Effect of helium preinjection and prior thermomechanical treatment on the microstructure of type 316 SS

Samples of 316 SS were preinjected with 15 appm helium either hot (650°C) or cold (room temperature) and irradiated with 3 MeV Ni⁺ ions to a dose level of 25 dpa at 625°C in order to test the validity of helium preinjection as a means of simulation of transmutant helium production. Results for preinjected and single-ion irradiated samples were compared to samples irradiated at 625°C to a 25 dpa dose level with 3 MeV Ni⁺ and simultaneously injected with helium at a rate of 15 appm He/dpa (dual-ion irradiated samples). Preinjected samples exhibited bimodal cavity size distributions. Preinjected samples of solution annealed or solution annealed and aged material showed lower swelling than dual-ion irradiated samples. However, He preinjected 20% cold worked samples showed greater swelling than dual-ion irradiated samples.     

Influence of helium injection schedule and prior thermomechanical treatment on the microstructure of type 316 SS

The influence of different helium injection schedules on microstructure development in Ni⁺ ion-irradiated 316 SS at 625°C is discussed. Injection schedules were chosen to either approximate the MFR condition or mimic the mixed-spectrum reactor condition. Dual-ion irradiation to 25 dpa produced strongly bimodal cavity size distributions in solution annealed and solution annealed and aged samples, whereas single-ion irradiation followed by dual-ion irradiation to the same dose produced a cavity size distribution with a substantial component of intermediate-size cavities. Dual-ion irradiation produced only very small cavities in 20% CW material, while single-ion followed by dual-ion irradiation produced some intermediate size cavities and greater swelling.     

Void Swelling of Solution-Annealed Type 316 Stainless Steel under Long Time Proton Irradiation

Solution-annealed type 316 stainless steel was irradiated by 150 keV proton to a dose of about 6 dpa at the irradiation temperature ranging 450–700°C. To examine the effect of aging during irradiation, the present proton irradiation was carried out for about 25 h at a low dose rate of 7×10–^?5dpa/s. The specimens without He preinjection showed much smaller void swelling than those preinjected with He to the content of 10 at.ppm. Similarly to the case of neutron irradiations, the void swelling in the He preinjected specimens showed the temperature dependence with double peaks, and the peak swelling temperatures were about 550 and 650°C. In these specimens with He preinjection. void number density decreased and average void diameter increased with the increase of irradiation temperature in the range of 450–600°C, but these trends were reversed between 600 and 650°C. The volume of the grain boudary M₂₃C6 precipitates increased with the increase of irradiation temperature from 600 to 700°C, and it was concluded that the decrease of soluble carbon due to the precipitation of M₂₃C6 caused the second swelling peak at 650°C.     

The migration of intragranular fission gas bubbles in irradiated uranium dioxide

The mobility of intragranular fission gas bubbles in uranium dioxide, irradiated at 1600–1800°C, has been studied following isothermal annealing at temperatures below 1600°C. The intragranular fission gas bubbles, average diameter approximately 2 nm, are virtually immobile at temperatures below 1500°C. The bubbles have clean surfaces with no solid fission product contamination and are faceted to the highest observed irradiation temperature of 1800°C. This bubble faceting is believed to be a major cause of bubble immobility. In fuel operating below 1500°C the predominant mechanism allowing the growth of intergranular bubbles and the subsequent gas release must be the diffusion of dissolved gas atoms rather than the movement of entire intragranular bubbles.     

Thermal Recovery of Radiation Defects and Microstructural Change in Irradiated UO2 Fuels

Abstract

Thermal recovery of radiation defects and microstructural change in UO₂ fuels irradiated under LWR conditions (burnup: 25 and 44 GWd/t) have been studied after annealing at temperature range of 450-1,800°C by X-ray diffractometry and transmission electron microscopy (TEM). The lattice parameter of as-irradiated fuels increase with higher burnup, which was mainly due to the accumulation of fission induced point defects. The lattice parameter for both fuels began to recover around 450-650°C with one stage and was almost completely recovered by annealing at 850°C for 5 h. Based on the recovery of broadening of X-ray reflections and TEM observations, defect clusters of dislocations and small intragranular bubbles began to recover around 1.150–1,450°C. Complete recovery of the defect clusters, however, was not found even after annealing at 1,800°C for 5h. The effect of irradiation temperature on microstructural change of sub-grain structure in high burnup fuels was assessed from the experimental results.     

