Creep properties of oxide and carbide fuels under irradiation

The creep behaviour of uranium dioxide and uranium carbide has been examined in both bend and compression experiments in DIDO Materials Test Reactor. In UO₂ no significant variation in creep rate with dose and temperature occured above ~10²⁵ fissions m^?3 between 450°C and 1230°C, the high strain rates measured in compression at low doses being largely attributable to pore sintering. Both a linear rating and stress dependence were observed up to 40 MNm^?2 and creep rates were found to be independent of grain size. At higher doses ($\text{>6 × 10}{}^{26}\text{fissions m}{}^{\mathrm{?3}}$) transient strains were incurred on varying stress and temperature due to the development of grain boundary gas bubbles. This also resulted in a six fold increase in the radiation creep constant between $\text{6 × 10}{}^{26}$ and $\text{1.2 × 10}{}^{27}$ fissions m^?3. A similar pattern of behaviour with respect to rating and stress was observed in hyperstoichiometric UC between 450 and 800°C up to $\text{1 × 10}{}^{27}$ fissions m^?3. However the nominally steady state creep rate was a factor 8 lower than in UO₂ irradiated under the same conditions. The experimental results also suggest that the primary creep contribution to the initial strain in compression is much higher than in UO₂. There was no evidence of either transient strain on changing stress or of an increasing creep rate at high doses. The experimental observations are reported and discussed in relation to models for irradiation induced low temperature creep in ceramic fuels.     

Irradiation induced creep of ceramic nuclear fuels

This contribution gives a review of the experimental results and accompanying theoretical considerations. The mechanisms considered for irradiation creep are: relaxation of elastic stresses by fission spikes, promotion of dislocation slide by thermal spikes, preferential, stress-orientated nucleation of dislocation loops and preferential growth of dislocation loops. A survey over the irradiation creep rates attributed to steady-state creep shows $\text{ε}{}_{\mathrm{irr}}\text{~ σ · F}$ for oxide fuel in the stress and fission rate ranges of $\text{σ = 10–50}\text{MN/m}{}^2$ and $\text{F = 3 × 10}{}^{12}\text{–1 × 10}{}^{14}\text{f/cm}{}^3\text{· s}$ at burnups < 3%. There seems to be a continuous increase of the irradiation creep rate of oxide fuels with increasing temperature. However, that increase cannot be directly interpreted through a thermally activated process. It seems that the irradiation creep rate will also depend on fuel porosity, on plutonium distribution in mechanically blended UO₂-PuO₂, but not substantially on the plutonium content per se. Some results were already given for carbide and nitride fuels, which show the irradiation creep rate to be lower by about a factor of 10 than for oxide fuel under comparable conditions. Primary irradiation creep has been observed up to $\text{(3–5) × 10}{}^{19}\text{f/cm}{}^3$ and could prevail up to $\text{1 × 10}{}^{20}\text{f/cm}{}^3$.     

Fission-enhanced self-diffusion of uranium in UO2 and UC

In-pile self-diffusion measurements in stoichiometric UO₂ sinters and single crystals and in arc-cast stoichiometric UC have been performed using the thin layer condition and ²³³U as tracer. The nominal irradiation temperature was 900°C. The resulting diffusion coefficients $\text{D1}$ of $\text{1.5 × 10}{}^{\mathrm{?16}}\text{cm}{}^2\text{· sec}{}^{\mathrm{?1}}$ for UO₂ and $\text{2.2 × 10}{}^{\mathrm{?17}}\text{cm}{}^2\text{· sec}{}^{\mathrm{?1}}$ for UC for a fission rate $\text{S}$ of 1 × 10¹³f/cm³ · sec represent radiation enhanced diffusion and are higher by factors of 10³ to 10⁴ than (extrapolated) coefficients of thermal diffusion. The data are of immediate relevance for understanding and predicting such important quantities as in-pile sintering and densification, diffusion controlled creep and fission gas behavior in the outer zones of the fuel. They are at the upper limit of expected values.     

