Influence de la dose de neutrons rapides sur la formation des cavites dans l'aluminium

The results of neutron irradiation of aluminium at 55°C under a fast flux of $\text{(1.4 ± 0.2) 10}{}^{14}\text{n · cm}{}^{\mathrm{?2}}\text{· sec}{}^{\mathrm{?1}}$ to doses ranging between 10¹⁹ and $\text{3 × 10}{}^{21}\text{n · cm}{}^{\mathrm{?2}}$ are presented. Experimental observations suggest that dislocations play an important role in void nucleation and growth. Monte-Carlo calculations of void growth in the presence of a dislocation describe reasonably well the experimental results.     

Damage in neutron-irradiated molybdenum: (I). Characterization of as-irradiated microstructure

The as-irradiated microstructure of molybdenum, irradiated in the EBR-II reactor at six different temperatures in the range 430–1000°C (0.24–0.44 $\text{T}{}_{\mathrm{m}}$) to a fast neutron fluence of ≈ 1 × 10²² n · cm^?2 ($\text{E > 1 MeV}$), has been characterized as black spot clusters, loops, rafts, voids (random and ordered) and dislocations. Present results show that both the void density, $\text{N}$_v and the void size, $\text{d}{}_{\mathrm{v}}$, are independent of irradiation temperature in the range 430–700° C. Above the 700° C irradiation temperature the void density decreases and the void size increases exponentially with increasing irradiation temperature and they have been expressed empirically as $\text{N}{}_{\mathrm{<mtext>v</mtext>}}\text{= 3.6 × 10}{}^{20}\text{exp}\text{(?26.9 T/T}{}_{\mathrm{<mtext>m</mtext>}}\text{)}$, $\text{d}{}_{\mathrm{<mtext>v</mtext>}}\text{= 1.5}\text{exp}\text{(9.44 T/T}{}_{\mathrm{<mtext>m</mtext>}}\text{)}$, where $\text{T/T}{}_{\mathrm{<mtext>m</mtext>}}$ is the irradiation temperature presented as a fraction of the melting point. The void density of all available published data has been used to show that the void density is (a) a strong function of irradiation temperature for a constant number of displacements per atom (dpa) and (b) a function of reactor power and spectrum when normalized to dpa.     

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.     

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.     

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.     

The oxygen snoek damping and an abnormal peak of neutron-irradiated vanadium

Measurements of low-frequency internal friction and electron microscope observations were made on neutron-irradiated vanadium with various oxygen contents. Irradiation was carried out at about 60°C to a fast fluence of $\text{2 × 10}{}^{17}$ or $\text{5 × 10}{}^{19}$ n/cm² ($\text{E ? 1 MeV}$). The oxygen Snoek damping was decreased by irradiation and post-irradiation annealing below 200 or 250° C, while it began to recover by annealing above this temperature. Complete recovery was attained by 30 min anneal at 450°C in the case of the lower fluence, whereas in the other case it was not observed after the same treatment. The results of electron microscope observations were consistent with those of internal friction measurements. The specimens irradiated to $\text{5 × 10}{}^{19}$ n/cm² showed an abnormal peak after annealing above 250°C near the nitrogen Snoek temperature. The height of this peak, $\text{P}{}^{\mathrm{?1}}{}_{\mathrm{<mtext>max</mtext>}}$, was expressed as $\text{P}{}^{\mathrm{?1}}{}_{\mathrm{<mtext>max</mtext>}}$ ∝ exp ($\text{2.72 × 10}{}^3\text{/RT}$) $\text{Q}{}^{\mathrm{?1}}{}_{\mathrm{<mtext>max</mtext>}}$, where $\text{Q}{}^{\mathrm{?1}}{}_{\mathrm{<mtext>max</mtext>}}$ the heiβht of the oxygen Snoek damping after each annealing. The mechanism for radiation-anneal hardening and the abnormal peak were considered in the light of these experiments.     

Neutron irradiation effects in uranium dicarbides

Changes of electrical resistivity and lattice parameter in UC_1.96 after neutron irradiation from $\text{9 × 10}{}^{14}$ to $\text{2 × 10}{}^{18}$ nvt were studied. The resistivity was increased with the dose up to $\text{1 × 10}{}^7$ nvt, and saturated at that dose. Above 10¹⁸ nvt a steep increase was observed. In the lattice-parameter changes, on the other hand, a gradual increase was observed in the dose range between $\text{2 × 10}{}^{16}$ and $\text{8 × 10}{}^{17}$ nvt; above that dose, an abrupt increase followed. Annealing experiments on the resistivity were performed up to 1000°C using the specimens irradiated to the low dose of $\text{5 × 10}{}^{16}$ nvt, and the increased resistivity was completely recovered in three steps. The activation energies of each step were estimated to be 0.3, 0.5 and $\text{1.6 ± 0.2}$ eV.     

