Some observations of density and porosity changes in UO2 fuel irradiated in water-cooled reactors

A technique has been developed for the hot-cell measurement of the apparent density of irradiated UO₂ fuel after extraction from a fuel pin. A single determination is accurate to $\text{± 3 \%}$ at the 95 % confidence limit. The method has been applied to fuel irradiated in thermal neutron fluxes in the Winfrith SGHWR and in the Halden BWR. Material has been examined at ratings of 1–51 W/g and in the burn-up range $\text{0.09–5.79 × 10}{}^{20}\text{fissions/cm}$. It is concluded that pellets with peak temperatures below 1100°C densify during irradiation, but at higher temperatures the pellets begin to swell. Fuel micrography has shown that the densification is principally due to the loss of micropores with a temperature dependency given by an activation energy of 5200 cal/mol. Above 1000°C the densification is masked by the formation and growth of intergranular fission gas bubbles, whose volume may exceed that of the manufactured pores which have sintered. In solid fuel pellets central swelling did not balance densification in the cooler rim until the fuel centre temperature exceeded 1700°C.     

Postirradiation Annealing of Boron Carbide Pellet Irradiated in Fast Breeder Reactor

This paper reports some irradiation effects and recovery behavior of neutron irradiated boron carbide pellets that were used as control rod elements in the Enrico Fermi Fast Breeder Reactor. Measurements were carried out on changes in lattice parameters, thermal expansion, helium release, elastic moduli and microstructure observations by annealing the irradiated pellets at elevated temperatures. The increase in unit cell volume of B₄C upon irradiation was found to be 0.22%. The recovery in lattice parameter began at around 500°C and completed at 1,000°C. It was found that the pellet showed a sharp increase in a dimensional change at about 700 to 800°C with a large amount of helium release, and the pellet which showed larger swelling released smaller amount of helium.     

The temperature dependence of nickel-ion damage in nickel

The temperature dependence of void and dislocation structures was studied in high-purity nickel irradiated with 2.8 MeV ⁵⁸Ni⁺ ions to a displacement density of 13 displacements per atom (dpa) at a displacement rate of $\text{7 × 10}{}^{\mathrm{?2}}\text{dpa/sec}$ over the temperature range 325 to 625°C. Dislocation loops, with no significant concentrations of voids, were observed in specimens irradiated at 475°C and below. Specimens irradiated between 525 and 725°C contained both voids and dislocations. The maximum swelling was measured as 1.2% at 625°C. Analysis of the data by theoretical models for void nucleation and growth indicated that the swelling in the present experiment was principally limited by void growth at low temperatures and by void nucleation at high temperatures. The data were also compared with previously reported neutron and nickel-ion irradiation results.     

Observation of a partially-ordered void lattice in aluminium irradiated with 400 keV Al+ ions

Transmission electron microscopy observations of voids formed in aluminium during irradiation at 50°C and 75°C with 400 keV Al⁺ ions, have shown that partially-ordered void arrays are often present. These arrays occur in high-purity annealed aluminium, which has been implanted with 10^?4 atom/atom helium before ion irradiation. The void concentration is found to be $\text{～3 × 10}{}^{16}\text{/ cm}{}^3$, and the void lattice parameter ～ 700 Å. The ratio of void lattice parameter to void radius is ～ 12. Ordered void lattices have been observed frequently in irradiated body-centred cubic metals but the only previous observation for a face-centred cubic metal was in nickel. Theoretical predictions of void lattices in metals are discussed and related to the observations reported herein.     

Neutron damage and fission product gas release in the initial burst region of uranium monocarbides

Thermal neutron damage and fission product gas (¹³³ Xe) release in a burst region of uranium monocarbides were studied. After neutron irradiation, the electrical resistivity was measured from room temperature to 800° C. Three recovery stages were revealed in the resistivity of UC irradiated to $\text{4.0 × 10}{}^{16}$ nvt. The lattice parameter of UC with the same irradiation also showed three stages of recovery up to 1050°C. The initial burst of Xe from UC was studied in a dose range between $\text{1.6 × 10}{}^{15}$ and $\text{2.9 × 10}{}^{18}$ nvt. The burst occurred in three steps for lightly irradiated specimens, while there were two steps of the burst in heavily irradiated specimens. The activation energies for each burst step were calculated. From the results obtained here, we concluded that the burst was correlated with the recovery of damage in the neutron-irradiated UC.     