Elevated-temperature tensile properties of irradiated 9 Cr-1 MoVNb steel

Normalized-and-tempered 9 Cr-1 MoVNb steel tensile specimens were irradiated in the Experimental Breeder Reactor-11 (EBR-11) at 390, 450, 500, and 550°C to ～2.1 and 2.5 × 10²⁶ neutrons/m² (> 0.1 MeV), which produced displacement damage levels of ～10 and 12 dpa, respectively. Tensile tests were conducted at the irradiation temperature and at room temperature. In addition to the irradiated specimens, as-heat-treated specimens and as-heat-treated specimens thermally aged at the irradiation for 5000 h were also tested.Thermal aging had no effect on the unirradiated tensile properties. Irradiation at 390°C increased the 0.2% yield stress and the ultimate tensile strength above those of the unirradiated control specimens. The ductility decreased slightly. After irradiation at 450, 500, and 550°C, the tensile properties were essentially the same as the unirradiated values. The hardening at 390°C was attributed to the dislocation and precipitate structure formed during the irradiation. The lack of hardening at 450°C and higher correlates with an absence of an irradiation-induced damage structure.     

Void formation in solution-treated aisi 316 and 321 stainless steels under 46.5 mev ni6+ irradiation

The swelling and radiation damage structure developed in solution-treated 316 and 321 stainless steels bombarded by 46.5 MeV Ni⁶⁺ ions in the Variable Energy Cyclotron (VEC) have been determined. Foils were pre-injected with 10^?5 a/a He at room temperature and subsequently bombarded by Ni⁶⁺ ions in the temperature range 450–750°C at a damage rate of 1–3 × 10^?3 dpa per second to doses up to 300 dpa and specimens from the foils were examined by transmission electron microscopy. The data obtained were compared with data from other experiments aimed at simulating the fast-neutron irradiation of 316 and 321 steels, in particular previous work with 20 MeV C²⁺ ions and with data on fast-reactor bombarded material. The swelling rates in Ni-ion bombarded specimens were about a factor two less than those in C-ion bombarded specimens and in good agreement with swelling rates in 5 MeV Ni⁺- and neutron-bombarded material. The peak swelling temperature after a dose of 40 dpa was 650°C in 316 steel and 625°C in 321 steel where the swelling was about 5.8% and 4.6% respectively.     

Effects of Annealing under High Static Pressure on Precipitation Behavior of Fission Gas Bubbles in Irradiated UO2

Annealing experiments were carried out on irradiated UO₂ in argon gas under high pressure (600 and 1,000 kg/cm²) as well as atmospheric, at temperatures of 1,400°–1,600°C. The effects of high external pressure on the behavior of fission gas bubbles in the irradiated UO₂ were studied by comparing replica electron micrographs of fractured surfaces of specimens annealed under different temperatures and pressures. The results indicate that high pressures such as above 600 kg/cm² can be effective in surpressing the growth of fission gas bubbles in both intergranular and intragranular zones, and in inhibiting the joining together of intergranular bubbles to form direct passages for fission gas release.     

Helium bubbles and trace of lithium in B4C control rod pellets used in JOYO experimental fast reactor

B₄C pellets used in the control rod of the experimental fast reactor ‘JOYO’ with different ¹⁰B burnups from lower than 10 × 10²⁰ captures/cm³ to 80 × 10²⁰ captures/cm³ and irradiated at less than 800 °C were examined by transmission electron microscopy (TEM). In a B₄C pellet irradiated in an irradiation capsule of ‘JOYO’ at 800 °C up to 30 × 10²⁰ captures/cm³, intragranular helium bubbles appeared in flat plate-shapes with the plane of the plate parallel to the (111) rhombohedral plane. However, in the other specimens that were taken from an actual control rod, the helium gas formed very tiny spherical intergranular bubbles with a diameter of a few nanometers . These tiny bubbles make wavy arrays roughly parallel to the (111) plane. The B₄C specimens were heated on a TEM in situ heating holder up to 1040 °C for 10 min. Clustering of tiny bubbles was observed, but did not extend to the plate-shaped bubbles. In high burnup specimens, large bubbles/cracks were rarely found along the {100} planes, which may correspond to the amorphous bands caused by the slip. While heating the specimens in TEM over 800 °C, liquid phases of lithium-bearing compoundsappeared on the surface of specimen.     

