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Abstract: We thank R. Konings et al. for their interest and their valuable critical discussion of our article regarding the fission product release from irradiated oxide fuel during thermal treatment and reply to their comments appearing in this issue. Their feedback stimulated us to give more details on the sampling procedure of investigated materials as well as the measurement procedure in order to exclude misunderstandings. The release curves for iodine and cesium are compared to blank profiles and reanalyzed to demonstrate the features of inductive heating approach applied in authors’ recent study on FP release under inert and oxidizing conditions. [Copyright &y& Elsevier]

In nuclear power plants, CuCrZr alloy and pure copper (T2) are widely used as the structural materials and service at high temperatures, and the comprehensive understanding of the mechanical properties is essential. In this paper, the small punch test (SPT) is used to understand the variations of mechanical properties and fracture mechanism with temperature for them. The soft temperature of CuCrZr alloy is identified as 400 °C, with the sharp decrease in SPT strength parameters and fracture energy, which is caused by the oxidation. The SPT fracture mode changes from the central line crack to "O" shape circumferential crack, and the mixed oxidation and plastic deformation fracture mechanism is observed at 500 °C. The soft temperature of T2 is 300 °C, with the sharp decrease in the SPT fracture displacement. The "O" shape circumferential crack and the necking phenomenon are unchanged for the fracture mode of T2, but the fracture mechanism changes from transgranular fracture to intergranular fracture, with increase in temperature. This paper describes the deformation behaviour of CuCrZr alloy and T2 using SPT as a function of temperature, investigates the fracture behaviour and attempts to bring out the effect oxidation during high temperature tests. [ABSTRACT FROM AUTHOR]

This paper is devoted to experiments on testing of the Sn-Li capillary porous system (CPS) under conditions of deuterium plasma irradiation, carried out at a plasma-beam installation (PBI). As a metal CPS matrix, molybdenum mesh was used. The paper presents a detailed description of the development of technology and procedure for the fabrication of the investigated sample, intended for plasma testing, as well as experiments on irradiation of Sn-Li CPS with deuterium plasma. As a result of the experiments performed, the dependences of the temperature of the investigated sample on the energy of the plasma contacting with the CPS, time dependence of the gas phase composition in the PBI chamber during irradiation of Sn-Li CPS with deuterium plasma, optical spectra depending on the temperature were obtained. Post-experimental material science studies with a sample of Sn-Li CPS were also performed and the results of microstructure, thermal and X-ray phase analysis were obtained. Analysis of the experimental data obtained showed that the energy of deuterium plasma precipitated on the Sn-Li alloy is re-emitted into the energy of optical radiation on the evaporating neutral atoms of lithium and tin. A complete evaporation of the alloy from CPS, during interacting with deuterium plasma, leads to partial destruction of the metal matrix of molybdenum. [ABSTRACT FROM AUTHOR]

This paper presents a comprehensive nuclear forensics signatures analysis of UO 2 fuel pellets, allowing their provenance to specific manufacturers. The study is based on a synthetic database designed for the 4th international Galaxy Serpent exercise circulation, which mimics real fuel pellet characteristics. The scenario of the exercise is completely fictional, including the data used and manufacturers' names, locations, etc. The signatures investigated include chemical form, pellet dimensions, equivalent boron content, Gd additive, impurities distribution, and isotopic composition. The consistency levels of the results, which determine the confidence in the provenance process, were divided into four levels due to the varying resolving power of the signatures. The paper emphasizes the importance of a graphical representation of the results to simplify interpreting multi-degree-of-freedom data, typical to nuclear forensics investigations, including the amalgamation of quantitative and qualitative information. The analysis connects 11 orphan pellets and a single UO 2 powder to five manufacturers. The presented methodology offers a straightforward approach to data interpretation in nuclear forensic investigations and highlights the need for continued research and development in this field. [ABSTRACT FROM AUTHOR]

Lithium-based ceramics exhibit good thermophysical properties, have low activation and chemical activity, and also release tritium well. These characteristics make lithium ceramics the best candidates for use as a functional material for a solid breeder blanket. To date, the most accurate understanding of the processes of tritium and helium production and release occurring in the breeder blanket materials can only be obtained from the results of reactor experiments. On the other hand, many important parameters can only be estimated in post-irradiation experiments (PIE). This paper describes studies of tritium and helium release in post-irradiation experiments (PIE) on the thermal desorption from ceramic pebbles Li 2 TiO 3 + 5mol% TiO 2 (with 96% enrichment on lithium-6 isotope), irradiated in the WWR-K reactor for 223 days. In PIE a peak of helium release was recorded for each sample in the temperature range of 1300–1500 K. An assumption was made in the paper to explain the nature of this helium release peaks from lithium ceramic pebbles. [ABSTRACT FROM AUTHOR]

The IFMIF-DONES Facility is built with the purpose of irradiating materials under DEMO Tokamak-like conditions and is a first-of-a-kind project that is foreseen to be built in Granada, Spain. A systematic top-down approach is used to design its Systems and to aid the harmonization and interaction between them. One of these main systems is the Test Systems which serves as the meeting point for other major systems such as the Lithium System (LS) and Accelerator Systems (AS), while also providing a connection to the Facilities for Complementary Experiments (FCE) for additional irradiation campaigns to take place. The Test Cell is a confined space for the experiments to take place, with critical functions for operation and safety. The TC is a subsystem serving as a convergent space for other systems, therefore the design of the TC has to fulfil the needs and requirements of the connecting systems also. In this paper, the design evolution of the Test Cell is discussed in detail from the early concept to the last more mature design, while also describing the connecting systems and the challenges these provide. [ABSTRACT FROM AUTHOR]

