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Under transient conditions, rapid fuel swelling in UO2 can occur through fission gas and vacancy capture by intra-granular bubbles. These swellings can lead to significant cladding deformations and clad failure. To understand the mechanisms of intra-granular swelling a number of key experiments have been conducted by the UK in conjunction with the Halden Reactor. Post-irradiation annealing enables the study of the thermal behaviour of the intra-granular bubbles in the absence of complicating processes such as irradiation-induced bubble destruction and fission gas resolution. Bubble-like features are frequently observed in straight lines and it has always been assumed that these were nucleated in the wake of energetic fission fragments and served as nuclei for bubble growth. However, most of these bubbles do not grow significantly and viable intra-granular bubbles are nucleated at much lower densities. At the temperature and pressure that these form, the xenon and krypton are above their critical points and the elements are supercritical fluids. However, bubble growth occurs at temperatures and pressures much greater than the critical points and the gases may be treated with simple equations of state such as those by Clausius or Van der Waals. Of greater importance is the fact that the large Van der Waals forces in these atoms generated by fluctuations in the electron dispositions lead to extensive adsorption on the inner surfaces of the bubbles and vacancy capture by the bubbles can lead to the diffusive loss of fission gas atoms resulting in bubble growth at slight over-pressure conditions. A diffusion model based on Langmuir Adsorption has been developed and shown to be consistent with the out-of-pile annealing studies. [ABSTRACT FROM AUTHOR]

Under transient conditions, rapid fuel swelling in UO2 can occur through fission gas and vacancy capture by intra-granular bubbles. These swellings can lead to significant cladding deformations and clad failure. To understand the mechanisms of intra-granular swelling a number of key experiments have been conducted by the UK in conjunction with the Halden Reactor. Post-irradiation annealing enables the study of the thermal behaviour of the intra-granular bubbles in the absence of complicating processes such as irradiation-induced bubble destruction and fission gas resolution. Bubble-like features are frequently observed in straight lines and it has always been assumed that these were nucleated in the wake of energetic fission fragments and served as nuclei for bubble growth. However, most of these bubbles do not grow significantly and viable intra-granular bubbles are nucleated at much lower densities. At the temperature and pressure that these form, the xenon and krypton are above their critical points and the elements are supercritical fluids. However, bubble growth occurs at temperatures and pressures much greater than the critical points and the gases may be treated with simple equations of state such as those by Clausius or Van der Waals. Of greater importance is the fact that the large Van der Waals forces in these atoms generated by fluctuations in the electron dispositions lead to extensive adsorption on the inner surfaces of the bubbles and vacancy capture by the bubbles can lead to the diffusive loss of fission gas atoms resulting in bubble growth at slight over-pressure conditions. A diffusion model based on Langmuir Adsorption has been developed and shown to be consistent with the out-of-pile annealing studies. [ABSTRACT FROM AUTHOR]

Assessing the embrittlement and hardening of reactor pressure vessel steels is critical for the extension of the service lifetime of nuclear power plants. This paper summarises the tensile test results on the irradiation behaviour of realistic VVER-1000 welds from the STRUMAT-LTO project. The welds were irradiated at the HFR (Petten, the Netherlands) to a fluence of up to 1.087 × 1020 n·cm−2, and their irradiation hardening was studied by means of tensile testing. The four grades, with different Mn and Ni contents, show different hardening behaviours. The highest degree of irradiation hardening is observed for the weld that has the highest combined Ni + Mn content. The results show that there is a synergetic effect of Mn and Ni on the irradiation hardening behaviour of the VVER-1000 welds. Besides irradiation hardening, the effectiveness of post-irradiation annealing treatments on the recovery of the tensile properties is studied in the present work. Post-irradiation annealing treatments conducted at 418 °C and at 475 °C proved to be effective for three of the four investigated welds. For the realistic weld with the highest combined Ni + Mn, only the annealing at 475 °C led to the complete recovery of the tensile properties. [ABSTRACT FROM AUTHOR]

Aiming to newly-developed FeCrAl alloys for accident-tolerant fuels (ATF) in light water reactors (LWR), an in-situ investigation of scan transmission electron microscopy (STEM) observation and annealing was performed in order to understand the evolution of radiation-induced defects (RID) and the Cr segregation during elevating the temperature. In this study, the FeCrAl alloys were irradiated with 6Mev Au+ ions up to a dose of 10 dpa at 400 °C, and the RID in depth of 1 μm were assess by in-situ TEM during annealing. A Cr layer, the dislocation loop raft and Cr-decorated dislocation loops were formed in the near-surface region I. Cavities, Cr-rich precipitates were formed in the interior region II. By calculating thermal equilibrium vacancies theoretically, the cavity formation is ascribed to the surface effect of the foil and the damage of TEM electrons. Shrinkage of some loops could be ascribed to produced vacancies. Formations of the raft and Cr-decorated loops are due to the strain or stress constructed by the surface C species or damages when annealing below 500 °C. The pronounced Cr-rich precipitates during 550–600 °C are due to the Cr atomic segregation initiated by vacancies via a long-time recovery towards low-energy sites. This work indicates the defect-caused stress profile is important for the Cr segregation.