Effect of cyclotron-injected helium on void-formation characteristics of type 304 stainless steel

Samples of Type 304 stainless steel were injected with helium by cyclotron bombardment to concentrations ranging between $\text{1.1 × 10}{}^{\mathrm{?7}}$ and $\text{1 × 10}{}^{\mathrm{?4}}$ ppma. Following cyclotron injection, the samples were given a variety of heat treatments prior to insertion in EBR-II for irradiation at 450 °C to a total dose of $\text{1 × 10}{}^{21}\text{n/cm}{}^2$. Samples that were not heat treated or that were annealed at 650 °C following cyclotron injection formed few voids and dislocation loops after EBR-II irradiation. This behavior is apparently due to the precipitate clusters that were formed during the helium injection. These precipitates were analyzed by electron microscopic techniques and found to have spherically symmetric strain fields that were of interstitial character. Samples that were annealed at 760 °C following cyclotron injection formed a larger number density of both voids and dislocation loops than did the control sample after EBR-II irradiation. The void volume also exceeded that of the control. Clustering of the dislocation loop population near grain boundaries and precipitate particles was observed in the control and low helium concentration samples.     

The damage structure formed in molybdenum by irradiation in a fast reactor at 650°c

Transmission electron microscopy has been used to study the damage structure of commercially pure and zone-refined molybdenum irradiated in a fast reactor to a total dose of $\text{～3 × 10}{}^{22}\text{n/cm}{}^2$ at ～650°C. In all cases the structures consist of coarsely distributed dislocation segments and loops, and a considerably finer distribution of small voids. The voids tend to be ordered on a bcc superlattice parallel to the underlying host lattice. Although the loops are predominantly interstitial in nature, a significant number of small vacacy loops are present in both the commercially pure and zone-refined materials. The formation of vacancy loops during irradiation can only be accounted for by in-site collapse of displacement spikes. The occurrence of such a process implies that the formation and growth of voids is dependent on emission of vacancies from the loops. An important practical consequence of having a high density of voids compared to other sinks for point defects is that the voids themselves act as dominant neutral sinks for vacancies and interstitials, leading to an early saturation in void growth.     

Nickel ion bombardment of type 304 stainless steel: Comparison with fast reactor swelling data

Annealed Type 304 stainless steel containing 15 atomic ppm of helium has been bombarded with 5 MeV nickel ions at 525°C to 700°C. A pronounced swelling peak occurs at 625°C, compared to a swelling peak temperature of about 475°C in reactor. TEM measurements of void swelling at 625°C as a function of ion dose show a swelling of almost 40% at 124 dpa without evidence of saturation. Measurements of gross swelling of the ion-bombarded material by a new step-height method provide information that is in good agreement with TEM data, and can be extended to larger swellings. The step-height results indicate a swelling of over 90% at 290 dpa at 625°C. The ion-produced swelling agrees well with in-reactor data when the two are compared at the respective peak swelling temperatures, and the void concentrations and average void diameters are comparable for the two cases. The high ion dose results are used to guide extrapolation of reactor data to higher fluences, leading to the following predictions for swellings at the peak swelling temperature in reactor: 18% swelling at $\text{1× 10}{}^{23}\text{n/cm}{}^2$ (fast), 50% at $\text{2 × 10}{}^{23}$, and 80% at $\text{3 × 10}{}^{23}$.     

Annealing behavior of neutron irradiated β-SiC

The post-irradiation annealing behavior of β-SiC for use as a monitor of irradiation temperature is discussed. Powder and rods of polycrystalline β-SiC were irradiated to $\text{1.5 × 10}{}^{17}$ to $\text{5.0 × 10}{}^{19}\text{n/cm}{}^2\text{(E > 0.18 MeV)}$ at temperatures between 290 and 500°C. The estimated temperatures deduced from the changes in lattice constant and specific electric resistivity during progressive annealing, and from thermal expansion measurement by high-temperature X-ray diffraction agreed with values determined by means of a thermocouple. Thermal expansion measurement in a conventional dilatometer resulted in an over estimate of the irradiation temperature, and further improvement of this method is required for experimental application.     