MOX fuel is one of the most widely used nuclear fuels in nuclear power plants. Due to the inhomogeneity of Pu, MOX fuel is prone to the formation of Pu-rich agglomerates and the occurrence of pore aggregations during irradiation. This paper aims to obtain microstructural information of the fuel using spatial statistics methods and investigate the impact of pore aggregation on MOX fuel damage based on the configurational forces of the M -integral. Firstly, the pore aggregation model is established by random sequence adsorption method twice, and the relationship between fuel model size and pore diameter is determined by point pattern criteria based on spatial statistics methods. Secondly, through spatial statistics methods, the reliability of the pore aggregation model to characterize different aggregation levels is verified. Simultaneously, it is confirmed that the M -integral delineating the global state is a better indicator of fuel damage compared to physical quantities that delineate the local state. Finally, a damage assessment model based on the M -integral is established by the basic characteristics of MOX fuel. The effects of the volume fraction of Pu-rich agglomerates, the Pu weight fraction within Pu-rich agglomerates and the radial position of the fuel pellet on the fuel damage are analyzed. It is concluded that the concentrated distribution of Pu promotes the damage of MOX fuel. The MOX fuel pellet center suffers the most damage. This study has proposed a damage evaluation method for MOX fuel microstructure with pore aggregation and established a relationship between the spatial distribution of fuel pores and fuel damage. [ABSTRACT FROM AUTHOR]

• The pre-precipitation TMT greatly increases MX phase. • The pre-precipitation TMT greatly decreases M 23 C 6 phase. • The pre-precipitated MX phase delays the coarsening of the lath. China Low Activation Martensitic steel (CLAM) is a cladding material used in thermonuclear fusion reactors. Its high temperature performance is closely related to precipitation strengthening. The MX phase with strong thermal stability is the key to improving the high temperature performance of CLAM steel, and the M 23 C 6 phase which is easy to coarsen under long-term high temperature service conditions is one of the important factors causing creep failure. In order to increase the content of MX phase and reduce the content of M 23 C 6 phase, this paper first pre-precipitates near the precipitation temperature of M 23 C 6 phase (which is also close to the maximum precipitation temperature of MX phase), which limits the precipitation of M 23 C 6 phase and promotes the maximum precipitation of MX phase. Then the subsequent thermomechanical treatment (TMT) is carried out. The results show that the content of M 23 C 6 phase in the samples after pre-precipitation TMT is greatly reduced, and the size is slightly reduced compared with the samples with only thermal deformation. The size of the MX phase is reduced by 10.6%, the volume density is increased by 88%, the volume fraction is increased by 34.1%, and the dislocation density is slightly increased. At the same time, it was found that a large number of MX phases precipitated at the lath boundary and pinned the boundary, which delayed the coarsening of the lath during tempering. The changes of these microstructures improve the precipitation strengthening, dislocation strengthening and boundary strengthening, so the room temperature tensile properties and high temperature tensile properties are improved, but the work hardening makes the total elongation and room temperature impact properties decrease. [ABSTRACT FROM AUTHOR]

Reflood plays an important role in preventing the serious accident process. In this paper, the reflood oxidation performances of Cr-coated Zr-Sn-Nb alloy cladding tubes at 1000 °C, 1100 °C and 1200 °C were studied by high-temperature steam oxidation followed by in-situ water quenching. The oxidation kinetics, macroscopic morphology, phase transformation and microstructural evolution were investigated. Cr coating can provide excellent anti-reflood oxidation protection for Zr-Sn-Nb alloy, resulting in an oxidation rate constant of only 1/4–1/7 of that of uncoated Zr-Sn-Nb alloys. However, the cladding tube degraded into fragments due to excessive thermal stress during the in-situ water quenching after reflood oxidation at 1200 °C for 4000 s. After reflood oxidation at 1000 °C and 1100 °C, the most prominent feature of the surface is a porous flocculation structure main due to the formation of volatile products, while after 1200 °C, it is a fine mackerel scale-like structure accompanied by protruding whiskers main owing to Cr3+ rapidly migrates outward to the outer surface via short circuit paths. After oxidation, the cross-section of Cr-coated Zr-Sn-Nb alloys exhibits a multi-layer structure. The thickness of each layer changes approximately linearly with oxidation time. However, due to the redox reaction, compared to that after oxidation for 2000s, the Cr 2 O 3 layer becomes thinner and the residual Cr coating becomes thicker after oxidation at 1200 °C for 4000 s. [ABSTRACT FROM AUTHOR]

Weibull stress-based methods for failure probability assessment have been developed and analyzed to assess the integrity of tristructural isotropic (TRISO) fuel particles during fuel life cycles and accident operating conditions. While simple, these methods entail a number of drawbacks when stress concentrates near crack tips, including finite element mesh size dependency when the Weibull stress is averaged over the finite element domain. Fracture mechanics approaches involving the use of interaction integrals eliminate this lack of mesh convergence and produce consistent fracture predictions. In this work, we use an interaction integral approach to computing stress intensity factors for a crack in the inner pyrolytic carbon layer perpendicular to the silicon carbide layer, which is simplified representation of a failure mode in TRISO particles. The interface between these two TRISO layers has been shown to become porous, which we simulate by considering a transition of mechanical properties over such porous length, i.e. the layers are modeled as a functionally graded material. Aspects such as porosity and thermal and irradiation eigenstrains are considered in computing the stress intensity factor from a fracture mechanics approach and compared with the known Weibull stress failure approach. The methodology introduced in this paper enables a more general fracture probability assessment in TRISO particles and eliminates the need to identify best suited parameters when using local or averaged stress-based failure criterion. Finally, the numerical sensitivity studies show how parameters such as the porous transition zone length, the material stiffness, and creep affect the probability of TRISO fuel particle failure. [ABSTRACT FROM AUTHOR]

Several in-core components in nuclear power systems are exposed to high-temperature water in the presence of radiation fields. The dynamic effect of radiation and water chemistry on material performance in these environments is not well understood partly due to significant experimental challenges. A facility consisting of a high temperature/pressure corrosion loop coupled with in-situ Raman spectroscopy has been commissioned to examine material behaviour in more realistic reactor conditions. The in-situ Raman component of the facility has been validated by conducting experiments with both pre-oxidized, and freshly abraded SS304L in water at 80 °C and 300 °C. Testing in water also revealed the detection limitations of the system. The assessment reported in this paper highlights the capabilities to perform degradation studies of key nuclear components by conducting in-situ characterization of materials exposed to high temperature water via Raman spectroscopy, thereby providing chemical, structural and semi-quantitative kinetic information. [ABSTRACT FROM AUTHOR]