Assessing the embrittlement and hardening of reactor pressure vessel steels is critical for the extension of the service lifetime of nuclear power plants. This paper summarises the tensile test results on the irradiation behaviour of realistic VVER-1000 welds from the STRUMAT-LTO project. The welds were irradiated at the HFR (Petten, the Netherlands) to a fluence of up to 1.087 × 1020 n·cm−2, and their irradiation hardening was studied by means of tensile testing. The four grades, with different Mn and Ni contents, show different hardening behaviours. The highest degree of irradiation hardening is observed for the weld that has the highest combined Ni + Mn content. The results show that there is a synergetic effect of Mn and Ni on the irradiation hardening behaviour of the VVER-1000 welds. Besides irradiation hardening, the effectiveness of post-irradiation annealing treatments on the recovery of the tensile properties is studied in the present work. Post-irradiation annealing treatments conducted at 418 °C and at 475 °C proved to be effective for three of the four investigated welds. For the realistic weld with the highest combined Ni + Mn, only the annealing at 475 °C led to the complete recovery of the tensile properties.

Beryllium oxide (BeO) is being re-emphasized and utilized in Micro Modular Reactors (MMR) because of its prominent nuclear and high temperature properties in recent years. The implications of the research about effects of neutron irradiation on the microstructure and properties of BeO are significant. This article comprehensively reviews the effects of neutron irradiation on BeO and proposes the maximum permissible neutron doses at different temperatures for BeO without cracks in appearance according to the data in the previous literature. This maximum permissible neutron dose value has important reference significance for the experimental study of BeO. The effects of neutron irradiation on the thermal conductivity and flexural strength of BeO are also discussed. In addition, microstructure evolution of irradiated BeO during post-irradiation annealing is summarized. This review article has important implications for the application of BeO in MMR.

• Thermal diffusivity of tungsten irradiated in a spallation target up to 26.5 dpa was studied. • Irradiated tungsten showed a drastic reduction of the thermal diffusivity by up to 50 %. • Thermal diffusivity was mildly recovered after post irradiation annealing. • EDM cut sample surfaces are suitable for LFA measurements. Tungsten is chosen as the target material of European Spallation Source (ESS), where it will be irradiated by a high energy (2 GeV) and high power (5 MW) pulsed proton beam to produce neutrons to be used by neutron scattering intruments. For designing a target with a high availability, it is important to determine the evolution of the temperature and thermal stresses in tungsten during operation. The thermal stress in the target is mainly determined by temperature gradients which depend on thermal diffusivity of tungsten. Reduced thermal diffusivity results in increased beam-induced thermomechanical stress in the target, which affects the survivability and lifetime of the target. As irradiation alters material properties, knowledge on the impact of irradiation on thermal diffusivity is crucial to the ESS target design and operation. In this work, the Laser Flash Analysis (LFA) technique was used to investigate the thermal diffusivity of three highly irradiated samples with displacement damage doses 9.5, 25.1 and 26.5 dpa. Due to the high radioactivity of the specimens, no surface polishing or blackening could be applied on the samples prior to the LFA measurements. Therefore, an attempt was made to study the effect of measured surface roughness on LFA data and the obtained results for irradiated samples were calibrated accordingly. The data shows a drastic reduction of the thermal diffusivity for all three samples by up to 50 %. The thermal diffusivites of the three specimens are overlapping within measurement uncertainty ranges, and are independent of the damage dose. Further, the recovery of the thermal diffusivity of irradiated tungsten was also observed after annealing at 1000 °C for an hour. The results showed the annealing effect was less pronounced compared to specimens irradiated at lower dose and lower temperature. These phenomena are speculated to be owing to the formation of larger dislocation loops and defect clusters in the high dose samples during radiation, resulting in a smaller contribution to the recovery achieved by post irradiation annealing. [ABSTRACT FROM AUTHOR]

Single crystal W is offered for Physics understanding of irradiation induced defects and their annealing as its structure is well defined and it is almost defect free. W(100) single crystal was neutron irradiated to a damage of 0.11 displacements per atom at 600 °C and subsequently isochronally annealed from 700 up to 1500 °C in 100 °C steps. Irradiation causes the formation of dislocation loops and vacancy clusters and a 45 % increase in hardness. After the annealing of 1500 °C the positron lifetime annihilation spectrum shows a defect free material and its hardness has been reverted to the pre-irradiation value and only clusters of Re, WRe and WOs2 have been detected by grazing incidence X-ray diffraction. The total line density of dislocations, number density of voids and their size versus annealing temperature have been determined. From hardness, the critical resolved stresses arising from dislocations and voids have been derived and correlated with their densities. The kinetics of defect annihilation versus annealing temperature is discussed.