Thermal conductivity of neutron-irradiated pyrolytic β-silicon carbide

Samples of pyrolytic β-silicon carbide deposited at 1400 °C (grain size $\text{～ 1 μ}\text{m}$) and at 1750 °C (grain size $\text{～ 3 μ}\text{m}$) were irradiated with fast neutrons to 2.7-7.7 × 10²¹ n/cm² ($\text{E > 0.18 MeV}$) at 550 °–1100 °C. Irradiation reduced the room-temperature thermal conductivity from $\text{～0.15}\text{cal/cm}\text{·}\text{sec}\text{· °}\text{C}$ to $\text{～ 0.02}\text{cal/cm}\text{·}\text{sec}\text{· °}\text{C}$ after irradiation at 550 °C and to $\text{～ 0.05}\text{cal/cm}\text{·}\text{sec}\text{· °}\text{C}$ for an irradiation temperature of 1100 °C. The thermal conductivity of unirradiated samples decreased with increasing measurement temperature, while that of the irradiated samples was much less temperature dependent. No difference in behaviour was found between the samples with $\text{～ 1 μ}\text{m}$ grain size and the samples with $\text{～ 3 μ}\text{m}$ grain size.     

Elevated-temperature tensile properties of irradiated 214 Cr-1 Mo steel

The effect of neutron irradiation on the tensile properties of normalized-and-tempered $\text{2}\text{1}\text{4}$ Cr-1 Mo steel was determined for specimens irradiated in Experimental Breeder Reactor II (EBR-II) at 390 to 550°C. Two types of unirradiated control specimens were tested: as-heat-treated specimens and as-heat-treated specimens aged for 5000 h at the irradiation temperatures. Irradiation to approximately 9 dpa at 390° C increased the strength and decreased the ductility compared to the control specimens. Softening occurred in samples irradiated and tested at temperatures of 450, 500, and 550 °C; the amount of softening increased with increasing temperature. The tensile results were explained in terms of the displacement damage caused by the irradiation and changes in carbide precipitates that occur during elevated-temperature exposure.     

Voids resulting from fast neutron irradiation of a stabilized stainless steel

A 16 Cr-13 Ni-niobium stabilized stainless steel (Type 1.4988) was irradiated as fuel pin cladding in the DFR reactor. The irradiation temperatures ranged from 300 to 650 °C. Neutron fluences from 3.3 to 4.0 n/cm²$\text{(E > 0.1 MeV)}$ were achieved. Swelling behavior was studied in this material by transmission electron microscopy, immersion density determinations and diametral change measurements.Void formation was observed in the temperature range 360–610 °C. Void concentration values at lower temperatures (< 480 °C) are comparable to those cited for the unstabilized stainless steels AISI 304 and 316, however, they decrease more strongly with increasing irradiation temperatures. The mean and maximum diameters show a pronounced maximum at about 530 °C. The decrease of the void diameters at higher temperatures can be explained by the nature of the cavities observed. Annealing experiments with specimens irradiated at 610 °C have revealed a bubble or a mixed void-bubble character for these cavities.Swelling in type 1.4988 was found to be lower than that reported previously for types 304 and 316 at comparable irradiation and material conditions. This is especially pronounced at higher temperature. Diameter changes of the pin at $\text{T >- 610 °C}$ cannot be explained by void-formation.     

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.     

Results of in-pile creep measurements on pressurized capsules made of stabilized stainless steels WNr. 1.4981, CW and WNr. 1.4970, CW,A

An in-pile creep experiment is being performed at present in the RAPSODIE reactor with the objective of getting basic data for the SNR-reference materials. Pressurized capsules with pressure levels up to 300 bars, manufactured out of two heats of the German grade stainless steels X 8 Cr Ni Mo Nb 1616 (WNr. 1.4981,cw) and of X 10 Cr Ni Mo Ti B 1515 (WNr. 1.4970,cw, a), are irradiated between 400 and 505°C up to the target fluence of $\text{1 × 10}{}^{23}\text{n/cm}{}^2$ total (48 dpa NRT standard), in the core mid-plane. Measurements of the diameter changes of the capsules after two irradiation runs, have been carried out. Results of the second intermediate examination (29 dpa NRT standard) are reported and discussed in terms of dose dependence, creep rate dependence and interrelation of irradiation-induced creep and swelling. It is observed that irradiation creep is enhanced with the onset of swelling. A linear relationship exists between creep deformation and stress in the whole flux range investigated.     