Compared to austenitic stainless steels, largely employed in the early fission reactors, high chromium Ferritic/Martenstic (FM) steels, developed since the first half of the 20th century for fossil-fuel power-plants, have a number of advantageous properties among which lower thermal expansion, higher thermal conductivity and better void swelling resistance. At the beginning of the 1970s, FM steels found their first nuclear application as wrapper and fuel cladding materials in sodium-cooled fast reactors. They are now the reference materials for in-vessel components of future fusion reactors, and are considered for in-pile and out-of-pile applications in Generation IV reactors as well as for various other nuclear systems. In this paper, after an introductory historical overview, the challenges associated with the use of FM steels in advanced reactors are addressed, including fabrication, joining and codification issues. The long term evolution of mechanical properties such as the creep and creep-fatigue behaviors is discussed and the degradation phenomena occurring in aggressive environments (lead alloys, high temperature gases, super-critical water and CO 2 , molten salts) are detailed. The paper also provides a brief overview of the radiation effects in FM steels. The influence of the key radiation parameters e.g. temperature, dose and dose rate on the microstructure and mechanical properties are discussed. The need to better understand the synergistic effects of displacement damage and helium produced by transmutation in fusion conditions is highlighted. [ABSTRACT FROM AUTHOR]

The Master Curve approach allows the full characterisation of the ductile to brittle transition region (DBTR) of ferritic steels to be performed with a reduced number of tests. In this paper, the approach has been combined with the application of the small punch (SP) test. Modified SP specimens have been successfully employed to estimate the fracture toughness values of a pressure vessel steel and three structural steels. In addition, a methodology has been proposed, including a validity criterion for the performed tests. The estimated reference temperatures have been compared to the values obtained by means of full-scale conventional techniques. A unique simple relationship between both methodologies has been established for all the analysed materials. Therefore, this paper confirms the suitability of the small punch testing technique for the characterisation of the DBTR of several ferritic steels. It is a promising, simple and cost-effective test, which can be performed with simple equipment. • Small punch notched specimens can be applied to estimate the reference temperature. • A methodology to estimate the reference temperature has been proposed. • A single correlation has been obtained for the ferritic steels analysed. [ABSTRACT FROM AUTHOR]

Abstract When boron carbide is irradiated in nuclear reactors, large helium quantities are produced due to the (n,α) neutron absorption reactions. In a recent paper, we have identified the trapping sites for helium, grain boundaries and damaged zones. In this paper, we propose the determination of an apparent diffusion coefficient for helium. 3He implantations then annealing were performed in B 4 C samples of different grain sizes. The 3He(2H,α)1H nuclear reaction was used to profile helium before and after annealing. Helium profiles with two superimposed components were observed. The narrow component is attributed to helium trapped in the implanted, damaged zone, either in clusters too small to be seen by TEM or as helium-defect complexes. The large component evolution is bounded by the distribution of the grain boundaries surrounding the implanted zone, confirming the high trapping efficiency of the grain boundaries. From an Arrhenius plot, an activation energy for helium diffusion in grains of 2.0 ± 0.2 eV was deduced in a large grain size material. [ABSTRACT FROM AUTHOR]

Abstract The influence of tensile stress on hysteresis loop of Reduced Activation Ferrite & Martensitic steel (RAFM steel) in elastic region was quantitatively evaluated through experiments in this paper. In view of in vessel structures of the RAFM steel in the Tokamak devices, the variation of magnetic material properties caused by forces or mechanical deformation may lead to unexpected influences on their dynamical behaviors, even safety problem. In this paper, tensile experiments of a RAFM steel were conducted under magnetic field up to 60 kA/m to measure its hysteresis loops under different tensile stress. Tensile specimens of a RAFM steel developed in China (CLAM steel) were fabricated and measured with a multi-field coupling experimental system. The maximum relative permeability under different experimental conditions was extracted from the measured hysteresis loops. Experimental results show that a clear variation of the maximum magnetic permeability was observed when the external tensile stress was enlarged. In addition, based on a linear theory of magneto-solid coupling mechanics, a model of constitutive relation between tensile stress and maximum magnetic permeability was put forward and validated based on the experimental results. [ABSTRACT FROM AUTHOR]

Abstract Silicon carbide (SiC) fiber reinforced SiC matrix composites continue to undergo development for fusion applications worldwide because of inherent advantages of the material including low activation, high temperature capability, relatively low neutron absorption, and radiation resistance. This paper presents an international overview of recent achievements in SiC-based composites for fusion applications. Key subjects include applications in fusion reactors, high-dose radiation effects, transmutation effects, material lifetime assessment, and development of joining technology (processing, test method development, irradiation resistance, and modeling capability). This paper also discusses synergy among research for fusion materials and non-fusion materials (for fission and aerospace applications). Finally, future research directions and opportunities are proposed. [ABSTRACT FROM AUTHOR]

The assessment of the post-cracking behaviour of the reactor core components requires knowledge of the fracture properties, such as the fracture toughness, K IC , and work of fracture, γ f , of irradiated graphite. The measurement of these properties is a proven technique for linearly elastic materials but application on small nuclear graphite specimens demands further consideration. The purpose of this work is to develop this technique for use on irradiated graphite specimens whose size and geometry are restricted by the reactor core trepanning or Materials Test Reactor experiments. This paper describes the theoretical basis of the method, the work undertaken to prove the measurement technique and demonstrate its applicability to small irradiated and oxidised graphite specimens. Finally, the paper presents work of fracture values of irradiated graphite trepanned from AGR core bricks; for the first time, the relationship between work of fracture and other graphite properties, such as density, modulus and strength is experimentally demonstrated. [ABSTRACT FROM AUTHOR]

The diffusion of lanthanide fission products plays an important role in the growth of the fuel-cladding chemical interaction (FCCI) region in metallic fuels. The use of a Zr interdiffusion barrier may mitigate the transport of lanthanides from the fuel to the cladding, but the efficacy of such a liner is not yet known. In this paper, the stability and vacancy-mediated diffusion of La, Ce, Pr, and Nd in hexagonal close-packed (HCP) Zr is investigated via density functional theory (DFT) calculations and self-consistent mean field (SCMF) analysis. DFT is used to calculate the formation, binding, and migration energies of vacancies and vacancy-solute pairs. The DFT energetics are used in the KineCluE code to calculate the Onsager transport coefficients. La is found to be the fastest diffusing species in HCP Zr and experiences an almost isotropic diffusion behavior. The other three species (Ce, Pr, and Nd) demonstrate anisotropic diffusion where the diffusion in the basal planes is significantly faster than that along the c-axis. The calculated lanthanide diffusivities in HCP Zr are fitted to an Arrhenius relation and the activation energies and prefactors are reported for the first time. Furthermore, the vacancy drag and the segregation tendencies were analyzed using the calculated off-diagonal transport coefficients. According to our vacancy-mediated diffusion model, lanthanides are expected to be enriched at vacancy sinks at low temperatures, while at high temperatures, lanthanides are depleted at sinks and will preferably diffuse into the bulk. The enrichment/depletion transition temperature depends on the diffusion direction (basal or axial) and hence will be controlled by the grain texture and orientation. [ABSTRACT FROM AUTHOR]