The study of the recrystallization resistance of engineering grade tungsten (W) prior to and after neutron irradiation as well as subsequent thermal annealing provides valuable insight into neutron induced defect interactions and their kinetics as well as their correlation with recrystallization mechanisms. The outcome of such investigations would be of assistance for appropriate grade selection for a fusion reactor and the design of healing processes enabling the lifetime extension of fusion reactor components. Within this framework, samples from ITER specification W, forged bar and cold-rolled plate, were neutron irradiated to a dose of 0.18 dpa at 600 °C in the Material Test Reactor BR2, Mol, Belgium, and subsequently isochronally annealed from 700 to 1550 °C for 24 h with a 50 °C step with two non–irradiated counterparts. X-ray diffraction, optical microscopy and positron annihilation lifetime spectroscopy, combined with Vickers Hardness are used to correlate the recrystallization process and recovery mechanisms. It is found that the non-irradiated plate has a higher recrystallization resistance compared to that of the bar, with the former requiring 100 °C higher temperature for 24 h annealing time to recrystallize. Despite the irradiation induced dislocation loops and voids being of similar sizes and densities for both grades, the different initial microstructure affects the recrystallization process upon annealing, delaying it by 50 °C on the plate. Further in the plate a complete void dissolution is observed before recrystallization, which is not the case for the bar, for which the two effects occur at the same temperature. • Isochronal annealing up to 1550 °C of two neutron irradiated ITER specification tungsten. • Higher recrystallization resistance of cold-rolled tungsten compared to forged bar. • Recovery stages identified including dislocation annihilation and void dissolution. • Complete void dissolution occurs before recrystallization in cold-rolled tungsten. [ABSTRACT FROM AUTHOR]

The post-irradiation anneal tests for dispersion fuel plates are commonly employed to determine the blistering threshold temperature in order to evaluate their resistance to accidents. In this study, a three-dimensional simulation method for UMo/Zr dispersion fuel plates is developed to predict the normal stresses in the fuel skeleton and some related variables during in-pile irradiation and out-of-pile anneal tests. The effects of the intrusion of extra fission gas atoms into gas bubbles are investigated, and the contributions of the creep performances of fuel particles and the matrix are examined. The predictions reveal the following insights: (1) the thermo-mechanical behaviors during anneal tests are strongly dependent on the intrusion of extra fission gas atoms into gas bubbles and the additional fission-gas induced swelling during post-irradiation anneal tests; the maximum tensile stress in the fuel skeleton increases significantly, from 57.4 MPa to 346.0 MPa, as the fraction of generated fission gas atoms in gas bubbles shifts from approximately 23.4 % to 100 % in the annealing process; (2) the thermo-mechanical responses during anneal tests are greatly influenced by the creep performances of fuel particles and the matrix; the significantly weakened creep ability of fuel particles is found to be the critical factors responsible for particle cracking in the annealing process; the fuel particle fracture will hardly occur when the irradiation creep rate model is maintained during the considered annealing scheme, because the skeleton tensile stress does not appear. It is important to note that the particle cracking and the subsequent blister behaviors under in-pile neutron irradiation conditions are possible to occur if a rapid temperature increase takes place under accident conditions. Consequently, the out-of-pile annealing tests are of significance. [ABSTRACT FROM AUTHOR]

Significant embrittlement in reactor pressure vessel (RPV) steels can be caused by the formation of nanometer-scale Mn–Ni–Si precipitates (MNSPs) and annealing is a promising technique for reducing embrittlement. To achieve better understanding of the evolution of these precipitates at the atomic scale, a kinetic lattice Monte Carlo (KLMC) model, parameterized using CALPHAD and recent atom probe tomography (APT) data, is used to simulate post-irradiation annealing of MNSPs. The model predicts MNSP volume fractions, number densities and sizes that agree well with the experimental observations. The model also predicts that the initial structure of the precipitates may be B2 bcc phases with one sublattice occupied by Ni and the other sublattice occupied by Mn and Si, as well as shows a modest temperature dependence of the MNSP composition. The results show that the simple approach can be used to model MNSP evolution and supports that these precipitates are stable thermodynamic phases. [ABSTRACT FROM AUTHOR]