Effects of Temperature Cycling and Heating Rate on Fission Gas Release of BWR Fuels

The effects of temperature cycling and heating rate on the release behavior of ⁸⁵Kr have been studied for U0₂ pellets irradiated in a commercial BWR during 3 and 4 cycles (burn-up: 23 and 28GWd/t), by using a post irradiation annealing technique. In addition, characteristics of intergranular bubbles in base-irradiated and annealed specimens (burn-up: 6~28GWd/t) have been examined by SEM fractography.

No significant difference in the release of ⁸⁵Kr was observed between the cyclic heating from 700 to 1,400°C and isothermal heating at 1,400°C. The maximum release rate of ⁸⁵Kr during heating up to 1,800°C became lower with decreasing heating rate in the range of 0.03–10°C/s, while its cumulative fractional releases were about 20~30%, almost independent of heating rate. The fractional coverage of the grain face area occupied by intergranular bubbles saturated around 40~50 for the specimens annealed at 1,600-1,800°C, independent of specimen burn-up and heating conditions (temperature, heating rate and duration). A relationship between intergranular bubble concentration N_g per unit area of grain face and average bubble diameter d_g was expressed as N_g∝d_{g 2.1}     

Correlation between blister skin thickness,the maximum in the damage-energy distribution,and projected ranges of helium ions in Nb for the energy range 10–1500 keV

The skin thickness of blisters formed on polycrystalline niobium under irradiation at room temperature by ⁴He⁺ at energies of 15 to 80 keV have been measured. Similar measurements were conducted for 10-keV ⁴He⁺ irradiation at 500°C to increase blister exfoliation, and thereby allow examination of a larger number of blister skins. For energies less than 100 keV the skin thicknesses are compared with both the projected range and the damage-energy distributions constructed from moments interpolated from Winterbon's tabulated values. The projected ranges and damage-energy distributions have also been computed with a Monte-Carlo program. For energies greater than 100 keV the projected ranges of ⁴He⁺in Nb were calculated using either Brice's formalism or the one given by Schiøtt. The thicknesses of blister skins for 60 and 80-keV irradiations, and those reported earlier for 100 to 1500-keV irradiations correlate well with calculated projected ranges. For irradiation energies less than 60 keV the measured thicknesses are greater than the calculated ranges.     

Quantifying helium distribution in dual-ion beam irradiated SiCf/SiC composites by electron energy loss spectroscopy

In this study, we report a method to quantify the helium distribution in the SiC_f/SiC composites, which are used as the first-wall materials of fusion reactor. The helium-bubble formation in Hi-Nicalon Type-S (HNS) was observed in the irradiated SiC_f/SiC composites at a level of 100 dpa and at 800 °C and 1000 °C, respectively. We applied transmission electron microscopy and electron energy loss spectroscopy to investigate the helium-gas-bubbles-formation mechanisms. To simulate the practical first-wall environment of Deuterium–Tritium (D–T) fusion reactor, a dual-ion beam (6 MeV Si³⁺ and 1.13 MeV He⁺) was performed to irradiate the SiC_f/SiC composites. The relationship between the energy shift of He K-edge and the radius of the bubble of the SiC composites was estimated by electron energy loss spectroscopy analysis. The results show that all of the helium atoms irradiated at 1000 °C and formed the bubbles. On the other hand, at 800 °C, only 25.5% of the helium atoms form the helium bubbles. A clear thermal-dependent formation mechanism is found.     

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.     