In-pile measurement of fission enhanced creep and swelling of uranium nitride

An irradiation device for the continuous and accurate in-pile measurement of fission enhanced creep in ceramic nuclear fuels has been designed and operated. The specimen stack was a hollow cylinder of 0.7 by 1 cm diameter and 2 cm gauge length, doubly contained, with sodium bonding in the inner container. The operational parameters were variable up to a fission rate of $\text{5 × 10}{}^{13}\text{cm}{}^{\mathrm{?3}}\text{sec}{}^{\mathrm{?1}}$, a temperature of 1100°C and a compressive stress of 70 MNm^?2. The strain was measured differentially on the gauge length by an inductive transducer. The device performed satisfactorily during one year of inpile operation. Deformations due to creep and swelling were measured at different stresses and temperatures on uranium nitride specimens. Routine measurements of deformations due to thermal expansion, and elastic strain at ambient and high temperature have proven the reliability and accuracy of the measuring and loading systems.     

Steady-state creep of zircaloy-4 fuel cladding from 940 to 1873 K

Steady-state creep rates of as-received zircaloy-4 fuel cladding have been determined from 940 to 1073 K in the α-Zr range, from 1140 to 1190 K in the mixed (α + β) phase region and from 1273 to 1873 K in the β-Zr phase region. Strain rates of between 10^?6 and 10^?2/s were determined under constant uniaxial load conditions. Assuming that creep rates can be described by a power law-Arrhenius equation, the creep rate for α-phase zircaloy-4 is given by: $\text{ge}{}_{\mathrm{ss}}\text{?}\text{= 2000 σ}{}^{5.32}\text{exp}\text{(?284 600/kT) s}{}^{\mathrm{?1}}$; for the β-phase zircaloy-4 by: $\text{ge}{}_{\mathrm{ss}}\text{?}\text{= 8.1 σ}{}^{3.79}\text{exp}\text{(?142 300/kT) s}{}^{\mathrm{?1}}$; and for the mixed (α + β) phase of zircaloy-4 (for creep rates ?3 × 10^?3 s^?1) by: $\text{ge}{}_{\mathrm{ss}}\text{?}\text{= 6.8 × 10}{}^{\mathrm{?3}}\text{σ}{}^{1.8}\text{exp}\text{(?56 600/kT) s}{}^{\mathrm{?1}}$. For the both the α and β phases, the activation energies for creep are in agreement with those of self-diffusion. For the mixed (α + β) phase region, the low creep rate range is controlled by grain boundary sliding at the α/(α + β) phase boundary.     

The mechanisms of fission gas release from (Th,U)O2

The releases of xenon from three (Th, U)O₂ specimens with different U contents were measured over a wide range of fission dose from $\text{2.9 × 10}{}^{19}$ to $\text{2.2 × 10}{}^{22}$ fissions m^?3 by using a post-irradiation technique. The releases were found to decrease with dose and to level off at higher doses. Measurements of the changes in lattice parameter and specific surface area of the same specimens enabled one to conclude that the decrease in release originates in the trapping of xenon by the vacancies and vacancy clusters induced by fission fragments. And the release mechanisms of fission gas were proposed based on the proper evaluation of the observation on radiation damage and recovery in oxide fuel.     

Re-solution-controlled fission-gas behavior in UO2 irradiated in a fast flux

Uranium dioxide irradiated in a fast neutron flux to a burnup of $\text{2 × 10}{}^{20}$ fissions/cm³ between 650 and 1400°C has been examined by transmission- and scanning-electron microscopy and replication metallography. The fission-gas distribution in the fuel matrix and grain boundaries has been characterized as a function of irradiation temperature and fission rate. The majority of fission gas produced even at the highest irradiation temperature was in the UO₂ matrix either in solution or in the form of bubbles < 20 Å in diameter. The results are explained on the basis of an irradiation-induced re-solution mechanism whereby fission gas from within bubbles is reinjected into metastable solution in the UO₂ lattice. Calculated fission-gas solubilities are given as a function of temperature for 10¹³, $\text{3 × 10}{}^{13}$, and 10¹⁴ fissions/cm³ · sec, and, based on these results, it is concluded that the re-solution process is operative over a substantial fuel volume of both light-water-reactor and fast-breeder-reactor oxide fuels.     