This work shows an ongoing study of the influence of neutron irradiation on the hardness of first wall structure materials. EUROFER97 and boron-doped steels have been neutron-irradiated in the European High Flux Reactor (HFR) with an average dose of up to 16 dpa at temperatures between 250 and 450 °C. The irradiated samples are investigated by the advanced method of a registering hardness measurement that is used to perform experiments at temperatures of up to 450 °C. The first part of this work examines the hardening of EUROFER97 depending on the irradiation temperature at different test temperatures. In addition, post-irradiation annealing is applied to EUROFER97 and to the boron-doped steels. The different contents of boron lead to different amounts of helium generated by neutron-irradiation. Since post-irradiation annealing is known to enable the neutron-induced Frenkel pairs and nanoscale defects to recombine, the remaining hardening after the annealing process can be attributed mostly to products of transmutation reactions. Thus, the potential as well as the limits of this promising treatment are explored. [ABSTRACT FROM AUTHOR]

The effect of isochronal annealing on the evolution of dislocation loop and void population in proton irradiated ThO 2 has been investigated. Post-irradiation annealing in other actinide oxides like UO 2 shows significant loop coarsening. ThO 2 samples were irradiated with 2 MeV protons up to a dose of 0.1 dpa at 600 °C. Post-irradiation isochronal annealing was performed at 600, 800, 1000 and 1100 °C for 1 h at each temperature using in-situ TEM. Only faulted 1/3<111> type dislocation loops were observed, and their sizes and distribution were characterized. The population of self-interstitial atom (SIA) dislocation loops did not show any significant growth and coarsening. Additionally, nanometric voids were observed at annealing temperatures of 1000 and 1100 °C. Using cluster dynamics (CD), we have studied the nucleation and growth of point defects and defect clusters, i.e., SIA prismatic dislocation loops and nanometric and sub-nanometric voids in proton irradiated ThO 2. The CD model was further utilized to predict the growth and coarsening of loops and voids during isochronal annealing at the experimental and higher temperatures. The model did not predict significant SIA loop growth which closely corresponds to the TEM observations. CD predicted SIA loop coarsening is insignificant even at high annealing temperature of 1500 °C because the model only considers the growth of defect clusters by absorption of like point defects, i.e., SIA loops absorb interstitials and voids absorb vacancies, and cannot account for their migration and coalescence due to elastic interaction. The CD model also predicts the evolution of nanometric voids having mean size within the error bounds of TEM observations. [ABSTRACT FROM AUTHOR]

To study the nature of irradiation-induced nanofeatures in oxide dispersion strengthened (ODS) Fe-Cr alloys, post-irradiation isochronal thermal annealing up to 600°C was performed for ODS Fe-9%Cr and Fe-14%Cr alloys ion-irradiated at 300°C and 500°C. Nanoindentation indicated hardening for all as-irradiated alloys and complete hardness recovery upon post-irradiation annealing. Cross-sectional TEM indicated an irradiation-induced defect band near peak damage mainly consisting of dislocation loops. Candidate mechanisms of recovery were critically evaluated. Shrinkage of loops via capture of thermal vacancies was found to correctly reflect the annealing behaviour of ODS Fe-9Cr irradiated at 300°C. [ABSTRACT FROM AUTHOR]

Abstract—The influence of neutron irradiation, post-irradiation annealing, and re-irradiation on the fracture toughness and mechanical properties of anticorrosive cladding material for reactor pressure vessels was investigated. Dependences for predicting the deformation diagram and fracture toughness of the weld metal, with the consideration of the effect of annealing and repeated neutron irradiation were proposed. The obtained curves may be used to calculate the stress-strain state of reactor pressure vessels, as well as to assess the brittle fracture of the material during operation of the reactor vessel after annealing. [ABSTRACT FROM AUTHOR]

Abstract—The influence of neutron irradiation and post-irradiation annealing on the fracture toughness of anticorrosive cladding material for reactor pressure vessels with different phosphorus and δ-ferrite content was investigated. The effect of the annealing temperature on the degree of degradation and recovery of the weld metal was considered. The processes during annealing exerting both positive and negative effects on recovery of the cladding properties were studied. The results of the research can be used for justification of the resistance to brittle fracture of reactor pressure vessels after annealing. [ABSTRACT FROM AUTHOR]

