Neutron Irradiation Conditions Research Articles

Effects of neutron irradiation on failure behavior analysis of high-density bumps under thermo-coupling conditions

As the space environment has conditions such as high vacuum, strong irradiation, and large temperature difference, spacecraft will face a series of challenges such as high-energy particles and extreme temperatures during on-orbit operation, especially neutron irradiation, which has a larger mean free range and a deeper depth of penetration, and which will cause serious damage to the electronic devices. Solder joints, the most failure-prone part of electronic devices, have received extensive attention from related researchers. In this study, the evolution of the microstructure of bumps under neutron irradiation and thermal field coupling conditions is analyzed utilizing scanning electron microscopy (SEM)\\electron backscatter diffraction (EBSD), and the changes in the mechanical properties of the bumps are also analyzed by nanoindentation tests and push-ball experiments. The results show that the incidence of high-energy neutrons generates a large number of atomic vacancies, which exacerbates the Kirkendall effect. The point defects caused by neutrons gather together to form defect clusters. These microscopic defects impede the movement of slip dislocations, leading to the hardening of the material and an increase in elastic modulus. In contrast, the accumulation of a large number of defects leads to the failure of the bump.

Comparison of PM-HIP to forged SA508 pressure vessel steel under high-dose neutron irradiation

Powder metallurgy with hot isostatic pressing (PM-HIP) is an advanced manufacturing process that is envisioned to replace forging for heavy nuclear components, including the reactor pressure vessel (RPV). But PM-HIP products must at least demonstrate comparable irradiation tolerance than forgings in order to be qualified for nuclear applications. The objective of this study is to directly compare PM-HIP to forged SA508 Grade 3 Class 1 low-alloy RPV steel at two neutron irradiation conditions: ∼0.5–1.0 displacements per atom (dpa) at ∼270 °C and ∼370 °C. PM-HIP SA508 experiences greater irradiation hardening and embrittlement (total elongation) than forged SA508. However, uniform elongation and approximate toughness are comparable across all irradiated materials, suggesting irradiated PM-HIP SA508 exhibits superior ductility at maximum load-bearing capacity. The irradiation hardening mechanism is linked to composition rather than fabrication method. Since PM-HIP SA508 has higher Mn and Ni concentration, it is more susceptible to irradiation-induced nucleation of Mn-Ni-Si-P (MNSP) nanoprecipitates and dislocation loops, which both contribute to hardening. Conversely, the forged material nucleates fewer MNSPs, causing dislocation loops to control irradiation hardening. These results show promise for the irradiation performance of PM-HIP SA508 and can motivate future nuclear code qualification of PM-HIP fabrication for RPVs.

The critical influencing factors responsible for the particle cracking in UMo/Zr dispersion fuel plates during post-irradiation anneal tests

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.

Comparative analysis of gas release from biphasic lithium ceramics pebble beds of various pebbles sizes and content under neutron irradiation conditions

This paper presents the results of 4 reactor campaigns on the irradiation of biphasic lithium ceramics containing different ratios of lithium orthosilicate (LOS) and lithium metatitanate (LMT) components (25 and 35 mol% LMT in LOS). The size distribution of pebbles in pebble beds was 250–1250 μm and 500–710 μm, respectively. The studies were carried out sequentially with each type of ceramics. In experiments carried out using the vacuum extraction method, the composition of gases released from lithium ceramic samples was registered in in-situ mode. The absence of purge gas during the experiments minimized the possibility of T2O and HTO formation, significantly simplifying processing of the results and providing more opportunities for results analysis. The main goal of the present paper was to identify the general patterns of the release of gases with mass numbers M2 (H2), M4 (He + HT), M6 (T2) and M18 (H2O) from ceramic samples throughout the entire irradiation experiment in 4 campaigns. Release trends of main gases with mass numbers M2, M4, M6 and M18 for all four campaigns are presented and their comparative analysis was performed. The average partial pressure of tritium release in the form of T2 and HT molecules for all campaigns was determined as 5.8 × 10-7 Torr. The dependences of formation rates of helium release peaks on the irradiation time were plotted. The nature of peak emissions does not have a monotonic relationship; upon irradiation, both an increase in the frequency of peaks and a decrease in it are observed. During irradiation, the process of peak helium release does not stop.The simulation was carried out by the finite element method, assuming that tritium release from the sample is determined by diffusion and desorption processes from the sample surface. The experimental curves are satisfactorily described by a number of sets of desorption and diffusion parameters. One way or another they lie in the range of specified values. The Arrhenius dependences of the effective diffusion coefficient and desorption coefficient obtained for lithium ceramics 35 LMT are equal to:D=5,2×10-11m2sexp-21kJmoleRT,K=1.21×10-4m2sexp-64kJmoleRT.The values of the effective diffusion coefficient and tritium desorption coefficient in 25 LMT ceramics were 15 and 20% lower than in 35 LMT ceramics.