Observations on the low-temperature creep properties of uranium carbide under irradiation

The creep behaviour of 97% dense hyperstoichiometric UC has been examined during irradiation in three-point bend tests carried out at 450°C up to a dose of $\text{1.65 × 10}{}^{26}\text{fissions/m}{}^3$. A rapid decrease in measured strain rate with dose was observed at each stress level, nominally steady-state creep being established above $\text{≈ 1 × 10}{}^{26}\text{fissions/m}{}^3$ when the creep rate was a factor of 8 lower than that observed in UO₂ irradiated under identical conditions. Creep rates were found to be directly proportional to stress at high doses. Comparison of results from this test with data from other experiments up to $\text{2 × 10}{}^{25}\text{fissions/m}{}^3$ in compression and tension indicates little variation in the radiation-creep constant between 450°C and 800°C. The creep rate for UC, much lower than that observed in UO₂, is consistent with recently reported determinations of the effective uranium self-diffusion coefficients under irradiation in those materials.     

The tensile deformation behavior of fine-grained α-plutonium

Tensile deformation of extruded monoclinic α-plutonium with an average grain size of 4 μm was studied at stress from 2 500 to 100 000 psi (17.3 to 689 MN/m²) and temperatures from 22 to 108°C. The strain rate varied from 10^?9 to 7 × 10^?3 sec^?1. The relation, $\text{? = 2.86 × 10}{}^{\mathrm{?7}}\text{σ}{}^{4.2}\text{exp}\text{(?25 600/}\text{RT}\text{)}\text{sec}{}^{\mathrm{?1}}$, was obeyed from 12 000 to 60 000 psi (71.7 to 414 MN/m²) for strain rates greater than about 10^?6 sec^?1. Stress and temperature dependences of creep rate over this stress range were in accord with a dislocation climb controlled creep model, although the power law behavior occurred at stresses higher than theory predicts. The value of 25 600 cal/mole proved a reasonable value for the activation energy for self-diffusion in α-plutonium. At lower stresses the apparent activation energy for creep increased with decreasing stress, and the stress exponent $\text{n}\text{(=}\text{d log}\text{?/}\text{d log}\text{σ)}$ increased from 4.2 to 7.9. The high apparent activation energies for creep and high n values at low stresses were attributed to grain growth during creep. Tensile elongation increased with decreasing strain rate and increasing temperature over the entire stress range. Low elongation at high stresses was attributed to lack of grain boundary sliding. Grain size changed during creep toward a size determined by stress. At the highest test temperatures and lowest stresses grain growth occurred during large strains, while at high stresses the average grain size decreased.     

Fluage du pyrocarbone sous irradiation

Three massive samples of pyrocarbon were irradiated at 1100°C for a maximum fast-neutron dose of $\text{1.6 × 10}{}^{21}$ DNE. They were subjected to stresses in the range $\text{1.33 × 10}{}^2\text{–2 × 10}{}^2$ Kg/cm². The pyrocarbon was deposited from methane in a rotating furnace. Its density, its isotropy, its structure according to X-rays and TEM relate closely to its homologue deposited from methane in fluidised conditions. A study of creep under irradiation showed that a brief stage of primary creep is followed by a stage which is linear with respect to both stress and fast-neutron dose. Creep is thus well represented by an expression of the form $\text{? = Kσφ}$, where $\text{K}$ is $\text{2 × 10}{}^{\mathrm{?25}}$ (Kg · cm^?2. DNE)^?1, which is a value ten times greater than previously estimated. Irradiation is accompanied by densification, a slight increase in anisotropy and a reduction in $\text{L}{}_{\mathrm{c}}$ (apparent crystallite size measured along the $\text{c}$ axis). The variation of these parameters with dose does not, however, differ appreciably between the three creep samples and the unstressed sample.     

