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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]

Abstract: The paper summarizes original results of irradiation embrittlement study of EUROFER 97 material that has been proposed as one candidate of structural materials for future fusion energy systems and GEN IV. Test specimens were manufactured from base metal as well as from weld metal and tested in initial unirradiated condition and also after neutron irradiation. Irradiation embrittlement was characterized by testing of toughness properties at transition temperature region – static fracture toughness and dynamic fracture toughness properties, all in sub-size three-point bend specimens (27×4×3mm3). Testing and evaluation was performed in accordance with ASTM and ESIS standards, fracture toughness K JC and K Jd data were also evaluated with the “Master curve” approach. Moreover, J–R dependencies were determined and analyzed. The paper compares unirradiated and irradiated properties as well as changes in transition temperature shifts of these material parameters. Discussion about the correlation between static and dynamic properties is also given. Results from irradiation of EUROFER 97 show that this steel – base metal as well as weld metal – is suitable as a structural material for reactor pressure vessels of innovative nuclear systems – fusion energy systems and GEN IV. Transition temperature shifts after neutron irradiation by 2.5dpa dose show a good agreement in the case of EUROFER 97 base material for both static and dynamic fracture toughness tests. From the results it can be concluded that there is a low sensitivity of weld metal to neutron irradiation embrittlement in comparison with EUROFER 97 base metal. [Copyright &y& Elsevier]

Abstract In this paper, the evolution of cracks from the surface of a thick oxide layer on Zircaloy-4 cladding has been analyzed for the normal operation of pressurized-water reactors. The conditions for the propagation of radial cracks towards the interface, channeling along the axis of the cladding, and the possible subsequent spalling of the oxide have been studied. The analysis was conducted by first calculating how the stresses in the oxide developed during operation, using a numerical model that incorporates multiple mechanisms such as creep, swelling and oxidation. These calculations demonstrate that the circumferential stresses within the oxide increase as oxidation proceeds and as the cladding expands under the effect of fuel swelling. Although the intrinsic growth stresses of the oxide are compressive, tensile stresses can eventually develop in the outer region of the oxide. Within this regime, the energy-release rate for the radial propagation of a crack was determined using the J -integral along with the calculated stress profile. By assuming a suitable value of toughness for the oxide, it is possible to determine the depth to which such a radial crack can grow. A second fracture-mechanics calculation was then conducted to explore the conditions under which such a surface crack can subsequently channel along the axial direction of the cladding. A third fracture-mechanics analysis considered oxide spalling. This showed the possibility of spalling from the radial cracks, which leaves a thin layer of the oxide adhered to the metal that can result in a local cold spot responsible for hydride formation. Graphical abstract Image 1 Highlights • Model of crack evolution in the oxide layer of cladding and oxide spalling. • Outer region of oxide layer becomes tensile after a critical oxide thickness. • Newly formed oxide induces a tensile contribution to the oxide formed earlier. • Three types of crack may form depending on oxide thickness and toughness. • A radial crack may channel; spall depth is typically smaller than oxide thickness. [ABSTRACT FROM AUTHOR]

Abstract: This work comments on Kim’s rebuttal of this authors’ critical review that is cited in the title of this paper. The author shows that Kim’s rebuttal of the author’s review is based on essentially the same physical misconceptions concerning the driving force for DHC rate in the Dutton–Puls model as are given in Kim’s original papers. The author shows that his previous criticisms of Kim’s model as well as the author’s defense of the correctness of the thermodynamic basis of the various versions of the Dutton–Puls models have been validated. In contradistinction, the author shows that it is Kim’s DHC model that has a faulty thermodynamic basis and, therefore, is not viable in underpinning the physical understanding of DHC rate. To make the reasons for these conclusions clearer than given previously, the author has attached an appendix to these comments that provides detailed steps of the mathematical derivations of the various versions of the Dutton–Puls models. This work also provides an interpretation of the results of the experiments of Kammenzind et al. (2000) that shows that these results can be interpreted according to the Dutton–Puls mechanism for hydride growth. Thus the results of the Kammenzind and co-workers’ tests do not provide support, as Kim claims, that a stress gradient in a closed system cannot cause hydrogen concentration redistribution. [Copyright &y& Elsevier]