Swelling of type-316 stainless steel at high fluences in EBR-II

A test to measure swelling induced by fast neutron irradiation in unstressed specimens of type-316 stainless steel has completed irradiation in the EBR-II reactor. Results are reported and discussed which describe the swelling as a function of neutron fluence, temperature of irradiation and extent of cold work in the alloy. Density determinations showed swellings of up to 15% $\text{ΔV}\text{V}{}_{\mathrm{f}}$ for 20% cold worked type-316 stainless steel at a neutron fluence level of 1.4 × 10²³n/cm², E > 0.1 MeV (70 dpa). The peak swelling temperature range was 550°C–600°C regardless of the extent of cold working. Increasing the cold work level reduced the swelling and tended to broaden the swelling temperature peak. Transmission electron microscopy (TEM) investigations showed that cold working had reduced the average void sizes compared to those observed in the solution annealed material.     

Defect development in neutron irradiated type 316 stainless steel

The effects of fast neutron irradiation on the defect development in unstressed solution treated Type 316 stainless steel were investigated by transmission electron microscopy. The irradiation conditions investigated covered the fluence range from 0.75 to $\text{5.1 × 10}{}^{22}\text{n/cm}{}^2$ ($\text{E > 0.1}\text{MeV}$) and temperatures from 380 to 850°C. Empirical equations were developed relating the void volume, void number density, mean void size, Frank faulted loop diameter, Frank loop number density and dislocation density with the neutron fluence and irradiation temperature. Void nucleation changes from homogeneous at low irradiation temperature (? 400°C) to heterogeneous at higher temperatures in that voids are preferentially associated with irradiation induced rod shaped precipitates. The void number density decreases while the void diameter increases with irradiation temperature. The total faulted loop line length per unit volume and dislocation density increases with fluence and decreases with temperature. The Frank loop diameter increases and number density decreases with temperature. The range of temperature in which Frank faulted loop formation occurs decreases with neutron fluence.     

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.     

Residual stress behavior in fast neutron irradiated SA AISI 304l stainless steel cylindrical tubing

Residual stress measurements were made on solution-annealed (SA) AISI 304L stainless steel (SS) irradiated in EBR-II over a temperature range from 402 to 524°C by axially slitting short sections of tubing. The data were analyzed by using SA AISI 304 SS physical properties and SA AISI 304L SS swelling and irradiation creep empirical equations to calculate the slit width change (δ) versus fluence (φt) curve. At temperatures equal to and above 445°C, δ versus φt calculations indicate that the stress effect on swelling is sufficiently large to reduce the swelling rate temperature gradient, and consequently the on-power stress gradient, to zero. This behavior is confirmed by void volume gradient measurements. At lower temperatures, δ versus φt calculations indicate that stress affected swelling is smaller and does not relax the swelling rate temperature gradient. Void volume gradient measurements confirm the presence of a swelling gradient. Calculations of the δ versus φt curve were made with four different empirical swelling equation fluence dependencies, and the best agreement with the δ versus φt data was obtained with a power form type swelling equation. The equations fit the immersion density data ($\text{ΔV}\text{V}{}_0\text{versus}\text{φt}$) within experimental scatter, but predict significantly different δ versus φt behavior. These results show that the slit tube results are very sensitive to the empirical swelling equation form.     