An alloy-based fuel comprised of U-Pu-Zr is under development for the futuristic fast reactors to improve fissile atom utilization. As the technical data available with the metal fuel reactors is limited, it is imperative to have an exhaustive comprehension of behavior of fission and activation products before deploying the U-Pu-Zr metal fuel for commercial applications. Towards meeting this objective, the present paper highlights the analytical estimation of Zr and Am in the irradiated U-Zr metal that has experienced a low burn-up. The estimation of Zr was carried out by a modified spectrophotometric method developed in our laboratory. The estimation of Am was accomplished through initial liquid-liquid extraction-based removal of U and Pu by tri- n ‑butyl phosphate followed by the separation of Am by TODGA from the raffinate. Am was estimated with the help of alpha spectrometry of the Am-separated TODGA phase, and the same has been substantiated by comparing it with the results of gamma spectrometry. [ABSTRACT FROM AUTHOR]

• The effect of LBE on fatigue crack initiation was assessed through the analysis of surface microcracks in cylindrical samples tested in low cycle fatigue. • Marked differences in crack initiation (number of cracks and depth) were found for vacuum, air, and LBE environments. • Changes on sample stiffness due to surface cracks were estimated with a simple finite element model and correlated to experimental stress response behaviour. • Marked differences in the evolution of surface cracks were found for vacuum, air, and LBE environments through the comparison of surface cracks early in fatigue life and after fatigue failure. Active media such as liquid metals have an effect on the fatigue resistance of steels when compared to vacuum or inert environments. This effect might impact the initiation of fatigue cracks, their propagation, or both. This paper focusses on the initiation of fatigue cracks in an austenitic stainless steel when in contact with three environments: vacuum, air, and lead-bismuth eutectic (LBE), and the impact of such cracks on cyclic loading behaviour. Solid cylindrical samples of a 316L-type steel were fatigued under strain control in vacuum, air, and LBE in order to induce the initiation of cracks on their surface. The characteristics of fatigue damage in the form of surface microcracks were analysed with microscopy techniques. It was found that the nucleation of fatigue microcracks is enhanced by LBE, but the majority of these cracks do not propagate beyond the grain size of the steel (50 μm). An analysis with finite element methods showed that the large number of small (<10 μm) microcracks nucleated in LBE environments has a negligible impact on the sample's stiffness, as opposed to the fewer but deeper (100–500 μm) microcracks that initiate in air, which produce a loss of stiffness, observable on the mechanical stress response of the fatigue sample. A model based on the adsorption of LBE atoms along surface slip bands is proposed to explain the phenomenon of enhanced fatigue crack nucleation in LBE. [ABSTRACT FROM AUTHOR]

The phase transition behavior of RAFM steel during Hot Isostatic Pressing (HIP) diffusion bonding process would extensively impact the service performance of blanket component in the future fusion reactor. In this paper, the influence of cooling behavior on phase transformation and mechanical property of RAFM steel was studied using a combination of Electron Backscatter Diffraction (EBSD) and a thermal dilatometer. The results show that as the cooling rate decreases after austenitizing, the phase category transitions from a single-phase of martensite to the dual-phase of martensite and ferrite, along with an increase in the rate of ferrite, minor martensite packets and high-angle grain boundaries (HAGB). As the cooling rate decreases, the initial temperature for martensite transformation significantly increases, and the transformation driving force slowdown. Moreover, specimens processed with HIP cooling show lower strength, higher elongation, and a greater standard deviation of hardness compared to air cooling. These differences can be attributed to the formation of ferrite and the diffusion of carbon atoms within martensite. The research findings reveal the phase transition and microstructural evolution of martensite under varying cooling conditions, providing a reference for the controlling the cooling process after HIP bonding of blanket steel components in the future fusion reactor. [ABSTRACT FROM AUTHOR]

Pure iron has been irradiated with Fe2+ ions in a series of studies to identify neutron-atypical physical processes that influence the depth dependence of void swelling, focusing especially on suppression effects arising from injected interstitials and surface proximity. One paper in this series examined the surface influence in single-crystal Fe irradiated to 50 and 100 peak dpa over a range of temperature (425–525 °C) and ion energy (1.0, 2.5, 3.5, 5.0 MeV) while keeping the peak damage rate at 1.2 × 10–3 dpa/s, although the surface dpa rate was lower but increasing with decreasing ion energy, providing a small range of surface dpa rate. The observed denuded width Δ x was modified to incorporate sputtering loss. The activation energy governing the denuding process was found to be E Δ x 4 =1.65 ± 0.03 e V , higher than the vacancy migration energy E V m known to be 0.67 eV. This difference was attributed to the effect of dissolved carbon (103 appm) which reduces the effective vacancy mobility and thereby increases the effective migration energy. Since the previous study involved a factor of only 2.83 in near-surface dpa rate it is important to confirm that the dependence of E Δ x 4 on dpa rate is maintained over a larger range of damage rates. In this study polycrystalline Fe with 140 appm carbon was irradiated with 5 MeV Fe2+ ions to 50, 100 and 150 dpa over a range of peak dpa rates (2.0 × 10–4, 1.2 × 10–3, 6.0 × 10–3 dpa/sec) and temperatures (425, 475, 525 °C). At very high dpa levels sputtering and void growth lead to loss of voids via shrinkage, limiting the upper dose level where this technique can be applied. It was found that the single-crystal and polycrystal specimens yielded essentially identical behavior with E Δ x 4 = 1.65 eV, validating the application of this activation energy over a wider range of dpa, dpa rate, temperature, and crystal form. [ABSTRACT FROM AUTHOR]