The effect of static stresses on the acceleration of the decomposition processes in the ferritic-martensitic steels in conditions of neutron irradiation to a dose of 85 dpa

Behavior of the corrosion-resistant EP823 steel (Fe-11Cr-Mn-Mo-Nb-V-W-Ni-Si-0.17C) and its ODS modification, hardened by yttrium, yttrium-titanium oxides (EP823-ODS) in conditions of static mechanical stress under high dose exposure in a fast-neutron reactor was studied using Mössbauer spectroscopy and electron microscopy. Mechanical stresses were revealed to result in a deformation of the fuel elements walls and the acceleration of the decomposition processes in steel structure with a release of chromium into the intermetallic χ phase and to the Me23C6 carbide formation. It was also shown that the key factor accelerating the structure decomposition upon neutron irradiation to a dose of 85 dpa at 505 °C is anisotropy of stresses between longitudinal and transverse directions of fuel cladding tube surface.

Irradiation experiments of titanium beryllide samples in the WWR-K reactor

This paper presents the results primary characterization of unirradiated titanium beryllide samples and neutron analysis of irradiation experiments of titanium beryllide Be12Ti samples in the WWR-K reactor. The XRD analysis of the samples was carried out. The surface morphology of the samples was studied using optical and electron microscopy. Two options of the irradiation capsule design are considered. The first design is designed for low-temperature irradiation, and the second design is for high-temperature irradiation. The effect of the capsule design on the neutron irradiation conditions of titanium beryllide samples is shown. The dynamics of tritium, helium-4 and helium-3 generation in samples is given. The values of the specific energy release of samples and structural materials of the capsule are calculated. The rate of displacements per atom in the irradiated material is estimated.

Modeling and characterization of beryllium reflector elements under irradiation conditions

This study aims at modeling the beryllium reflector poisoning under neutron irradiation conditions and calculating the impact of beryllium poisoning on the core parameters of ETRR-2 research reactor. The CITVAP code was used to calculate the neutron flux and parameters of ETRR-2 core with beryllium reflector elements. The neutron flux in each reflector element was calculated to solve the modeling equations for the atomic densities of lithium-6 (6Li), tritium-3 (3H), and helium-3 (3He) using the BERYL program. The results are discussed based on CITVAP calculations of the core excess reactivity and cycle length Full Power Days (FPD). Possible solutions to minimize the degradation due to beryllium poisoning are also discussed and compared based on calculations.

The Effect of Cooling Rate on Crystallographic Features of Phase Transformations in Zr-2.5Nb.