Microstructural changes resulted from neutron irradiation and post-irradiation annealing in a high-temperature ultra-fine precipitate strengthened (HT-UPS) stainless steel were characterized using transmission electron microscopy (TEM) and atom probe tomography (APT). Three HT-UPS samples were neutron-irradiated to 3 dpa at 500 °C, and after irradiation, two of them were annealed for 1 h at 600 °C and 700 °C, respectively. Frank dislocation loops were the dominant defect structure in both the as-irradiated and 600 °C post-irradiation-annealed (PIAed) samples, and the loop sizes and densities were similar in these two samples. Unfaulted dislocation loops were observed in the 700 °C PIAed sample, and the loop density was greatly reduced in comparison with that in the as-irradiated sample. Nano-sized MX precipitates were observed under TEM in the 700 °C PIAed sample, but not in the 600 °C PIAed or the as-irradiated samples. The titanium-rich clusters were identified in all three samples using APT. The post-irradiation annealing (PIA) caused the growth of the Ti-rich clusters with a stronger effect at 700 °C than at 600 °C. The irradiation caused elemental segregations at the grain boundary and the grain interior, and the grain boundary segregation behavior is consistent with observations in other irradiated austenitic steels. APT results showed that PIA reduced the magnitude of irradiation induced segregations. [ABSTRACT FROM AUTHOR]

Post-irradiation annealing was performed on a 304L SS that was irradiated to 5.9 dpa in the Barsebäck 1 BWR reactor. Evolution of dislocation loops, radiation-induced solute clusters and radiation-induced segregation at the grain boundary was investigated following thermal annealing at 500 °C and 550 °C up to 20 h. Dislocation loops, Ni-Si and Al-Cu clusters, and enrichment of Ni, Si and depletion of Cr at the grain boundary were observed in the as-irradiated condition. Dislocation loop size did not change significantly after annealing at 550 °C for 5 h but the loop number density decreased considerably and loops mostly disappeared after annealing at 550 °C for 20 h. The average size of Ni-Si and Al-Cu clusters increased while the number density decreased with annealing. The increase in cluster size was due to diffusion of solutes rather than cluster coarsening. Significant volume fractions of Ni-Si and Al-Cu clusters still remained after annealing at 550 °C for 20 h. Substantial recovery of Cr and Ni at the grain boundary was observed after annealing at 550 °C for 5 h but neither Cr nor Ni was fully recovered after 20 h. Annihilation of dislocation loops, driven by the thermal vacancy concentration gradient caused by the strain field and stacking fault associated with the loops appeared to be faster than annihilation of solute clusters and recovery of Ni and Si at the grain boundary, both of which are driven by the solute concentration gradients. [ABSTRACT FROM AUTHOR]

Former investigations clearly revealed that embrittlement and hardening of reduced-activation ferritic-martensitic (RAFM) steels after neutron irradiation can be reduced remarkably by post-irradiation annealing (PIA). This study demonstrates the repeatability of damage and recovery mechanisms for EUROFER-based and EUROFER-ODS-HIP-steels by using specimens of previous 330 °C Bor 60 reactor (SSC RIAR) irradiations. Specimens were investigated in the irradiated state after a first PIA, after a subsequent further irradiation, and after a second annealing. Both tensile and Charpy-V impact tests clearly indicate that embrittlement occurs after the second irradiation of intermediately annealed (and, thus, recovered) material in the same order as after the first irradiation. In addition, a subsequent PIA leads to a complete recovery of the newly applied neutron damage, suggesting that lifetime of neutron-exposed fusion power plant components could be extended substantially by means of intermediate annealing. [ABSTRACT FROM AUTHOR]

• Coarsened voids & loops after post-irradiation annealing at 800 & 1000 °C. • Reduced voids and loop number densities after post-irradiation annealing at 1000 °C. • Reduced DC resistivity & hardness after post-irradiation annealing at 800 & 1000 °C. • Total dislocation line density determination from DC resistivity measurements. • Hardening after irradiation and post-irradiation annealing is dominated by voids. The microstructural evolution of neutron irradiated tungsten (W) after annealing and its correlation with the corresponding mechanical properties provides valuable insight on the defect interactions and their annihilation processes. This would result in the identification of recovery mechanisms, leading to the design of healing processes and thus, enabling the lifetime extension of fusion reactor components. Within this framework, samples from ITER specification forged W bar were neutron irradiated to a dose of 0.2 displacements per atom (dpa) at 600 °C in the Belgian Material Test Reactor (BR2) and subsequently annealed at 800 and 1000 °C. The evolution of the irradiation induced defects after annealing has been assessed by transmission electron microscopy (TEM), positron annihilation lifetime spectroscopy and electrical resistivity and its effect on the mechanical properties are discussed in terms of Vickers hardness. Neutron irradiation results in the formation of dislocation loops and voids. TEM observations show an increase in the size of both defect types after post irradiation (PI) annealing at both temperatures, accompanied by an initial increase in their density after PI annealing at 800 °C, followed by a subsequent decrease after PI annealing at 1000 °C. The presence of TEM-irresolvable defects of both types is revealed in the as-irradiated state, which is evidenced by the evolution of Vickers hardness, resistivity and positron lifetimes and their correlation after the PI annealing. In the as-irradiated state, as well as after PI annealing at both temperatures, the hardening is dominated by voids. [ABSTRACT FROM AUTHOR]