Irradiation creep in Zr single crystals

A single crystal of crystal bar Zr was irradiated, unstressed, at 570 K in a fast (> 1 MeV) neutron flux of $\text{5.5 × 10}{}^{16}\text{n/m}{}^2\text{-s}$. After a dose of $\text{6 × 10}{}^{23}\text{n/m}{}^2$ a tensile stress of 25 MPa was applied during a period of steady reactor power. The loading strain was an order of magnitude smaller than that observed when an identical, unirradiated, crystal was loaded to the same stress. There followed a period of primary creep during which the creep rate decreased to a value of $\text{5 × 10}{}^{\mathrm{?6}}\text{h}{}^{\mathrm{?1}}$ in the first 24 hours of the test. For the final 2000 hours of the test the specimen was observed to creep at a rate of $\text{1 × 10}{}^{\mathrm{?6}}\text{h}{}^{\mathrm{?1}}$ when the reactor was at full power. During shutdowns, the creep rate decreased with time. The results will be discussed and compared with predictions from current theories for the mechanism of irradiation enhanced creep in light of the micro-structures observed.     

Etude par compression a hautes temperatures de la deformation plastique du bioxyde d'uranium polycristallin

We have studied the compression of stoichiometric, sintered polycrystalline UO₂ as a function of strain rate $\text{?e (7 × 10}{}^{\mathrm{?4}}\text{/min < ?e < 7 × 10}{}_{\mathrm{?2}}\text{/}\text{min}\text{)}$, and temperature $\text{T (600 < T < 1700°}\text{C}\text{)}$. The brittle-to-ductile transition temperature is about 1000°C and we have studied by fractography the characteristics of the fracture at 600 and 800°C. In the plastic deformation range, two types of behaviour have been observed, (i) At low stress ($\text{5 < < 10}\text{kg/mm}{}^2$), the dépendance of strain rate with flow stress σ is $\text{?e = (}\text{cte}\text{/T) σ}{}^{\mathrm{4,5}}\text{exp}\text{-(ΔH/RT)}$ with $\text{ΔH = 96}\text{kcal/mol}$. (The deformation is probably by dislocation climb.) (ii) At high stress ($\text{σ > 10}\text{kg/mm}{}_2$), the deformation is heterogeneous and one observes a compression yield point, the magnitude of which decreases with increasing temperature. In this range, the analysis of the activation parameters has not allowed us to establish a very satisfactory correlation with any of the available deformation schemes.     

The diffusion of fission products in thorium metal

A study was performed of the diffusion in α-thorium of fission products representing impurity atoms with a diversity of size and valance differences with respect to the solvent lattice. The atoms were recoil injected into thorium disks. Diffusion coefficients were determined for ¹³³Xe by monitoring its release during annealing, and for the other isotopes by post-annealing concentration profile analysis. The Arrhenius constants $\text{d}{}_{\mathrm{o}}\text{(cm}{}^2\text{/sec)}$, resp. $\text{Q(kcal/mole)}$ were obtained for the diffusion coefficients where. $\text{D = D}{}_{\mathrm{o}}$exp$\text{(?Q/RT)}$;${}^{99}\text{Mo: 7.6 × 10}{}^{\mathrm{?4}}$ and $\text{37.6;}{}^{132}\text{Te: 1.32 × 10}{}^6$ and $\text{95.6;}{}^{133}\text{I: 2.7 × 10}{}^{\mathrm{?1}}$ and $\text{66.2;}{}^{133}\text{Xe: 3.6 × 10}{}^2$ and $\text{82.3;}{}^{140}\text{Ba: 2.3 × 10}{}^{\mathrm{?2}}$ and $\text{59.4;}{}^{141}\text{Ce ?}{}^{143}\text{Ce: 1.8 × 10}{}^{\mathrm{?2}}$ and 60.0. The fission product diffusion behavior, in general, fit either the vacancy or the substitutional-interstitial diffusion mechanisms for impurity atoms in a fcc metal. Both valence and ionic radius correlations were found. The data indicate low rates of diffusion for the operating temperatures at which α-thorium-based fuel might be used.     