Irradiation damage in Zircaloy-2 produced by high-dose ion bombardment

Damage produced in Zircaloy-2 by 5 MeV Ni ion bombardment was examined by transmission electron microscopy. Nickel ion bombardment was done over an integrated flux ranging from about $\text{5.6 × 10}{}^{15}$ to $\text{7 × 10}{}^{16}\text{Ni}{}^{\mathrm{++}}$ ions/cm² and over the temperatures from 300 to 600°C. The gross damage which occurred at low temperature consisted of dark unresolved areas and uniformly distributed ‘black dots’, somewhat similar to those observed in neutron-irradiated Zircaloy. With increasing irradiation temperature and dose, two features have appeared in the microstructures. One of the features was randomly oriented dislocation loops and the other well-defined alignment of defects with the projection of aligned defects oriented in the «101&#x0304;0å> directions. Detailed examination of these aligned defects revealed that they exhibit two kinds of images; one is associated with precipitation of second-phase material and the other with lattice displacement of matrix material. With further increase in irradiation temperature, the damage structure changed markedly to that of random dislocation network.     

Study of 3He+-bombarded palladium using nuclear magnetic resonance techniques

The nuclear magnetic resonance (NMR) from ion-implanted ³He atoms has been observed in ³He⁺-bombarded palladium. Two 1 μm palladium thin films, one on each side of a copper foil substrate, were bombarded at 75°C with 75 and 140 keV ³He⁺-ions to a fluence of $\text{～10}{}^{18}{}^3\text{He}{}^{\mathrm{+}}\text{/}\text{cm}{}^2$ for each energy. Spin-lattice ($\text{T}{}_1$) and spin-spin ($\text{T}{}_2$) relaxation times were measured at 10, 20, and 35 MHz in the temperature range 1–4K. The ³He nuclear relaxation data indicate that the implanted atoms after the 75°C bombardment are situated in small clusters within the Pd thin films.     

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.     

The transformation behavior of a U-16.4 at % Nb-5.6 at % Zr alloy

The isothermal transformation kinetics of a γ-phase U-16.4 at % Nb-5.6 at % Zr alloy have been studied by resistometric techniques between 130 and 590° C. Three distinct regions of transformation were observed. Above 420° C the transformation proceeded by nucleation and growth and progressed to the equilibrium $\text{α + γ}{}_2$ phases. Aging at temperatures between 270 and 400° C resulted in the appearance of a transition $\text{α}{}^{\mathrm{″}}$ monoclinic phase that remained stable after even 7 days of aging at 380° C. Aging below 250° C produced a tetragonal $\text{γ}{}^0$ phase. A negative temperature coefficient of resistance is observed at temperatures below 400° C when the material is quenched from a 900° C anneal. It is suggested that 400° C represents an athermal γ to $\text{γ}{}^{\mathrm{<mtext>S</mtext>}}$ transformation and that the negative resistance coefficient results from a continual change of lattice parameters of $\text{γ}{}^{\mathrm{<mtext>S</mtext>}}$ with decreasing temperature.     

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.     

Tensile properties of annealed type 316 stainless steel after EBR-II irradiation

The effects of fast reactor neutron irradiation on the tensile properties of annealed Type 316 stainless steel were determined over a wide range of irradiation temperatures and reactor fluences. These effects are described as a function fluence and temperature to $\text{7 × 10}{}^{22}\text{n/cm}{}^2$ ($\text{E}{}_{\mathrm{n}}\text{> 0.1}\text{MeV}$) at 430 to 820 °C. The usual flow stress increase and ductility decrease were observed with increasing neutron fluence. Strengthening decreases continuously from 480 °C to ≈ 700 °C with no hardening at or above 760 °C. Elongation values increase with temperature in the 430–540 °C range and generally decrease with temperature above 540 °C.     

The effect of minor constituents on swelling in stainless steel

The small quantities of solute interstitial elements in stainless steel (C, N and possibly Si) reduce the swelling under neutron irradiation ($\text{～ 2 × 10}{}^{22}\text{neutrons/cm}{}^2$) by more than an order of magnitude between 500 and 600° C over high purity material. The solute interstitials reduce both the numbers and sizes of irradiation-caused voids. Current swelling models ignore — of necessity — this gross effect. Several possible mechanisms are suggested to account for the effect.     

The effect of temperature and neutron irradiation upon the fatigue-crack propagation behavior of ASTM A533-B steel

The fatique-crack propagation behaviour of A533-B steel was studied within the framework of linear-elastic fracture mechanics. Tests were conducted at 75° F (24° C) and 550° F (288°C) on unirradiated material, and on material irradiated at 550° F to $\text{2.3 – 2.8 × 10}{}^{19}\text{n/cm}{}^2$ and $\text{5.3 – 5.7 × 10}{}^{19}\text{n/cm}{}^2\text{(E > 1 MeV)}$. In general, at the cyclic frequency used (600 cpm), neither temperature nor neutron irradiation had a significant effect on the fatigue-crack propagation.     

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.