Severe plastic deformation, intense single-beam He-ion irradiation and post-irradiation annealing were performed on a nanostructured reduced activation ferritic/martensitic (RAFM) steel to investigate the effect of grain boundaries (GBs) on its microstructure evolution during these processes. A surface layer with a depth-dependent nanocrystalline (NC) microstructure was prepared in the RAFM steel using surface mechanical attrition treatment (SMAT). Microstructure evolution after helium (He) irradiation (24.8 dpa) at room temperature and after post-irradiation annealing was investigated using Transmission Electron Microscopy (TEM). Experimental observation shows that GBs play an important role during both the irradiation and the post-irradiation annealing process. He bubbles are preferentially trapped at GBs/interfaces during irradiation and cavities with large sizes are also preferentially trapped at GBs/interfaces during post-irradiation annealing, but void denuded zones (VDZs) near GBs could not be unambiguously observed. Compared with cavities at GBs and within larger grains, cavities with smaller size and higher density are found in smaller grains. The average size of cavities increases rapidly with the increase of time during post-irradiation annealing at 823 K. Cavities with a large size are observed just after annealing for 5 min, although many of the cavities with small sizes also exist after annealing for 240 min. The potential mechanism of cavity growth behavior during post-irradiation annealing is also discussed. [ABSTRACT FROM AUTHOR]

Abstract: Fission gas bubbles are one of the evolving microstructures that affect thermal mechanical properties, such as thermal conductivity, gas release, volume swelling, and cracking, in operating nuclear fuels. Therefore, fundamental understanding of gas bubble evolution kinetics is essential to predict the thermodynamic property and performance changes of fuels. In this work, a generic phase-field model was developed to describe the evolution kinetics of intragranular fission gas bubbles in UO2 fuels under post-irradiation thermal annealing conditions. Free energy functional and model parameters are evaluated from atomistic simulations and experiments. Critical nucleus size of gas bubbles and gas bubble evolution were simulated. A linear relationship between logarithmic bubble number density and logarithmic mean bubble diameter was predicted, which is in good agreement with experimental data. [Copyright &y& Elsevier]

The present paper investigates the tensile properties of post-irradiation annealed Cu-Ni alloy. The specimens were irradiated with a 15 MeV electron beam at room temperature and the post-irradiation annealing (PIA) of the specimens was carried out under vacuum at 450 °C for 15-120 min. The yield stress (YS), ultimate tensile stress (UTS), percentage elongation, stress relaxation rate and activation volume of both as-irradiated and post-irradiation annealed specimens were examined at room temperature using a universal testing machine. The results show that PIA of the specimen at 450 °C for 15 min decreases its YS and UTS, whereas the percentage elongation is increased. The changes in the tensile parameters become more pronounced with increases in annealing time. Effects of PIA on the stress relaxation rate and activation volume indicate that the relaxation rate of post-irradiation annealed specimens increases, and the activation volume decreases, with an increase in annealing time. [ABSTRACT FROM AUTHOR]

This paper summarises the review of 92 papers most of them selected from the ODIN database made within the framework of the project 'Nuclear Knowledge Preservation & Consolidation, WWER - WS2' managed by JRC EC. The papers contain the experimental results concerning postirradiation annealing and re-irradiation of WWER RPV materials. The results published in these papers have been obtained in RRC KI, VTT, NRI Rez, Forschungszentrum Dresden-Rossendorf, with the international cooperation between RRC KI, VTT-Fortum, JRC EC and ITEP, RRC KI and ORNL, and IAEA International Project. The following problems are examined in the review: evolution of RPV materials properties and structure under irradiation; recovery of irradiated RPV materials properties and structure under annealing; chemical composition effect on transition temperature of irradiated RPV materials recovery; effect of fluence on WWER RPV materials mechanical properties recovery due to annealing; microstructure evolution due to postirradiation annealing; effect of post-irradiation annealing on mechanical properties of irradiated WWER-1000 RPV materials; re-irradiation behaviour; surveillance programme for the post-annealing operation; assessment of the RPV state; technical problems concerning annealing; and licensing problems for annealing. [ABSTRACT FROM AUTHOR]

Thermal evolution and elevated temperature annealing of the dislocation microstructure of an irradiated metal, represented by an ensemble of elastically interacting interstitial dislocation loops, is explored using discrete dislocation dynamics simulations. The two fundamental microscopic processes driving the evolution of dislocations are the pipe diffusion of atoms along the dislocation lines, giving rise to the dislocation self-climb, and bulk diffusion of vacancies, resulting in the conventional dislocation climb. Simulations show that the coalescence and coarsening of the prismatic dislocation loop microstructure, observed at lower temperatures, is driven primarily by the dislocation self-climb. In tungsten, dislocation self-climb gives rise to a pronounced change in the dislocation loop microstructure at temperatures close to 800 ° C , see Ferroni et al. (2015) [1], whereas a similar microstructural transformation of the dislocation network driven by self-climb in α -iron is predicted to occur at ∼ 270 ° C. Simulations also show that the diffusion of vacancies in the crystal bulk is able to explain the observed annihilation rates of interstitial loops in tungsten. [ABSTRACT FROM AUTHOR]