Tritium breeder is an important functional material in d-T fusion reactor blanket serving the purpose of tritium re-production and energy conversion. Presently, Li 2 TiO 3 ceramics is considered as the most promising solid tritium breeder for its good chemical stability, high mechanical strength and excellent tritium release performance. However, low lithium content and lithium evaporation at high temperature lead to a low tritium breeding ratio. In the paper the neutron multiplier element, Pb, is introduced into Li 2 TiO 3 to obtain lithium lead titanate (Li 2 Pb x Ti 1-x O 3) as tritium breeder, which combined the advantages of Pb and Li 2 TiO 3 ceramic. Li 2 Pb x Ti 1-x O 3 powders were synthesized via a sol-gel method, employing lithium acetate, tetrabutyl titanate, and lead acetate, with varying lead contents. The characterization results indicate that at a molar ratio of titanium to lead of 1:2, Li 2 TiO 3 -Li 2 PbO 3 eutectic phase could be formed. The tritium breeding ratio (TBR) of Li 2 Pb x Ti 1-x O 3 materials were evaluated by Monte Carlo Code, which was notably influenced by the lead content, exhibiting a continuous increase with higher Pb content and neutron energy. The Li 2 Pb x Ti 1-x O 3 pebbles were prepared by freeze-drying route. The mechanical strength of Li 2 Pb x Ti 1-x O 3 pebbles could be affected significantly by the lead content. Compared to the conventional Li 2 TiO 3 ceramics, the Li 2 Pb 0.4 Ti 0.6 O 3 ceramics exhibits a higher crushing load of 25.33 N. [ABSTRACT FROM AUTHOR]

Chromium (Cr) has been recognized as a promising diffusion barrier candidate to mitigate the Fuel Cladding Chemical Interaction (FCCI) failure in metallic fuels for sodium-cooled fast reactor. This paper, for the first time, conducted an in-depth post-irradiation examination of the microstructure and composition evolution, and micromechanics of the Cr diffusion barrier in U-10Zr fuel/HT9 cladding irradiated in the Advanced Test Reactor (ATR) to 8.7% burnup at an averaged Peak Inner Cladding Temperature of 540–550 °C. Transmission Electron Microscope (TEM) characterization confirmed the preferential intergranular diffusion of Zr and U in the Cr diffusion barrier, suggesting that grain boundaries served as fast path for the diffusion of Zr and U into the Cr diffusion barrier. The interaction zone is dominated by nano crystalline α- ZrCr 2 Laves phase. Despite the interaction, there is no microcracks being observed in the preserved Cr diffusion barrier and HT9 cladding, serving as a good barrier to mitigate FCCI under the studied in-reactor irradiation conditions. High density cavities in uniform distribution are observed inside Cr grains, nano particles contains Cr, Mn, and O are confirmed by Electron Energy Loss Spectroscopy (EELS) analysis. However, it is unclear whether the cavities will become an issue to the barrier integrity at higher burnup. In-situ Scanning Electron Microscopy (SEM) micro-tensile testing uncovered mechanical softening in the HT9 cladding nearing the Cr diffusion barrier, possibly due to the coarsening of lath structure and carbides precipitates. [ABSTRACT FROM AUTHOR]

The significant microstructural changes that U-Mo fuel undergoes during operation degrades its mechanical properties and structural integrity. Microstructural evolution entails the formation, evolution, and redistribution of porosity in conjunction with grain refinement. In the present paper, we employ numerical approaches to assess the impact of the various microstructural features—grains, nanoscale intragranular fission gas bubbles, and mesoscale intergranular voids—on the degradation of elastic constants. Phase-field microstructure models are combined with the asymptotic expansion homogenization technique in order to derive the effective elastic constants as a function of porosity and fission density. The results are verified and compared against theoretical bounds. Using this approach, elastic degradation in operating nuclear fuels can be quantified when the distributions of microstructural features are known. [ABSTRACT FROM AUTHOR]

IFMIF-DONES is a powerful neutron source which is being designed with the main purpose of qualifying structural materials (EUROFER being the first candidate) to be used in DEMO and fusion power plants envisaged after it. This source relies on one high current (125 mA) deuterons beam accelerated to 40 MeV which impacts on a liquid lithium target to produce an intense neutron flux through Li(d,n) stripping reactions able to simulate the nuclear responses expected on the first wall of the reactor. The engineering design of IFMIF-DONES is presently being carried out mainly in the framework of the EUROfusion Work Package Early Neutron Source (WPENS). Since 2021, a new phase has started with the launch of the FP9 WPENS workplan whose objective is to continue advancing the engineering design of the facility, putting special effort on the experimental validation of those aspects which still need to be qualified to demonstrate the fulfillment of functional and safety requirements. The ENEA Brasimone Research Centre has been and still is strongly committed in several validation tasks concerned in particular with the lithium systems design and the Remote Handling (RH) maintenance. In this paper, an overview of the most relevant validation activities carried out in recent years or still in progress or planned at the ENEA Brasimone Research Centre in both of the aforementioned areas is presented, including the erosion/corrosion tests in the Lifus 6 loop; the nitrogen-gettering materials qualification in the ANGEL facility; and the RH and prototypes testing in the DRP laboratory for the validation of the High Flux Test Module electric connectors and the pre-heating of the Target Assembly. [ABSTRACT FROM AUTHOR]

In this paper, the microstructural changes and mechanical characteristics of alloy 690TT heat transfer tube with a triangular prism dent are investigated by combining experimental and simulation methods, and subsequently examines their influence on stress corrosion cracking (SCC) behavior. The results indicate that the deformation mechanism caused by impact dent is slip and twinning. The structural transformation of high-density dislocations, as well as Copper and Brass deformation textures, are formed at dent bottom, resulting in stress concentration through work hardening, which greatly increases SCC sensitivity. Oxidation preferentially occurs along grain boundaries and dislocations, and the corrosion products are formed, which exhibit a semi-coherent orientation relationship with the matrix. Oxidation is essentially a de-alloying process in which oxygen diffuses inward and Cr migrates towards the oxidation zone. The wedge force generated by corrosion products provides another driving force for initiation and propagation of SCC cracks, in addition to external loads and residual stress. SCC cracks propagate in the form of intergranular stress corrosion cracking and transgranular stress corrosion cracking in a high temperature lead-containing alkaline solution. [ABSTRACT FROM AUTHOR]