The irradiation creep of nickel and AISI 321 stainless steel during 4 MeV proton bombardment

An apparatus has been developed to study the creep of thin metal specimens under tensile stress during bombardment by 4 MeV protons from the Harwell Van de Graaff Accelerator. The specimen is held in a helium atmosphere and the proton beam reaches it through a thin metal window at the end of the accelerator beam line. The proton beam passes through the thin (25 μm) specimen, losing ~1.5 MeV in the process (most of which contributes to heating the specimen) and creating almost uniform radiation damage at the rate of $\text{(1–10) × 10}{}^{\mathrm{?7}}$ displacements per atom per second (dpa s^?1). The specimen temperature is monitored by infra-red pyrometry and controlled to ± 0.2°C by additional DC heating via the infra-red pyrometer output to compensate for ion beam fluctuations. The irradiation creep strain of the specimen is continuously measured with a sensitivity of $\text{5 × 10}{}^{\mathrm{?6}}$ by a linear variable differential transformer. Irradiation times up to about 100h with reasonable beam stability are possible. Results are presented of the irradiation creep behaviour of pure Ni and both solution treated and cold-worked AISI 321 stainless steel bombarded in the temperature range 400–600°C under tensile stresses in the range 20–250 MPa.     

Irradiation-induced creep of mechanically blended porous UO2-PuO2 fuel

With regard to the behaviour of fast breeder reactor fuel, irradiation creep of mechanically blended, porous U_natO₂-15% PuO₂ was investigated. Some results for UO₂ are also quoted to clarify the dependence of creep rate on stress and temperature. The sintered density of the UO₂-PuO₂ samples amounted to 86% TD and 93,5% TD, their irradiation temperatures were between 300 and 1000°C, the stress in the samples at 15 and 40 MN/m², the fission rates between 2.5 and $\text{5 × 10}{}^{\mathrm{?9}}$ f/(U + Pu)-atom · s, and the maximum burnup at about 1%. The creep rates of UO₂-PuO₂ are much higher than previously measured on UO₂ of high density, but there was a good correspondence of the stress and temperature dependence. The difference of the creep rates cannot be explained only by the porosity of the UO₂-PuO₂ samples. Therefore the PuO₂ portion of the fuel, whose distribution is heavily inhomogeneous, is treated as additional “effective” porosity. By this means a suitable interpretation is obtained for the results below about 650°C. At higher temperatures, UO₂-PuO₂ of 86% TD showed a rapid initial densification up to about 93% TD, apparently together with a simultaneous homogenization of the fission-density distribution. The results measured could be interpreted without considering an influence of the Pu-content as such.     

Phase identification,microstructural characterization,phase microanalyses and leaching performance evaluation of SYNROC-FA crystalline ceramic waste form

SYNROC-FA, a crystalline ceramic waste form designed to contain 50 wt% Amine process, uranium-rich, high-level, radioactive waste for ultimate deep-geologic disposal, has been characterized using X-ray diffractometry (XRD), scanning electron microscopy (SEM) and transmission electron microscopy (TEM). Phase identification was carried out using XRD, electron diffraction in TEM, and backscattered electron imaging in SEM. Phase microanalyses were carried out using energy-dispersive X-ray analyzers (EDX) during SEM and TEM examinations. X-ray diffraction and grain microanalyses using EDX revealed the existence of a pyrochlore-structured phase CaU(Ti³⁺, Ti⁴⁺)₂O₇, perovskite (Ca, U)(Ti³⁺, Ti⁴⁺)O₃ and uraninite (U, Ca, Ti)O₂, while Ba-hollandite Ba(Al³⁺, Ti³⁺)₂Ti₅O₁₄ was identified using only XRD.The leaching resistance of SYNROC-FA was determined by carrying out a modified MCC-1 leach test in a simulated Canadian Shield groundwater at 90°C for 120 days. The normalized leach rate of Ba was $\text{6 × 10}{}^{\mathrm{-3}}\text{g · m}{}^{\mathrm{-2}}\text{· d}{}^{\mathrm{-1}}$ while the concentrations of U and other simulated fission products in the leachants were below the detection limits of inductively coupled plasma spectrometry and atomic absorption techniques. The leach rates of U and Ti were estimated to be less than $\text{6 × 10}{}^{\mathrm{-5}}$ and $\text{3 × 10}{}^{\mathrm{-5}}\text{g · m}{}^{\mathrm{-2}}$, respectively.     