Abstract: The aim of this paper is a reply to McRae et al.’s paper entitled “The first step for delayed hydride cracking (DHC) in zirconium alloys” claiming that the first step of DHC is hydrogen diffusion, not nucleation of hydrides as demonstrated by Kim’s new model. Despite the authors’ claim that the crack tip concentration is higher than the bulk concentration due to the stress gradient, their claim violates the thermodynamic principle that the stressed region should have a lower potential of hydrogen or lower hydrogen solubility than the unstressed region. Furthermore, it is demonstrated that the Diffusion First Model (DFM) proposed by the author is defective in terms of kinetics because hydrogen diffusion before hydride nucleation just governs the rate of hydride nucleation, neither the rate of hydride growth nor the crack growth rate (CGR). [Copyright &y& Elsevier]

Abstract: This paper describes a finite element based fracture mechanics model to assess how hydrides affect the integrity of zircaloy cladding tubes. The hydrides are assumed to fracture at a low load whereas the propagation of the fractured hydrides in the matrix material and failure of the tube is controlled by non-linear fracture mechanics and plastic collapse of the ligaments between the hydrides. The paper quantifies the relative importance of hydride geometrical parameters such as size, orientation and location of individual hydrides and interaction between adjacent hydrides. The paper also presents analyses for some different and representative multi-hydride configurations. The model is adaptable to general and complex crack configurations and can therefore be used to assess realistic hydride configurations. The mechanism of cladding failure is by plastic collapse of ligaments between interacting fractured hydrides. The results show that the integrity can be drastically reduced when several radial hydrides form continuous patterns. [Copyright &y& Elsevier]

In this work, the hoop fracture toughness of ZIRLO ® fuel cladding is calculated as a function of three parameters: hydrogen concentration, temperature and displacement rate. To this end, pre-hydrided samples with nominal hydrogen concentrations of 0 (as-received), 150, 250, 500, 1200 and 2000 ppm were prepared. Hydrogen was precipitated as zirconium hydrides in the shape of platelets oriented along the hoop direction. Ring Compression Tests (RCTs) were conducted at three temperatures (20, 135 and 300 °C) and two displacement rates (0.5 and 100 mm/min). A new method has been proposed in this paper which allows the determination of fracture toughness from ring compression tests. The proposed method combines the experimental results, the cohesive crack model, finite elements simulations, numerical calculations and non-linear optimization techniques. The parameters of the cohesive crack model were calculated by minimizing the difference between the experimental data and the numerical results. An almost perfect fitting of the experimental results is achieved by this method. In addition, an estimation of the error in the calculated fracture toughness is also provided. [ABSTRACT FROM AUTHOR]

A fuel cladding mechanical test, performed under conditions of plane strain deformation in the transverse direction of tube axis, was originally developed at Pennsylvania State University. It was decided to implement this original test within the PROMETRA program using the same experimental procedure and its optimization for a ring mechanical testing on plane strain conditions (PST tests) in hot cells laboratory. This paper presents a detailed description and an interpretation of the Plane Strain Tensile (PST) tests performed in the framework of the PROMETRA program on fresh and irradiated claddings. At first, the context of the PST tests is situated and the specificities of these tests implemented at CEA are justified. Indeed, a significant adjustment of the original experimental procedure is carried out in order to test the irradiated fuel cladding in the best possible conditions. Then, the tests results on fresh Zircaloy-4 and on irradiated Zircaloy-4, M5™ and ZIRLO ® specimens are gathered. The main analyses in support of these tests, such as metallographies, fractographic examinations and finite element simulations are detailed. Finally, a synthesis of the interpretation of the tests is proposed. The PST test seems only representative of plane strain fracture conditions when the test material is very ductile (fresh or high temperature or low hydride material like M5TM). However, it provides a relevant representation of the RIA rupture initiation which is observed in irradiated cladding resulting from hydride rim damage due to the strong irradiation of a fuel rod. [ABSTRACT FROM AUTHOR]