• A single high-temperature strength equation expressing the mechanical strength in different deformation and rupture modes was derived for 9Cr-ODS TMS cladding tubes. • This equation can predict the rupture life of the cladding tubes under various stresses and temperatures over time. • The reason why the equation can be applied to different deformation and rupture modes is considered to be the effect of the fine-grain matrix of 9Cr-ODS TMS. • The study suggested the thermal activation process is dominant even for the high-temperature deformation modes exceeding yield stress in the 9Cr-ODS TMS with the fine-grained matrix. This paper discusses the applicability of Straalsund et al.'s technique for combining the Larson–Miller parameter (LMP) and life-fraction rule to form a single high-temperature strength equation for 9Cr-oxide-dispersion-strengthened (ODS) tempered martensitic steels (TMS). It uses the extensive dataset on creep rupture, tensile, and temperature-transient-to-burst tests of 9Cr-ODS TMS cladding tubes in the α-phase, α/γ-duplex, γ-phase matrices, which are accumulated by the Japan Atomic Energy Agency so far. The technique is adequately applicable to 9Cr-ODS TMS cladding tubes. A single high-temperature strength equation expressing the mechanical strength in different deformation and rupture modes (creep, tensile, temperature-transient-to-burst) is derived for 9Cr-ODS TMS cladding tubes. This equation can predict the rupture life of the cladding tubes under various stresses and temperatures over time. The applicable range of the high-temperature strength equation is specified in this study and the upper limit temperature for the equation is found to be 1200 °C. At temperatures higher than 1200 °C, the coarsening and aggregation of nanosized oxide particles and the γ to δ phase transformation are reported in previous studies. The high-temperature strength equation can be well applied to the creep and tensile strength in the α-phase matrix, the creep strength in the γ-phase matrix and the temperature-transient-to-burst strength in both phases except for the low equivalent stress (43 MPa) at temperatures exceeding 1050 °C. The mechanism of the notable consistency between creep and tensile strength in the α-phase matrix is discussed by analyzing the high-temperature deformation data in the light of existing deformation models. The study suggested the thermal activation process is dominant even for the high-temperature deformation modes exceeding yield stress in the 9Cr-ODS TMS with the fine-grained matrix. [ABSTRACT FROM AUTHOR]

The characterization of zirconium hydride is important in the nuclear industry because of the hydrogen-induced embrittlement of Zircaloy cladding and its use as a neutron moderator. This paper introduces the use of Raman spectroscopy for the characterization of zirconium hydride. First-principles density functional theory (DFT) calculations were used to predict the Raman spectra of ζ-ZrH 0.5 , γ-ZrH, δ-ZrH 1.5 , δ-ZrH 1.66 , and ε-ZrH 2 with all their predicted symmetries; ζ-ZrH 0.5 (P3m1, R 3 ¯ m, C2/m, Cm, Cmmm , and Pn 3 ¯ m); γ-ZrH (P222, Ccce , and P4 2 /mmc); δ-ZrH 1.5 (P 4 ¯ m2, P4 2 /mcm, Fmmm, Pn 3 ¯ m, Ibam, P2/c, PI , and P4 2 /nnm); δ-ZrH 1.66 (Fm 3 ¯ m); and ε-ZrH 2 (Fm 3 ¯ m, R 3 ¯ m , and I 4 /mmm). Two samples of Zircaloy-4 containing 133 wt ppm and 360 wt ppm hydrogen were characterized by Raman spectroscopy, showing two signal lines at 215 cm−1 and 1,187 cm−1, which were assigned to the presence of δ-ZrH 1.66. These signals had a good spatial correlation with visible hydride precipitates in Raman spectroscopy images. This work provides the basis for the characterization of all possible zirconium hydride compositions and structures using Raman spectroscopy. [ABSTRACT FROM AUTHOR]

In this paper, the volume control spark plasma sintering tool has been designed and applied to sinter porous U-10Zr metallic fuels, by which the sintered sample volume can be precisely controlled. Ethanol and NH 4 HCO 3 are used to control the powder compact or as pore formers to control the pore size and pore structure. Without pore formers, the fuel pellet displays an inhomogeneous microstructure consisting of highly porous and highly densified areas. Uneven powder stacking in the green body results in a non-uniform microstructure, and in the closed-packed area, Joule heating accelerates the neck formation and densification. The addition of ethanol reduces the friction between the powders, resulting in isolated pores formed by the stacking of powders during the sintering. By adding NH 4 HCO 3 , the pore size, and structure can be well controlled, and an interconnected pore structure can be obtained upon the decomposition of the NH 4 HCO 3. A uniform microstructure and pore distributions can be achieved through the U-10Zr fuel pellets by controlling current flow during the volume control SPS sintering. The microstructure and phase characterization of the sintered porous U-10Zr pellets show major phases of α-U and α-Zr for the sample with short dwelling. For the sample with long dwelling (30 min), the ω UZr 2 in the Zr-enriched area has been observed. The strategy of volume control SPS sintering with the assistance of pore formers could be used to fabricate porous U-10Zr metallic fuels to mimic the microstructure evolution of irradiated metallic fuels (including porosity) and could enable a possible solution for the design of new sodium-free metallic fuels for high burnup. [ABSTRACT FROM AUTHOR]

In this paper, the reaction of SrO in molten chloride and fluoride systems is investigated using Raman spectroscopy, combined with density functional theory calculations. The formation of Sr 4 OCl 6 was observed in various salt compositions, including LiCl, LiCl-KCl, and LiCl-LiF, at temperatures ranging from 298 K to 923 K. Thermodynamic calculations estimated that Sr 4 OCl 6 is more stable than SrO and SrCl 2 , while no stable strontium oxyfluoride compound was found. The results were also supported by X-ray analyses: Extended X-ray absorption fine structure analysis indicated that the local structure of Sr in LiCl-KCl and LiCl-LiF consists of one oxygen and six chlorides, corresponding to a complex with a similar local structure of Sr 4 OCl 6. Additionally, Sr 4 OCl 6 was found in LiCl-LiF, and no strontium compound was observed in LiF-KF using X-ray diffraction. These results provide a comprehensive understanding of the dissolution structures of strontium oxyhalide in molten chloride and fluoride systems. [ABSTRACT FROM AUTHOR]