Heterodiffusion de l'or,de l'argent et du cuivre dans les phases cubiques δ et ε du plutonium

The solute diffusion at infinite dilution of ¹⁹⁸Au and ^110mAg in cubic phases of Pu has been studied using the serial sectroning method. The solute diffusion coefficients in the b.c.c. ? phase can be expressed by: $\text{D}{}^{\mathrm{Au}}{}_{\mathrm{?Pu}}\text{= 5,7 × 10}{}^{\mathrm{?5}}$ exp(?10300/RT) cm²/s and $\text{D}{}^{\mathrm{Ag}}{}_{\mathrm{?Pu}}\text{= 4,9 × 10}{}^{\mathrm{?5}}\text{exp(?9600/RT) cm}{}^2\text{/s}$. The solute diffusion mechanism is interstitial of the dissociative type in both cases. These experiments confirm the activated interstitial model which has been proposed for self diffusion of ?Pu. Indeed the solute diffusion coefficients of Au and Ag are near of the self diffusion coefficients of Pu. The mechanisms are therefore interstitial in both cases. In the f.c.c. δ phase of Pu where self diffusion takes place by a vacancy mechanism, the solute diffusion coefficients of Au and Ag are near of the self diffusion coefficients of δ Pu. Solute diffusion takes place also by a vacancy mechanism. On the other hand, the extrapolation at infinite dilution of experiments of solute diffusion of Cu in ?Pu (Matano-Wagner coupling) gives the following results: $\text{D}{}^{\mathrm{Cu}}{}_{\mathrm{?Pu}}\text{= 1 × 10}{}^{\mathrm{?3}}\text{exp(?12300/RT)}$ cm²/s. The solute diffusion mechanism is interstitial of the dissociative type. In the ? phase the smaller the atomic radius the faster the migration: r_Co < r_Cu < r_?Pu < r_Ag = r_Au, and D^Co_?Pu > D^Cu_?Pu >D^Pu_?PU > D^Ag_?Pu ≈ D^Au_?Pu.     

Creep of cold-worked Zry-4 at 673 K

Creep data, at 673 K, up to times of about 400 h and stresses between 117.6 and 264.7 MPa, in flat specimens of cold-worked Zry-4, are reported. When viewed in a σ-/.ε diagram, the data can be represented by Hart's equation of state $\text{/.ε = /.ε}{}^1\text{[ln(σ}{}^1\text{/σ)]}{}^{\mathrm{<mtext>?1</mtext><mtext>\lambda </mtext>}}$, where σ¹ and /.ε¹ are related to the plastic strain and λ is a constant, with a value similar to that obtained by measurements of the stress-relaxation in bending of the same material and at the same temperature. No distinction is made between primary and steady-state creep, indicating that the same mechanism is controlling the plastic deformation in both regions. The apparent activation energy was found to be independent of stress with a value close to that for self diffusion.     

Investigations on swelling and fission gas behaviour in uranium dioxide

UO₂ irradiated at temperatures between 1000 and 2100 K was investigated with respect to fission gas behaviour and swelling. The amount of fission gas was measured in three steps as released fission gas, fission gas retained in bubbles and pores, and fission gas in the fuel matrix. The retained fission gas reaches concentrations up to $\text{1.6 × 10}{}^{\mathrm{?2}}$ gas atoms per uranium atom at temperatures below 1250 K and decreases with increasing temperature. The swelling was evaluated by measuring the volume changes and by immersion density measurements. The maximum fission gas swelling without extensive bubble migration is about 20% at 2000 K. It diminishes to about 5% at 1250 K.     

Plastic flow of ordered Zr3Al

Room-temperature tensile experiments established that ordered Zr₃Al of the Ll₂ type obeys the relationship $\text{σ}{}_{\mathrm{?}}\text{= σ}{}_{\mathrm{0,?}}\text{+ kd}{}^{\mathrm{<mtext>?</mtext><mtext>1</mtext><mtext>2</mtext>}}$ where σ_? is the flow stress at a given strain ?, σ_0,? is a strain-dependent frictional stress, $\text{d}$ is the average grain diameter, and $\text{k}$ is a strain-independent constant of magnitude $\text{(2.8 ± 0.3) kg/mm}{}^{\mathrm{<mtext>3</mtext><mtext>2</mtext>}}$. Zr₃Al flows by fine, planar slip, and is susceptible to intergranular cracking.