Mechanical properties of ferrous alloys after fast neutron-irradiation at high temperature

The results are given of an international “round-robin” experiment to study the nature of the damage structure in neutron irradiated zirconium and zircaloy-2 using transmission electron microscopy. The damage structure consists entirely of $\text{1}\text{3}\text{α<11}\text{2}\text{?}\text{0>}$ dislocation loops and no evidence has been found for c-component loops. Both vacancy and interstitial loops were found in specimens irradiated at 400 °C, with an excess of vacancy loops. Quantitative measurements of loop size distributions and loop concentrations are reported. All specimens exhibited “corduroy” contrast to varying degrees. The importance of choice of imaging conditions to minimize the contrast from thin foil artefacts such as oxide films and surface hydrides is stressed. The significance of the results is briefly discussed with reference to current theories of irradiation growth.     

The effect of additives on the irradiation behaviour of UO2

Samples of UO₂ doped with small amounts of Nb₂O₅ or La₂O₃, and having various grain sizes, have been irradiated at 1500°C to 0.1% FIMA. At this low burn-up, gas release and swelling measurements show no dependence on dopant, but the Booth model prediction of swelling proportional to reciprocal grain size has been verified. Gas release does not fit the simple Booth model at the low releases measured, and shows a dependence on sample density, and hence surface area only. A model has been derived to explain these results. The rare gas diffusion coefficient in UO₂ at 1500°C has been measured to be $\text{1.6 × 10}{}^{\mathrm{?19}}$ m²/s.     

Neutron irradiation-induced voids in β-silicon carbide

Pyrolytic β-silicon carbide was irradiated at temperatures between 625°C and 1500°C to neutron fluences up to 12 × 10²¹ n/cm² ($\text{E > 0.18 MeV}$). Density changes were measured and the samples were examined by transmission electron microscopy. Irradiation below 1000°C created small Frank dislocation loops on {111} planes. Irradiation at 1250°C and 1500°C produced tetrahedral voids which caused continuing expansion of the samples. Void sizes increased with increasing fluence and with increasing irradiation temperature, while void concentrations decreased with increasing temperature. One-hour post-irradiation anneals at 1700°C to 2100°C reduced the void concentration and total void volume while increasing the maximum void size.     

Retention and release of water by sintered uranium dioxide

The release of water and hydrogen upon heating sintered $\text{UO}{}_2$ pellets was measured by a direct mass spectrometric method of vacuum outgassing. The technique avoids water loss by sample transfer and measures, rather than the cumulative release, the rate of release with a sensitivity of 1 μg of water (as D₂O) per hour. Exposure of high-density $\text{UO}{}_2$ pellets to water (liquid D₂O) results in negligible water adsorption. Water in pellets fabricated with especially high open porosity (5%) was driven off by a linear temperature ramp below 200°C. A drying model for this process was developed and applied to the data. Strongly bound water was introduced into high-density UO₂by sintering in an atmosphere of D₂O and D₂. Release of the water or hydrogen began at ~500°C and was complete only at the melting point of UO₂(2800°C). The release kinetics are not diffusion-controlled; rather the process is governed by the rates of desorption of bound hydrogen-bearing species from at least three binding sites in the solid characterized by interaction energies between 20 and 50 kcal/mol. The D₂O/D₂ ratio of the desorbed gas was > 1 and did not correspond to thermodynamic equilibrium with stoichiometric urania. Hydrogen and water release kinetics are comparable below 2&#x0303;000°C, suggesting a common bound precursor. The total hydrogen (as D₂O or D₂) absorbed in the specimens was between 2 and 4 μg/g $\text{UO}{}_2$.     

Einführung in die Werkstoffwissenschaft (in German): ed. W. Schatt (VEB Deutscher Verlag für Grundstoffindustrie,Leipzig, 1972, DM 35)

Copper samples were irradiated with fast neutrons at temperatures in the range 220–550 °C and at instantaneous fluxes in the range 2 × 10¹³-3 × 10¹⁴ n/cm².sec > 0.1 MeV. The maximum swelling was observed at 0.45 $\text{T}{}_{\mathrm{f}}$ for an instantaneous flux of 3 × 10¹⁴ n/cm²-sec. A fourfold reduction of the instantaneous flux, at constant dose, displaces the maximum to lower temperatures and slightly increases its magnitude. Cold work before irradiation does not appear to have a significant effect on swelling. Alloying with solutes which lower the stacking-fault energy appears to displace the domains of swelling towards lower temperatures for a fixed instantaneous flux and towards lower flux for a fixed temperature.