Corrosion characteristics of Hastelloy N alloy after He+ ion irradiation

With the goal of understanding the invalidation problem of irradiated Hastelloy N alloy under the condition of intense irradiation and severe corrosion, the corrosion behavior of the alloy after He⁺ ion irradiation was investigated in molten fluoride salt at 700 °C for 500 h. The virgin samples were irradiated by 4.5 MeV He⁺ ions at room temperature. First, the virgin and irradiated samples were studied using positron annihilation lifetime spectroscopy (PALS) to analyze the influence of irradiation dose on the vacancies. The PALS results showed that He⁺ ion irradiation changed the size and concentration of the vacancies which seriously affected the corrosion resistance of the alloy. Second, the corroded samples were analyzed using synchrotron radiation micro-focused X-ray fluorescence, which indicated that the corrosion was mainly due to the dealloying of alloying element Cr in the matrix. Results from weight-loss measurement showed that the corrosion generally correlated with the irradiation dose of the alloy.     

Synergistic effect of helium and hydrogen for bubble swelling in reduced-activation ferritic/martensitic steel under sequential helium and hydrogen irradiation at different temperatures

In order to investigate the synergistic effect of helium and hydrogen on swelling in reduced-activation ferritic/martensitic (RAFM) steel, specimens were separately irradiated by single He⁺ beam and sequential He⁺ and H⁺ beams at different temperatures from 250 to 650 °C. Transmission electron microscope observation showed that implantation of hydrogen into the specimens pre-irradiated by helium can result in obvious enhancement of bubble size and swelling rate which can be regarded as a consequence of hydrogen being trapped by helium bubbles. But when temperature increased, Ostwald ripening mechanism would become dominant, besides, too large a bubble could become mobile and swallow many tiny bubbles on their way moving, reducing bubble number density. And these effects were most remarkable at 450 °C which was the peak bubble swelling temperature for RAMF steel. When temperature was high enough, say above 450, point defects would become mobile and annihilate at dislocations or surface. As a consequence, helium could no longer effectively diffuse and clustering in materials and bubble formation was suppressed. When temperature was above 500, helium bubbles would become unstable and decompose or migrate out of surface. Finally no bubble was observed at 650 °C.     

Observation and analysis of damage structure in Al and Al/Mg (N4) alloy after irradiation with 100 and 400 keV aluminium ions

A comparative TEM study has been made of ion irradiation damage structure in pure aluminium, commercial aluminium (grade 1100) and in a modified N4 (Al/2.95% Mg) alloy of the type used in the construction of the calandria of the Winfrith prototype SGHW Reactor. Atom displacements equivalent to many years neutron irradiation were simulated by bombardment with 100 and 400 keV Al⁺ ions to doses of up to 200 dpa at temperatures between 30 and 250°C. Dynamic observations of damage formation were made during irradiation with 100 keV ions in a linked heavy-ion accelerator/200 keV electron microscope, and further results were obtained by 400 keV Al⁺ ion bombardment in a Cockcroft-Walton accelerator. Dislocation structure and voids were seen in aluminium irradiated with 100 and 400 keV A1⁺ at temperatures between 30 and 250°C. Void swelling of 8.7% at 104 dpa was a maximum at 1&#x0303;50°C in type 1100 aluminium. No voids were found at temperatures μ 250°C. No voids were seen in the Al/Mg N4 alloy after bombardments up to 200 dpa with 100 keV A1⁺, and 150 dpa with 400 keV Al⁺ at temperatures between 50 and 170°C. The void-resistant property is consistent with observations in the USA of neutron-irradiated 5052 Al alloy which has a similar magnesium content to the modified N4 alloy. The 1100 alloy and N4 results have been analysed using the rate theory of swelling. The absence of voids in the N4 alloy indicates an effective vacancy annihilation mechanism, which possibly occurs at small precipitates formed during irradiation.     