Zirconium (Zr) alloys are utilized as structural components for the cores of nuclear reactors due to the excellent combination of their mechanical properties and corrosion resistance under intense neutron irradiation conditions in water. The characteristics of microstructures formed during heat treatments play a crucial role in obtaining the operational performance of parts made from Zr alloys. This study investigates the morphological features of (α + β)-microstructures in the Zr-2.5Nb alloy, as well as the crystallographic relationships between α- and β-phases. These relationships are induced by the β→α(α″) displacive transformation that occurs during water quenching (WQ) and the diffusion-eutectoid transformation that takes place during furnace cooling (FC). To conduct this analysis, samples solution treated at 920 °C were examined using EBSD and TEM. The experimental distribution of α/β-misorientations for both cooling regimes deviates from the Burgers orientation relationship (BOR) at a discrete set of angles close to 0, 29, 35, and 43°. The experimental α/β-misorientation spectra are confirmed by crystallographic calculations for the β→α→β-transformation path based on the BOR. Similar spectra of misorientation angle distribution in α-phase and between α and β phases in Zr-2.5Nb after WQ and FC point to similar transformation mechanisms and the significant role of shear and shuffle in β→α-transformation.

Recent progress of plasma exhaust concepts and divertor designs for tokamak DEMO reactors

The power exhaust concept and an appropriate divertor design are common critical issues for tokamak DEMO design activities which have been carried out in Europe, Japan, China, Korea and the USA. Conventional divertor concepts and power exhaust studies for recent DEMO designs (Pfusion = 1 – 2 GW, Rp = 7 – 9 m) are reviewed from the viewpoints of the plasma physics issues and the divertor engineering design. Radiative cooling is a common approach for the power fusion scenario. Requirements on the main plasma radiation fraction (fradmain = Pradmain/Pheat) and the plasma performance constrain the divertor design concept. Different challenges contribute to optimizing the future DEMO designs: for example, (i) increasing the main plasma radiation fraction for ITER-level Psep/Rp designs and simplifying the divertor geometry, and (ii) extending ITER divertor geometry with increasing divertor radiation (Praddiv) for larger Psep/Rp ≥ 25MWm−1 designs. Power exhaust simulations with large Psep = 150 – 300 MW have been performed using integrated divertor codes considering an ITER-based divertor geometry with longer leg length (1.6 – 1.7 m), as in a common baseline design. Geometry effects (ITER like geometry or more open one without baffle) on the plasma detachment profile and the required divertor radiation fraction (fraddiv = Praddiv/Psep) were key aspects of these studies. All simulations showed that the divertor plasma detachment were extended widely across the target plate with a reduction in the peak heat load of qtarget ≤ 10 MWm−2 for the large fraddiv = 0.7 – 0.8, while the peak qtarget location and value were noticeably different in the partially detached divertor. Simulation results also demonstrated that radial diffusion coefficients of the heat and particle fluxes were critical parameters for DEMO divertor design, and that effects of plasma drifts on outboard-enhanced asymmetry of the heat flux, suggested the need for longer divertor leg to ensure the existence of a detached divertor operation with qtarget ≤ 10 MWm−2.Integrated design of the water cooled divertor target, cassette body (CB) and cooling pipe routing has been developed for each DEMO concept, based on the ITER-like tungsten monoblock (W-MB) with Cu-alloy cooling pipes. Engineering design adequate under higher neutron irradiation condition was required. Therefore, inlet coolant temperature (Tcool) was increased. In current designs, it still shows a large potential variation between 70 °C and 200 °C. The influence of thermal softening on the Cu-alloy (CuCrZr) pipe was fostered near the strike-point when the high qtarget of ∼10 MWm−2 was studied. Improved technologies for high heat flux components based on the ITER W-MB unit have been developed for EU-DEMO. Different coolant conditions (low- and high-Tcool) were provided for Cu-alloy and reduced activation ferritic martensitic (RAFM) steel heat sink units, respectively. The high-Tcool coolant was also considered for the CB and supporting structures. Appropriate conditions for the high-Tcool coolant, i.e. 180 °C/ 5 MPa (EU-DEMO) and 290 °C/ 15 MPa (JA-DEMO, CFETR and K-DEMO), will be determined in the future optimizations of the divertor and DEMO design.