Evaluating the mole fraction of hydrogen isotopes in a solid is a difficult task. Few methods allow it to be achieved. LIBS is a laser method based on the electronic excitation of elements and the spontaneous emission of characteristic optical lines. On a sample containing hydrogen isotopes, α -type lines can allow the estimate of their total mole fraction. In addition, LIBS is a discriminating method because it can separate the contributions of isotopes. This paper reports the implementation of this method on thin film-type samples containing hydrogen and tritium. They consist of nanometric layers of palladium and titanium on a silicon substrate. Under irradiation of nanosecond laser pulses reaching a fluence of the order of 200 J cm−2, LIBS was performed in argon at atmospheric pressure. A detailed spectroscopic study is performed around the T α line of wavelength 656.039 nm of the Balmer series (n = 3 → 2) of tritium. An analysis based on the reconstruction of the spectrum under conditions of local thermodynamic equilibrium is carried out. This leads to the estimate of the tritium mole fraction. [ABSTRACT FROM AUTHOR]

• Calorimetric phase transition peak data is reported for three U-Pu-Zr alloys. • Complex peaks were deconvoluted using Frasier–Suzuki peak fitting algorithm. • Critically compared DSC data, microstructure and phase characterization data with existing phase diagrams. • Increasing Pu concentration generally promotes stabilization of bcc-γ phase. • Constructed U-Pu-40 at.% Zr pseudobinary using TAF-ID database Ternary alloys consisting primarily of uranium, plutonium, and zirconium (U-Pu-Zr) are among the leading candidate fuel systems considered for fast spectrum nuclear reactors. Despite historical operation data from the testing of U-Pu-Zr rods in the Experimental Breeder Reactor-II, considerable uncertainty about the evolution of phases and microstructure across the ternary composition space exists. Due to sluggish kinetics and other difficulties in handling metal actinide specimens, quantitative measurements of phase-transitions in U-Pu-Zr alloys remain sparse in scientific literature, with most investigators reporting either phase-transition temperatures or phase identification data, but not both from the same specimens. The purpose of this paper is to critically compare experimental and calculated phase transition data and correlate with the microstructure and phase characterization data of as-cast and annealed U-Pu-Zr alloys. Phase transition peaks were measured using differential scanning calorimetry in the subsolidus regions (723−948 K) of three ternary U-Pu-Zr alloys with the same zirconium concentration but various U/Pu ratios. Overlapping peaks were deconvoluted using a Frazier-Suzuki peak fitting algorithm, and the critical peak temperatures and enthalpies were calculated. In general, increasing concentrations of Pu were associated with enhanced thermal stability of the body-centered cubic γ phase upon both heating and cooling. Experimental phase transition temperatures in this study tended to agree well with the predictions of the established ternary phase diagrams and other reported phase transition temperatures in literature. Additionally, the TAF-ID thermodynamic database was used to calculate a U-Pu-40 at.% Zr pseudobinary diagram as well as ternary diagrams from 773 to 973 K. The equilibrium phase transition temperatures tended to be considerably lower than measured peak temperatures upon both heating and cooling. Recommendations for improving the quality of data in future U-Pu-Zr characterization studies are also discussed. [Display omitted] [ABSTRACT FROM AUTHOR]

This paper describes the first VERDON test performed at the end of September 2011 with special emphasis on the behaviour of fission products (FP) and actinides during the accidental sequence itself. Two other papers discuss in detail the post-test examination results (SEM, EPMA and SIMS) of the VERDON-1 sample. The first VERDON test was devoted to studying UO 2 fuel behaviour and fission product releases under reducing conditions at very high temperature (∼2883 K), which was able to confirm the very good performance of the VERDON loop. The fuel sample did not lose its integrity during this test. According to the FP behaviour measured by the online gamma station (fuel sight), the general classification of the FP in relation to their released fraction is very accurate, and the burn-up effect on the release rate is clearly highlighted. [ABSTRACT FROM AUTHOR]

The spherical fuel element (SFE) dispersed with many tri-isotropic fuel (TRISO) particles has been primarily employed in the high-temperature gas-cooled reactor. Owing to the complex structure of SFE, behaviors will be evolved for both TRISO particles and the fuel matrix with the increasing burnup. To evaluate the fuel performance of SFE, the homogenization method has been conventionally adopted for SFE scale modeling, and dispersed TRISO particles are analyzed subsequently with fixed temperature boundary conditions. However, owing to the varying properties of TRISO particles under reactor operation, the homogenization models dedicated to certain circumstances can hardly reflect the properties evolution of TRISO with high fidelity, which may lead to uncertainties in analysis results. To this end, a hybrid model, namely the conventional homogenization model informed by the lower scale model online, has been developed in this paper, which may increase modeling fidelity for both SFE and particles. Specifically, property models for the TRISO particle and graphite matrix have been implemented in COMSOL. Subsequently, based on the linking of COMSOL and MATLAB, the hybrid model for SFE analysis was established. Through the hybrid model, the temperature field and stress distribution of TRISO, the evolution of effective properties, and the thermal and mechanical state of the SFE were obtained and analyzed. The results show that within the scope of this paper, the central temperature of TRISO in the innermost region rises with the increasing burnup for the decreasing thermal conductivity of materials, and the peak value is about 1324 K, which is about 60 K higher than that in the outermost region. The difference in the maximum plenum pressure is more than 5 MPa among TRISO particles, which is mainly caused by the variations in CO production. Consequently, the hoop stress of SiC in the outermost region is slightly elevated. The distribution of effective thermal conductivity (ETC) for TRISO is nearly 4.3 W/(m·K) at the beginning of life (BOL), while it was reduced to 3.6 W/(m·K) at the end of life (EOL). There are marginal variations along the radial direction for the ETC of TRISO. Meanwhile, the ETC of SFE decreases from about 21 W/(m·K) for fresh fuel to 13 W/(m·K) for spent fuel. The predicted effective elastic modulus and coefficient of thermal expansion of SFE increase gradually versus time. The former is caused by the densification of pyrolytic carbon, and the latter is mainly determined by the high volume fraction of matrix and the changes of CTE for pyrolytic carbon. In addition, the comparison in the temperature field between the proposed model and the Differential Effective Material Theory (D-EMT) model may further demonstrate the feasibility and effectiveness of the hybrid model. [ABSTRACT FROM AUTHOR]