Post-irradiation annealing (PIA) is beneficial in recovering degraded mechanical properties of irradiated materials due to the evolution of irradiation-induced defects. To understand the underlying responsible mechanisms, isothermal PIA experiments were carried out in this work on a neutron-irradiated 9%Cr reduced-activation ferritic-martensitic steel (EUROFER97) using in-situ transmission electron microscopy. In general, PIA at 600 °C led to an appreciable reduction of the dislocation loop density and increment in the loops mean size. This is found mainly associated with the vacancy-mediated loops shrinkage phenomenon. The driving force for this process is loop's curvature force. Nevertheless, the spontaneous evolution of the loops neighboring microstructure substantially alters their annealing kinetics. Therefore, in this study, loops' annealing kinetics in EUROFER97 is discussed in terms of various scenarios originating in terms of the presence or absence of the local external sources and sinks. In addition, influence of loops type, alloying elements segregation and free surface are also discussed. [ABSTRACT FROM AUTHOR]

Neutron-irradiation induced prismatic dislocation loops contribute to the degradation of materials' mechanical properties. Recent post-irradiation annealing (PIA) investigations on neutron-irradiated reduced-activation ferritic-martensitic (RAFM) steel "EUROFER97" have revealed an almost complete recovery of the mechanical properties. To identify underlying recovery mechanisms, thick-foil annealing experiments on neutron-irradiated EUROFER97 were carried out. It is observed that PIA at 550 °C results in a reduction of initial dislocation loops density with a simultaneous increase in their mean size. This is linked to the dislocation loops shrinkage/annihilation as well as their coarsening/merging phenomena. Moreover, a clear coarsening of the quasi-circular 〈100〉 loops into their rectilinear versions and simultaneous shrinkage/annihilation of the ½<111> loops/clusters are observed. The coarsening phenomenon is a non-conservative climb process, which is discussed in terms of the movement of jogs whilst considering the change in the iron's elastic anisotropy with temperature and loops lowest energy directions. Furthermore, the observed interaction of the coarsened 〈100〉 loops and thereafter their merging to form large loops of various geometries are presented and discussed. Lastly, a rationalization justifying pronounced microstructural transition from dislocation loops dominated to dislocation tangles dominated microstructure with continued annealing is presented. These thick-foil PIA results complement our previous works and provide complete description of the dislocation loops annealing behaviour in neutron-irradiated EUROFER97. [ABSTRACT FROM AUTHOR]

Recent post-irradiation annealing (PIA) investigations on the neutron-irradiated reduced-activation ferritic-martensitic (RAFM) steel EUROFER97 have led to an almost complete recovery of the mechanical properties. To identify responsible microstructural processes, transmission electron microscopy (TEM) investigations on neutron-irradiated (330 °C, 15 dpa) and subsequently annealed (550 °C, 3 h) EUROFER97 samples have been carried out. It is observed that PIA results in a reduction of dislocation loops density with a simultaneous increase in their mean size and formation of dislocation lines/networks. Additionally, a reduction of cavities density with increment in their mean size was observed. The loops' annealing behaviour appears to be vacancy/interstitial diffusion driven and is situated between the two extremes of isolated loops shrinkage/annihilation and their coarsening. Out of the two phenomena, the one that transpires depend on the presence or absence of the external sources and sinks in the loops neighbourhood. These competing mechanisms define the overall loops size distribution evolution upon PIA. Additionally, vacancy or their cluster migration and coalescence could be the main mechanism for cavities coarsening. The observed microstructural recovery upon PIA explains the recovery of mechanical properties by healing of irradiation induced defects. [ABSTRACT FROM AUTHOR]

This paper investigated the formation of bubbles and precipitates and the hardening effects in V–4Cr–4Ti alloy after irradiation of He and H ions and post-irradiation annealing. Microstructures observation shows that small bubbles formed in as-irradiated samples, and, during the post-irradiation annealing, lath-like precipitates formed and small bubbles grew upon. It was found that bubbles surrounding precipitates were significantly smaller than those away from, and tended to align along the precipitates. Composition analysis shows that the precipitates comprised TiO with FCC structure. The effects of bubbles and precipitates on hardening were evaluated by the nanoindentation test, as well as the dispersed barrier model that explicitly considered the small bubbles in as-irradiated sample, large bubbles and TiO precipitates in post-irradiated annealed samples. Compared with the as-irradiated samples, the higher hardness in post-irradiated annealed samples is attributed to the co-existence of bubbles and precipitates. [ABSTRACT FROM AUTHOR]