Zirconium is a kind of deuterium absorption material and can be used as a deuterium carrier in the nuclear industry. However, zirconium usually cracks after deuterium absorption and therefore fails in its application. In this paper, the effects of deuterium absorption temperature, deuterium absorption pressure and interrupted cooling on the cracking of deuterium absorption Zircaloy-4 alloys have been investigated. The results show that as the deuterium absorption temperature decreases from 900 °C to 800 °C or 700 °C, the deuterium absorption amounts and their tendency to crack increase. And with the increase of deuterium absorption pressure, the deuteride types change from δ -ZrD 1.66 to ε -ZrD 1.8 , which is believed to be the key factor in the cracking. Furthermore, the interrupted cooling also has a significant effect on the deuterium absorption cracking. Based on the experimental results, an explanation of why Zircaloy-4 alloy cracks after deuterium absorption is proposed. [ABSTRACT FROM AUTHOR]

Zirconium oxide layers are submitted to internal stresses that play a role in damage of the layer. Lateral cracking is often observed during Zr alloys oxidation. In this paper, we investigated the influence of the microstresses at the crystalline scale on the lateral cracking within a growing oxide on a plane substrate. A parametric study was carried out taking into account the crystallographic texture of the oxide and the presence of a tetragonal zirconia at the metal–oxide interface. Macroscopic computations and polycrystalline aggregate computations were performed. The result indicating the (1 0 6 ¯ ) fiber texture as the most favorable was recovered. It was found that under macroscopic compressive stresses parallel to the plane metal–oxide interface, positive microstresses perpendicular to the interface develops. They can trigger the lateral cracking and the phenomenon is promoted by the presence of tetragonal zirconia at the metal–oxide interface. [ABSTRACT FROM AUTHOR]

In this paper, the behavior of a dense UO 2 (porosity less than 2%) was studied experimentally on a range of temperatures (1100–1700 °C) and strain rates ( 10 - 4 – 10 - 1 /s) representative of RIA loading conditions. The yield stress was found to increase with strain rate and to decrease with temperature. Macroscopic cracking of the samples was apparent after the tests at 1100 °C. Scanning Electron Microscopy (SEM) image analyses revealed a pronounced grain boundary cracking in the core of the samples tested at 10 - 1 /s and at 1550–1700 °C. A hyperbolic sine model for the viscoplastic strain rate with a clear dependency on porosity was first developed. It was completed by a Drucker–Prager yield criterion with associated plastic flow to account for the porosity increase induced by grain boundary cracking. Finite Elements simulations of the compression tests on the dense UO 2 were then successfully compared to the stress–strain curves, post-test diameter profiles and porosities at the pellets’ center, periphery and top extremity. The response of the grain boundary cracking model was then studied in biaxial compression, this condition being closer to that of the pellet during a RIA power transient. [ABSTRACT FROM AUTHOR]

Delayed hydride crack (DHC) growth study was carried out on irradiated Indian Zr–2.5Nb pressure tube which had seen around 8 effective full power years of operation. Disc compact tension type specimens were used for the DHC tests at 210 °C, 250 °C, 265 °C and 290 °C. This paper discusses the test methodology, results generated and compares it with that obtained on the as-fabricated pressure tube of similar specification. [ABSTRACT FROM AUTHOR]

Fracture toughness of irradiated Zr–2.5Nb alloy pressure tube, fabricated by the cold pilgering and stress relieving route, was evaluated using disk compact tension type specimens. These specimens were punched out from the irradiated pressure tube (S-07), which was in service for about 8 effective full power years of reactor operation in the Kakrapar Atomic Power Station-2 (KAPS-2). The tests were carried out remotely inside a lead shielded enclosure. Crack growth during the test was measured using the direct current potential drop technique. The irradiated pressure tube showed low fracture toughness at 25 °C. The fracture toughness increased with increase in temperature up to 250 °C but was practically unaffected with further increase in temperature up to 300 °C. This paper discusses the fracture behavior of irradiated Indian pressure tube material and compares it with other data available. [ABSTRACT FROM AUTHOR]