Phase Formation Features of Reactor Pressure Vessel Steels with Various Ni and Mn Content under Conditions of Neutron Irradiation at Increased Temperature

In this paper the phase formation and mechanical properties of VVER-type reactor pressure vessel (RPV) steels with various Ni (1.57–5.95 wt.%) and Mn (0.03–0.76 wt.%) content after neutron irradiation up to fluences in the range of (53–120) × 1022 n/m2 at 400 °C were studied. The possibility of carbonitride formation under these irradiation conditions is shown. In case of sufficient Ni (>1.5 wt.%) and Mn (>0.3 wt.%) content formation of Ni-Si-Mn precipitates is observed. Their chemical composition is close to G-phase and Γ2-phase and differs from that of radiation-induced precipitates in VVER-1000 RPV steels. This indicates the prerequisites for thermally conditioned mechanism of Ni-Si-Mn precipitates formation and growth at 400 °C enhanced by irradiation. It is also shown that the optimized steel manufacturing technology coupled with an ultralow Mn content (≤0.03 wt.%) in steel with increased up to 5.26 wt.% Ni content facilitates suppressing the Ni-Si-Mn precipitates and carbonitrides formation. This, in turn, reduces the contribution of the hardening embrittlement mechanism and, correspondingly, facilitates high radiation resistance of the steels with ultralow Mn content at the increased irradiation temperature (400 °C).

Influence of Neutron Irradiation Conditions on Maximal Frequency of IR Absorption Spectra of LiF:OH Single Crystal

For many solids operating under irradiation conditions, the study of their behavior at the neutron irradiation is of importance. Herein, the influence of neutron irradiation conditions on the infrared (IR) absorption spectra of LiF crystals containing different amounts of hydroxyl and metallic impurities is considered. The investigations are carried out for discovering new peculiarities under the extreme irradiation conditions of LiF—being a very important material for practical applications. In the crystals with OH− impurity after the action of neutron irradiation, quite clear bands are appeared in the region of 1900–2200 cm−1. The delivered doses vary from 1014 to 3 × 1017nnvt. LiF crystals are investigated by differing both in qualitative and quantitative contents of OH− ions. With the increase in irradiation dose, the maximum IR absorption of all investigated groups of crystals shifts to the region of low energies. The appearance of a new band in the IR spectrum and its shift as a function of dose is associated with the radiolysis of OH− molecule.

Quantifying defect production in solids at finite temperatures: Thermally-activated correlated defect recombination corrections to DPA (CRC-DPA)

It has long been recognized from low-temperature defect production experiments and molecular dynamics (MD) simulations of displacement events in metals that the surviving defect fraction relative to the calculated Norgett-Robinson Torrens (NRT) displacements per atom (dpa) ranges from ∼1 for near-threshold displacements (representative of electron irradiation) to ∼0.3 for energetic displacement cascades (relevant for fast neutrons or energetic heavy ion irradiations). This correlated point defect recombination that occurs within a few picoseconds following the creation of a primary knock-on atom (PKA) has recently been formalized in the athermal recombination corrected dpa (ARC-DPA) parameter. However, at elevated temperatures additional thermally-stimulated correlated defect recombination processes need to be considered. For example, at irradiation temperatures above defect recovery Stage I (typically ∼40–100 K for most metals), mobile self-interstitial atoms induce additional thermally-activated correlated defect recombination that can amount to 70–80% reduction in the surviving defect fraction for low PKA energies and ∼20 to 30% reduction for high PKA energies. Since most nuclear material applications involve operation above room temperature, it is generally important to include this additional correlated defect recombination in radiation effects models. Published experimental defect recovery and defect production studies on 26 metals are analyzed to extract the surviving defect fraction following thermally-activated correlated recombination as a function of PKA energy. These experimental estimates of the athermal- plus thermal-activated correlated recombination-corrected DPA (CRC-DPA) are compared with new long-term MD simulation results on Cu and Fe. The analysis indicates the CRC-DPA fraction of surviving defects in most elements for irradiation at temperatures where interstitials are mobile is ∼25% of the NRT DPA over a wide range of PKA energies relevant for electron, ion and fast neutron irradiation conditions. Limited experimental and kinetic Monte Carlo studies suggest a further reduction to ∼10% of the NRT DPA over a wide range of PKA energies when correlated recombination due to thermally activated vacancy migration is considered. This temperature-dependent CRC-DPA should be used as the defect production parameter (rather than NRT DPA or ARC-DPA) for materials irradiated at all temperatures where defects are mobile.