Ferrite formation in neutron-irradiated austenitic stainless steel

Magnetic measurements were carried out on type 316, 321 and three modified heats of 316 austenitic stainless steels that had been irradiated to high fluences (1 ? 8 × 10²²n/cm², E > 0.1 MeV) in EBR-II at temperatures ranging from 450–700°C. Most of the specimens showed increases of magnetization after exposure to the reactor environment that can be attributed to formation of numerous small ferrite particles. The amount of ferrite formed during irradiation is a function of alloy composition as well as irradiation temperature and fluence. Specimens with low molybdenum concentrations had a greater ferrite content than specimens with the normal molybdenum content of type 316 stainless steel. A modified heat of type 316 with 0.23 wt% Ti had lower levels of ferrite under given irradiation conditions than the other heats. Some particles with diffraction patterns corresponding to the ferrite phase were found in an irradiated type 321 stainless specimen, but none were observed in the type 316 stainless specimens.     

Defect- and strain-enhanced cavity formation and Au precipitation at nano-crystalline ZrO2/SiO2/Si interfaces

Defect- and strain-enhanced cavity formation and Au precipitation at the interfaces of a nano-crystalline ZrO₂/SiO₂/Si multilayer structure resulting from 2 MeV Au⁺ irradiation at temperatures of 160 and 400 K have been studied. Under irradiation, loss of oxygen is observed, and the nano-crystalline grains in the ZrO₂ layer increase in size. In addition, small cavities are observed at the ZrO₂/SiO₂ interface with the morphology of the cavities being dependent on the damage state of the underlying Si lattice. Elongated cavities are formed when crystallinity is still retained in the heavily-damaged Si substrate; however, the morphology of the cavities becomes spherical when the substrate is amorphized. With further irradiation, the cavities appear to become stabilized and begin to act as gettering sites for the Au. As the cavities become fully saturated with Au, the ZrO₂/SiO₂ interface then acts as a gettering site for the Au. Analysis of the results suggests that oxygen diffusion along the grain boundaries contributes to the growth of cavities and that oxygen within the cavities may affect the gettering of Au. Mechanisms of defect- and strain-enhanced cavity formation and Au precipitation at the interfaces will be discussed with focus on oxygen diffusion and vacancy accumulation, the role of the lattice strain on the morphology of the cavities, and the effect of the binding free energy of the cavities on the Au precipitation.     

Hardening and Recovery of Neutron-Irradiated Pressure Vessel Steel (ASTM A533B)

Specimens of ASTM A533B steel were studied to gain information on the annealing process following irradiation, through measurements of internal friction and of hardness.

The specimens were quenched from 900°C and tempered at 650°C, then irradiated in the JMTR reactor at 65°–75°C to a neutron dose of 1.4–1.7×10²⁰ n/cm² (E _n >1MeV).

Peaks were observed on the internal friction curves from unirradiated specimens. These peaks disappeared upon irradiation, but reappeared with annealing treatment at 150°C.

Radiation-anneal hardening was observed at 250°C. The recovery of radiation hardening begins at a temperature between 250° and 350°C, but is not completed even at 550°C.     

Recovery of Technetium with Active Carbon Column in Partitioning Process of High-Level Liquid Waste

Blistering and exfoliation of several tungsten alloys, which cause surface damage, were investigated using 3-MeV He-ion bombardment at room temperature (RT), 400, and 550°C. The alloy W-0.3TiC, which was fabricated by the mechanical alloying method and had an ultrafine grain structure, a K-doped W alloy, and pure W metal were examined to explore a way of suppressing the surface damage. In RT irradiation, surface exfoliation occurred at a fluence of (1–2) × 10²² He/m² in all the tested specimens. In the case of 550°C irradiation, surface exfoliation was observed above 2 × 10²² He/m² irradiation in pure W and K-doped W, but no surface exfoliation was observed in W-0.3TiC up to a fluence of 2 × 10²³ He/m². The results showed that W-0.3TiC showed a higher resistance to surface exfoliation by He-ion bombardment and the level of resistance was temperature-dependent. The surface morphology, cross-sectional morphology, and microstructure were characterized by transmission electron microscopy. Helium thermal desorption spectrometry was carried out to determine the mechanism whereby the surface attained resistance to the damage through He-ion bombardment. The improvement in the resistance to the surface exfoliation could be attributed to the ultrafine grain structure and the intergranular enhanced He diffusion behavior of the MA-processed material.