C except after 100 h. Hardening by γ′ precipitation was well described by the dispersed barrier hardening model with a γ′ barrier strength of 0.23. Stress corrosion cracking behaviour of A-286 was investigated by means of constant elongation rate tensile tests at 1.5 × 10⁻⁷ s⁻¹ in simulated PWR primary water at 320 °C and 360 °C. In all cases, initiation was transgranular while propagation was intergranular. Grain boundary η phase precipitates were found to have no significant effect on stress corrosion cracking. In contrast, yield strength and to a lesser extent temperature were found to have significant influences on A-286 susceptibility to stress corrosion cracking.     

Behavior of thorium-uranium (IV) phosphate-diphosphate sintered samples during leaching tests. Part II. Saturation processes

Sintered pellets of thorium-uranium (IV) phosphate-diphosphate solid solutions (β-Th_4−xU_x(PO₄)₄P₂O₇, β-TUPD) were altered in several acidic media. All the results reported in the first part of this paper confirmed the good chemical durability of the samples. The evolution of the normalized weight loss showed that, in several media, thorium quickly precipitates in a neoformed phosphate-based phase while uranium (IV) is released in the leachate due to its oxidation into the uranyl form. The characterization of neoformed phases was carried out through several techniques involving grazing XRD, infrared and μ-Raman spectroscopies, EPMA, SEM and TEM. SEM micrographies showed that the dissolution mainly occurs at the grain boundaries, leading to the break away of the grains: only the first 15 μm are altered for 2 months in 10⁻¹ M HNO₃. From EPMA and BET measurements, neither the chemical composition nor the specific surface area are significantly modified. Near equilibrium, two neoformed phases were observed and identified by grazing XRD and/or μ-Raman spectroscopy at the surface of the leached pellets: one is found to be amorphous and progressively turns into the crystallized thorium phosphate-hydrogenphosphate hydrate (TPHPH). From the results obtained, a chemical scheme of the dissolution of β-TUPD sintered samples is proposed. The behavior of the actinides in the gelatinous phase appears mainly driven by their oxidation state: thorium remains in the tetrapositive state and is quickly and quantitatively precipitated while uranium (IV) is oxidized into uranyl then released in the leachate. The Th-precipitation as TPHPH first appears scattered then covers the entire surface of the pellet, inducing a delay of the actinides release in the leachate. Both phases act as protective layers and should induce the significant delay of the release of actinides (Th, U) to the biosphere.     

Development of a fundamental crack tip strain rate equation and its application to quantitative prediction of stress corrosion cracking of stainless steels in high temperature oxygenated water

A formulation for the quantitative calculation of the stress corrosion cracking (SCC) growth rate was proposed based on a fundamental-based crack tip strain rate (CTSR) equation that was derived from the time-based mathematical derivation of a continuum mechanics equation. The CTSR equation includes an uncertain parameter r₀, the characteristic distance away from a growing crack tip, at which a representative strain rate should be defined. In this research, slow strain rate tensile tests on sensitized 304L stainless steel in oxygenated high temperature water were performed. By curve fitting the experimental results to the numerically calculated crack growth rate, the parameter r₀ was determined. Then, the theoretical formulation was used to predict the SCC growth rates. The results indicate that r₀ is on the order of several micrometers, and that the application of the theoretical equation in predicting the crack growth rate provides satisfactory agreement with the available data.     

Threats, design limits and design windows for laser IFE dry wall chambers

The High Average Power Laser (HAPL) program is carrying out a coordinated effort to develop inertial fusion energy based on lasers, direct-drive targets and a dry wall chamber. The dry wall must accommodate the ion and photon threat spectra from the fusion micro-explosion over its required lifetime. This paper summarizes the current HAPL strategy on the armor/first wall configuration based on tungsten and ferritic steel as preferred armor and structural materials, respectively. The thermal performance of an example fully dense tungsten armor configuration on a ferritic steel first wall is described showing the basis for separating the high energy accommodation function of the armor from the structural function of the first wall. Example design operating windows for the armor, first wall and blanket are presented based on different requirements and constraints. The possibility of utilizing an engineered porous armor is discussed. Key chamber wall and armor issues are summarized.     

A review of refractory metal alloys and mechanically alloyed-oxide dispersion strengthened steels for space nuclear power systems

Mechanical and thermo-physical properties of refractory metal alloys and mechanically alloyed (MA)-oxide dispersion strengthened (ODS) steels are reviewed and their potential for use in space nuclear reactors is examined. Preferable refractory alloys for use in liquid metal and gas-cooled space reactors include Nb-1%Zr, PWC-11, Mo-TZM, Mo-xRe where x varies from 7% to 44.5%, T-111 and ASTAR-811C. These alloys are heavy, difficult to fabricate, and are not readily available. The advantages of the MA-ODS alloys are: (a) their strength at high temperatures (>1000 K), which decreases slower with temperature than those of niobium and molybdenum alloys; (b) relatively lightweight and less expensive; (c) low swelling and no embrittlement with exposure to high-energy neutrons (>0.1 MeV) up to 10²⁷ n/m²; and (d) high resistance to oxidation and nitration. The few data available on compatibility of MA-ODS alloys with alkali liquid metals up to 1100 K are encouraging, however, additional tests at typical temperatures (1000-1400 K) in space nuclear reactors are needed. The anisotropy of MA-ODS alloys when cold worked, and particularly rolled into tubes, should not hinder their use in space nuclear power systems, in which operation pressure is either near atmospheric or as high as 2 MPa, but joints weldability is an issue.     

Determining the shear properties of the PyC/SiC interface for a model TRISO fuel

The fracture behavior of TRISO-coated fuel particles is dependent on the shear strength of the interface between the inner pyrolytic carbon (PyC) and silicon carbide coatings. This study evaluates the interfacial shear properties and the crack extension mechanism for TRISO-coated model tubes using a push-out technique. The interfacial debond shear strength was found to increase with increasing sample thickness and finally approached a constant value. The intrinsic interfacial debond shear strength of ∼280 MPa was estimated. After the layer is debonded, the applied load is primarily transferred by interfacial friction. A non-linear shear-lag model predicts that the residual clamping stress at the interface is ∼350 MPa, and the coefficient of friction is ∼0.23, yielding a frictional stress of ∼80 MPa. These relatively high values are attributed to the interfacial roughness. Of importance in these findings is that this unusually high interfacial strength could allow significant loads to be transferred between the inner PyC and SiC in application, potentially leading to failure of the SiC layer.