Cavity evolution and void swelling in dual ion irradiated tempered martensitic steels

Microstructural evolutions in two variants of the F82H 8 wt%Cr tempered martensitic steel under dual (He+ plus Fe3+) ion irradiations at 500 °C have been characterized by transmission electron microscopy. Emphasis is on understanding the combined effects of displacements per atom (dpa) and the helium/dpa ratio (He/dpa) on cavity evolution and void swelling (fv). The dual ion database includes 231 alloy-dpa-He/dpa conditions for irradiations up to ∼ 82 dpa and 3700 appm helium. We hypothesize that the large “scatter” observed in the fv data is primarily due to local microstructural variations, especially those that shield regions at shallower depths from the effects of injected interstitial atoms, which would otherwise reduce fv. More generally, our hypothesis is that the highest fv data, at a specified dpa and He/dpa, provides the most appropriate estimate of swelling for fusion neutron irradiation conditions. The dose dependence of fv can be described by an incubation dpai, marking the onset of significant void growth, and a post-incubation swelling rate, fv’ = Δfv/Δdpa. Both dpai and fv’ decrease with increasing He/dpa. The fv’ data are quantitatively consistent with a simple model for defect partitioning between evolving bubble, void, and dislocation sinks strengths and biases. Most notably, the model correctly predicts both the incubation dpai, and the subsequent fv’, in spite of the fact that they reflect somewhat different physics. Our analysis also suggests that defect recombination is modest at 500 °C. Along with other considerations, this supports the hypothesis that, even at high dpa rates, dual ion irradiations can be used to reasonably emulate void nucleation and growth for a range of fusion-relevant neutron irradiation service conditions. A companion paper incorporates the results of this study in a reduced-order, data-driven void swelling model.

Theoretical study on the correlation of swelling peaks between neutron and heavy ion irradiated 15-15Ti stainless steel

Titanium-modified austenitic stainless steels (15-15TiSS) are currently used as the fuel cladding material of fast reactors, which are subject to higher radiation damage during their application. In this work, the radiation-induced swelling of 15-15TiSS under a fast reactor neutron and heavy ion irradiation conditions was studied by the rate theory (RT). The simulated swelling properties of 15-15TiSS under fast neutron conditions were calculated initially. The swelling peak, swelling rates, and swelling-related microstructural properties are consistent with the neutron irradiation results, indicating that the selected RT model and material parameters are reasonable. Then, the swelling properties of 15-15TiSS under various damage rates were predicted by changing the radiation damage rates from 1 × 10−6 to 1 × 10−3 dpa/s. It shows that swelling peaks are strongly dependent on temperature and the swelling peaks shift ∼50 °C toward the higher temperature with each order of magnitude increase of defects generation rate. The swelling rates and swelling-related defect evolution at 1 × 10−3 dpa/s (with a swelling peak temperature of 590 °C) are consistent with that under neutron irradiation with 1 × 10−6 dpa/s (with a swelling peak temperature of 460 °C). At length, the RT-predicted heavy ions irradiation results were verified by the previous positron annihilation lifetime spectroscopy results of ions irradiated 15-15TiSS. It indicates that heavy ion irradiation can be used to study the radiation effect of materials under neutron irradiation and should be a feasible technique used in the further screening of radiation-resistant materials.