The UCSB database on cavity evolution in 9-12Cr tempered martensitic steels (TMS), includes the results for both dual heavy and helium ion (DII), and High Flux Isotope Reactor (HFIR) in situ helium injection (ISHI) neutron irradiations at 500°C. These results were combined with literature single ion and fission neutron irradiation data to derive a model for the void volume fraction, f v , as a function of displacements per atom (dpa) and transmutant helium concentrations in atomic parts per million (appm). The scientific foundation for the paper is described in a companion paper entitled "Cavity Evolution and Void Swelling in Dual Ion Irradiated Tempered Martensitic Steels ". Here, we show that f v (dpa, He/dpa) is described by the incubation dose, dpa i , for the onset of void growth, and the post-incubation growth rate, f v '(%/dpa). Both dpa i and f v ' decrease with increasing He/dpa at > ∼ 5. The dpa i is also lower for the ISHI neutron irradiations at the same He/dpa. Single heavy ion and fission reactor neutron irradiations, with low He/dpa ratios, have a much larger dpa i. Based on a combined analysis of DII, single ion, ISHI and fission neutron data, we further show that the post-incubation f v data analyzed here have a common empirical curve shape, with f v ' reaching up to ∼ 0.2%/dpa at very high dpa. We also show that f v ' can be predicted based on a physical model of defect partitioning between evolving sinks. At 500°C and fusion relevant He/dpa ≈ 10, the best-fit model predicts nominal swelling, S = f v /(1-f v), of ∼ 1.1, 4.9 and 16% at 50, 100 and 200 dpa, respectively. The physically motivated, data-driven model includes estimated uncertainties for both dpa i and f v '. [ABSTRACT FROM AUTHOR]

In order to explore the core corrosion characteristics of horizontal lead-bismuth reactor in oxygen-controlled environment, a liquid lead-bismuth eutectic (LBE) corrosion model including the growth and removal process of oxide layer was established in this paper. The coupling method of double-layered oxidation model and computational fluid dynamics (CFD) was used to simulate the long-term oxidation corrosion of horizontal LBE reactor core numerically. The results show that the average total oxide layer thickness of fuel rod surface is 1.59 μm, and the thickest oxide layer is 3.48 μm on the fuel rod surface at the core outlet after 20,000 h. The fuel rods maintain a double-layer oxide layer structure. The growth rate and removal rate of the oxide layer on fuel rod surface increase with the increase of inlet temperature and the decrease of inlet flow rate. The growth rate of the oxide layer on fuel rod surface increases with the increase of inlet oxygen concentration, while the removal rate decreases. At higher temperature above 420℃, lower flow velocity below 0.25 m/s and lower oxygen concentration below 2 × 10−8wt%, the magnetite layer will first be completely removed from the surface of the fuel rods at the core outlet. The liquid LBE oxidation corrosion model and coupled CFD method presented in this paper can be used for numerical simulation of oxidation corrosion in different units of LBE cooling system. [ABSTRACT FROM AUTHOR]

• A review of data on phase equilibria in the U-Pb-Bi-Fe-Cr-Ni-N system has been carried out. • The Bi-Fe-Pb ternary phase diagram has not been built and must be constructed. • It is required to study the influence of steel alloying components on phase equilibria in the UN-Fe pseudobinary system. • The phase relationships in the Bi-Pb-U system are studied, but there are no studies on the high-temperature interaction of UN with Fe, Pb, Bi. • The review information may be useful for understanding the physicochemical interactions between the new reactor materials. This paper systematizes information on phase equilibria in the multicomponent system Pb-Bi-Fe-Cr-Ni-U-N. The components of this system correspond to the interaction of stainless steel fuel cladding, nitride fuel and heavy liquid metal coolant (HLMC) with each other under conditions of a severe accident. The review touches upon the temperature thresholds for the stability of materials in the core of a nuclear reactor. The phase diagrams in the multicomponent Pb-Bi-Fe-Cr-Ni-U-N system are discussed. Gaps and contradictions in the available information about phase equilibria in subsystems are considered. [Display omitted] [ABSTRACT FROM AUTHOR]

• Anti-LBE corrosion FeCrAlW x HEA coating was prepared by selective laser melting. • FeCrAlW 0.75 HEA coating has outstanding LBE corrosion resistance. • LBE corrosion protection of FeCrAlW 0.75 HEA coating and its mechanism analysis. • FeCrAlW 0.75 HEA coating exhibits good mechanical properties after LBE corrosion. The use of ferritic/martensitic (F/M) steel as a container material for liquid lead–bismuth eutectic (LBE) presents several challenges owing to the complex and diverse types of LBE corrosion, including dissolution corrosion and oxidation corrosion. This paper proposes a novel strategy to address these challenges, based on the optimization of the design and composition of FeCrAlW x high-entropy alloy (HEA) coating. The objective is to simultaneously enhance the resistance of F/M steel to both dissolution corrosion and oxidation corrosion. Following LBE corrosion, the surface of the coating exhibits oxide layers consisting of an outer oxide layer (OOL, composed of Fe oxides) and an inner oxide layer (IOL, composed of Al 2 O 3 and Cr 2 O 3). Among the tested samples, FeCrAlW 0.75 HEA coating exhibits the thinnest OOL and IOL layers when subjected to LBE corrosion over different durations, demonstrating superior resistance against LBE corrosion. The incorporation of W in the coating increases the covalency of the HEA coating, thereby improving the bonding strength. In addition, it decreases the adsorption energy of Pb and Bi on the surface, thereby effectively limiting the mutual diffusion of the coating elements and Pb and Bi and inhibiting dissolution corrosion. Furthermore, the IOL formed by Al and Cr exhibits protective properties and improves the oxidation corrosion resistance of the coating. These effects are further enhanced by the high adsorption capacity of the FeCrAlW 0.75 HEA coating to O, which helps accelerate the formation of the IOL. Additionally, the FeCrAlW 0.75 HEA coating retains satisfactory mechanical properties after being subjected to LBE corrosion for 2000 h, with the Hv, E, Hv / E, and Hv 3/ E 2 values being 6.92 GPa, 197.2 GPa, 0.0351, and 0.00852, respectively. "For Table of Contents Use Only." [Display omitted] [ABSTRACT FROM AUTHOR]