Post-irradiation annealing is an effective way to recover mechanical properties by removing neutron damage. Using in-situ transmission electron microscopy, a neutron irradiated Fe–9Cr alloy was annealed under specific time-temperature scheme. In general, dislocation loops density decreased and their mean size increased upon increasing annealing temperature and time. Concurrent shrinkage phenomenon was observed for both ½<111> and <100> type loops and is found to be vacancy diffusion driven. Essentially, loops' neighboring microstructure dictates their shrinkage behavior. Shrinkage rate is both loop size and type dependent. Loop size defined vacancy concentration gradient and loop type associated elemental segregation are main deciding factors. Image 1 [ABSTRACT FROM AUTHOR]

IG-430, a fine-grained, isotropic graphite grade is a promising candidate for the future Very High Temperature Reactors (VHTR). IG-430 which provides higher density, strength, and thermal conductivity, has already been developed as a graphite for next-generation HTGR, and is expected to be employed. This graphite grade, however, is lacking enough database that is needed for design. The present study aims to enhance the database with experimental data focusing on the low temperature regime (90–210 °C) by using 120–200 MeV protons to irradiate the IG-430 graphite to peak fluence of ~1.2 1025 m−2. It is anticipated that radiation-induced changes in the graphite properties and damage to be more pronounced in this low temperature regime than in elevated temperatures where damage annealing is taking place simultaneously. IG-430 graphite was characterized following irradiation for mechanical property changes (modulus and strength), dimensional stability and irradiation-induced growth as well as microstructural changes using high energy X-rays and different X-ray diffraction techniques. In assessing proton irradiation effects on the IG-430 graphite grade, comparison of radiation effects was made with the IG-43 grade, the un-purified version of IG-430, as well as other isotropic graphite grades. IG-430 was shown in this study to be better graphitized than other isotropic graphite grades. The study also revealed that during proton irradiation at low temperatures (~100 °C) the IG-430 exhibits stored energy release. [ABSTRACT FROM AUTHOR]

Post-irradiation annealing (PIA) was conducted on a 304L stainless steel irradiated to 5.9 dpa in the Barsebäck-1 BWR reactor, to investigate its effect on the mitigation of irradiation-assisted stress corrosion cracking (IASCC) susceptibility. IASCC susceptibility was measured for the as-irradiated and four PIA conditions (500 °C: 1 h and 550 °C: 1, 5, and 20 h) via interrupted constant extension rate tensile and four-point bend experiments under simulated BWR-NWC conditions. The annealing treatments were observed to progressively reduce IASCC susceptibility, as measured by the final intergranular fracture fraction (tensile) and crack length per unit area (four-point bend), with full removal of IASCC susceptibility being observed following annealing at 550 °C: 1 h for tensile specimens and 500 °C: 1h for four-point bend specimens. Among the microstructure and mechanical property parameters measured as a function of PIA, the average dislocation channel spacing was observed to decrease by ∼25% and ∼40% from the as-irradiated condition after annealing at 500 °C: 1 h and 550 °C: 1 h, respectively. The mitigation of IASCC susceptibility correlated well with the decrease in the average dislocation channel spacing and is consistent with a process in which crack initiation is controlled in part by the high tensile stress at dislocation channel-grain boundary intersections. [ABSTRACT FROM AUTHOR]

The 304L SS was irradiated in Barsebäck 1 BWR reactor to 5.9 dpa followed by post-irradiation annealing at 500 °C and 550 °C up to 20 h. Correlations of irradiation hardening with the main hardening features of dislocation loops and Ni/Si-rich clusters were made using the least squares regression. The dispersed-barrier hardening model tended to overestimate the contribution of Ni/Si-rich clusters in the as-irradiated condition. Contribution of Ni/Si-rich clusters to irradiation hardening appeared to be more appropriately accounted for by the short-range order strengthening mechanism and was dependent on the Ni and Si concentrations in the clusters. Irradiation hardening in the as-irradiated condition was predominantly from dislocation loops with minor contribution from the Ni/Si-rich clusters. This is consistent with recent molecular dynamics simulations which showed that Ni/Si-rich clusters were weak obstacles to dislocation glide. As dislocation loops were annihilated at a faster pace than the Ni/Si-rich solute clusters, the Ni/Si-rich clusters played a more important role in the remaining hardening after thermal annealing. Contribution of Ni/Si-rich clusters to remaining hardening outweighed that of dislocation loops after thermal annealing at 550 °C for 5 h and became predominant after annealing for 20 h. [ABSTRACT FROM AUTHOR]