Effects of self-irradiation on deuterium retention and reflectivity of molybdenum, fusion plasma-facing material: Combined experimental and modeling study

Molybdenum is used as plasma-facing material in tokamaks and as material for plasma optical diagnostics mirrors. Harsh conditions of neutron irradiation, exposure to hydrogen isotopes and helium ions, and high operating temperatures result in degradation of the molybdenum surface and ultimately limit their lifetime in a fusion power plant. In the current paper, intake and subsequent thermal release of deuterium from self-irradiated by high energy (1 MeV) ions molybdenum as a function of irradiation dose are investigated. Several characteristic temperature regions where deuterium release takes place are identified and attributed to trapping of deuterium in intrinsic and radiation-induced microstructure defects. This attribution is further validated by molecular dynamics modeling, which confirms that increase and saturation of vacancy concentration found in simulations follows increase and saturation of experimentally determined deuterium content. Deuterium inventory and vacancy content saturate at a damage level of around 0.2 dpa (displacement per atom), similar to recent modeling and experimental studies of iron and tungsten. Reflectivity measurements of irradiated molybdenum show that it is only slightly affected by damage up to 1 dpa.

On the theory of void nucleation in irradiated crystals

The theory of void nucleation in irradiated materials is critically analysed and further developed. It is shown that the existing theory overestimates the rate of empty void nucleation in ion-irradiated steels by several orders of magnitude and therefore requires revision for an adequate interpretation of ion bombardment tests. To extend the nucleation model to the conditions of reactor neutron irradiation, it is modified to take into account the effect of void stabilisation by transmutation gas (He) in steels or by fission gases (Xe, Kr) in fuels. It is shown that the presence of gas atoms in a solid solution strongly reduces the free energy of void formation, but does not affect the number of void nucleation sites, which is determined by the concentration of vacancies in the irradiated material. The model developed for nucleation of small gas filled voids (bubbles) is further modified to adequately consider their further growth, stabilisation (due to athermal resolution of gas atoms) and generation of a new population of larger bubbles. The obtained analytical expression for the nucleation rate is applied to a simplified analysis of experimental data on neutron irradiation of stainless steel in PWRs and FRs. For a more accurate analysis of these tests, it is recommended to implement the new model in a fuel performance code, additionally taking into consideration the migration and coalescence of small bubbles.

Studies of two-phase lithium ceramics Li4SiO4-Li2TiO3 under conditions of neutron irradiation

This work presents the preliminary experimental data on the study of gas release from two-phase lithium ceramics Li4SiO4-Li2TiO3 under neutron irradiation conditions. Experiments were carried out at the WWR-K research reactor (Almaty, Kazakhstan) for ∼4.3 days. The total neutron fluence during the irradiation was ∼1.8·1019n/cm2. In the course of irradiation, two experiments on ceramics heating during irradiation and two experiments with hydrogen isotopes (H2 and D2) supply into the experimental chamber with the sample were performed at a temperature of 680 °C and reactor power of 6 MW. During the entire irradiation, the gas composition in the continuously evacuated ampoule device with samples was recorded. The obtained dependences of the release of tritium-containing molecules and helium during the experiment were qualitatively analyzed.

Editorial: Special Issue on Materials for Extreme Environments, Part 2

Abstract The response of materials to extreme conditions of heat flux, stresses, strain rates, and corrosive environments provides an insight into new phenomena, reveals failure modes that limit technological possibilities, and presents novel routes for making new and improved materials. Exposing materials to extreme environments induces new physical phenomena, which are central to the challenges of science. Understanding and exploiting these extreme environments is critical for facing many of the future energy challenges, as materials for these usually operate at extremes of pressure, temperature, and chemical reactivity. Identifying materials that not only survive but also perform under extreme conditions to the design requirements is a major concern in energy research. It is in this context that this special issue on materials for extreme environment becomes relevant and useful for readers in their research endeavors. It has been a pleasure to be guest editors for the two-part special issue on Materials for Extreme Environments for Materials Performance and Characterization. The first part of special issue covered 16 papers, essentially covering the areas of materials for cryogenic applications and ultrahigh temperature, advanced materials, and manufacturing technologies. The second part of the special issue contains the remaining 18 papers, essentially showcasing recent advances in materials for energy applications. There are two review articles in this volume. The first is by Prasad Reddy et al. from the Indira Gandhi Centre for Atomic Research, India, on “Core Materials for Sodium Cooled Fast Reactors: Past to Present and Future Prospects,” which reviews materials that are subjected to extreme conditions of intense fast-spectrum neutron irradiation, high temperatures, and mechanical/chemical fuel-cladding interactions. The second article, by Patil et al., “A Review on the Material Development and Corresponding Properties for Power Plant Applications,” brings out the recent advances in materials for power plants. These are followed by six papers each on corrosion/oxidation and welding of advanced materials for use in extreme environments. We sincerely thank both the authors and reviewers for their hard work and dedication. We wholeheartedly thank the ASTM International staff dealing with the special issues for all their efforts. We sincerely hope that you enjoy the papers in this special issue.

Core Materials for Sodium-Cooled Fast Reactors: Past to Present and Future Prospects

Abstract The Sodium-Cooled Fast Reactor (SFR) is the most promising and technologically evolved among the six nuclear reactor concepts selected by the Generation IV International Forum. Although it is compatible with a closed fuel cycle for sustainable energy production, its competitiveness depends on being able to achieve high fuel burn-up up to 200 GWd/t. This would enable efficient fuel utilization to minimize fuel cycle costs. Hence, development of optimized core materials, especially for fuel cladding tubes that are subjected to extreme conditions of intense fast-spectrum neutron irradiation, high temperatures, and mechanical and chemical fuel–cladding interactions, continues to be a priority worldwide. Resistance to void swelling, irradiation creep, and embrittlement are required to be enhanced to minimize dimensional changes and loss of ductility in core components, which limit achievable burn-up. Current SFR core materials comprise austenitic stainless steels (SS) and ferritic-martensitic (FM) steels. Cold-worked austenitic SS grades 304, 316, 316Ti, niobium-stabilized grades FV548, EI-847, EP-172, and Ti-modified 15Cr-15Ni SS, such as D9, 1.4970, 15/15Ti SS, and ChS-68, have been used for fuel cladding and ducts in SFRs built to date, and they withstand neutron irradiation damage up to 80–100 dpa. Subsequent improvements have been made by optimizing minor elements (titanium, silicon, and phosphorous) and multi-stabilization to develop advanced grades such as Si-modified 15-15Ti, Indian Fast Reactor Advanced Clad (IFAC-1), PNC-316, EK-164, and HT-UPS, which could support burn-up of 150 GWd/t. In the case of FM steels, several commercial grades are found suitable in view of their inherent void swelling resistance to 200 dpa, and are being developed for improved creep strength by oxide dispersion strengthening to realize burn-up of 200–250 GWd/t. This article presents the development of structural materials for SFR fuel pin cladding and ducts and associated enhancement of permissible fuel burn-up. Indian and international experience resulting from extensive in-pile and out-of-pile mechanical testing and fuel–cladding interactions has been covered.

Large helium-vacancy clusters coalescence during helium bubble evolution under cascade in iron with edge dislocation: A MD simulation

The nucleation and growth of helium bubbles under the mutual effect of cascade and dislocation in bcc Fe was investigated by molecular dynamic simulations. The helium atoms along with primary knock-on atoms (PKA) were gradually added in the simulation system to imitate the real neutron irradiation condition. Several coalescence events of large helium-vacancy (He-V) clusters during the molecular dynamic simulation in bcc Fe were discovered. The detailed analysis of these coalescence events revealed that in the bulk region, if two helium-vacancy clusters are close enough, they are inclined to combine into one cluster by kicking out host atoms between the two clusters. However, two close helium-vacancy clusters near the dislocation line will not merge by this kick-out mechanism. Instead, it’s found that they could be promoted to combine together by cascade, even if the two clusters locate on the edge of the cascade thermal spike. The mutual influence of dislocation and cascade will lead to the formation of low content and large size helium-vacancy clusters around the dislocation line.