Physical mechanisms at different length scales have to be taken into account while predicting the overall failure of nuclear graphite structures of advanced gas cooled graphite reactors. In this paper, the effect of composition of meso graphite phases and porosity on the aggregate elastic properties is predicted using the Eshelby homogenisation method. Results indicate an overall decrease in elastic modulus with an increase in porosity. Subsequently, the moduli at different porosity levels are used to predict the critical strain energy release rates for crack propagation of graphite, and fracture behaviour is studied using compact tension and four point bending tests. Compared to flexural strength at zero porosity level, significant reduction in strength of up to 80% at 30% porosity level is observed. Evolution of flexural strength due to porosity is also compared against available experimental values of graphite from UK nuclear plants. [ABSTRACT FROM AUTHOR]

Abstract: Delayed hydride cracking (DHC) is an important failure mechanism that may occur in Zr alloys during service in water-cooled reactors. Two conditions must be attained to initiate DHC from a crack: the stress intensity factor must be higher than a threshold value called KIH and, hydrogen concentration must exceed a critical value. Currently the pressure tubes for CANDU reactor are fabricated from Zr–2.5Nb. In this paper the critical hydrogen concentration for DHC and the crack velocity of a developmental pressure tube, Excel, was evaluated and compared with that of Zr–2.5Nb. The DHC velocity values measured in Excel were higher than usually reported in Zr–2.5Nb. Due to the higher hydrogen solubility limits in Excel, its critical hydrogen concentration for DHC initiation is 10–50wppm over that of Zr–2.5Nb in the range of 150–300°C. [Copyright &y& Elsevier]

Abstract: This paper presents the development and validation of an empirical model of combined fast neutron damage and radiolytic oxidation effects on Young’s modulus for the Gilsocarbon graphites used in advanced gas-cooled reactors. The combined irradiation and oxidation model is based on a model of fast neutron damage in an inert environment and the degree of radiolytic oxidation as measured by weight loss. The change in Young’s modulus with exposure was found to vary from one power reactor to another and with the relative contributions of oxidation and fast neutron damage in data from materials test reactors. The model was calibrated to data from material trepanned from advanced gas-cooled reactor moderator bricks after varying operating times. The model also predicts materials test reactor data that extend beyond the range of currently-available trepanned data. The model is convenient to use because it is an analytic function that eliminates the interpolation on temperature and dose that is used in an earlier model. [Copyright &y& Elsevier]

Abstract: Ferritic 14%Cr and 18%Cr ODS steels produced at CEA in round bars or plates were tested mechanically. The present paper reports results obtained in tension, impact, fatigue, creep and toughness tests. These tests were carried out at various temperatures and in different directions. These materials show a pronounced anisotropy at all tested temperatures. No matter the loading, the transversal direction is always found to be far less resistant than the longitudinal one. This anisotropy is mainly observed in terms of damage mechanisms, with intergranular fracture preferentially occurring along the extrusion direction. This intergranular fracture mode leads to very low and anisotropic toughness values and to the absence of tertiairy creep stage, pointing out the unstable nature of fracture, even at high temperature. The unrealistically high values of the Norton exponent measured in creep suggests the existence of a threshold stress, which is consistent with the mainly kinematic nature of the stress as revealed by fatigue tests. [Copyright &y& Elsevier]

Abstract: This paper deals with the effect of irradiation on tensile properties and fracture toughness of CuCrZr and CuCrNiSi alloys, considered for use in some in-vessel components of ITER, where a combination of high strength and heat conduction is essential. The heat treatments were: [–] CuCrZr, quenching in water after annealing at 950°С, cold worked 40–45%, and aged at 475–500°С for 3h. [–] CuCrNiSi, quenching in water after annealing at 980°С and aged for 4h at 460°С. Specimens were irradiated in the IVV-2 reactor at ∼200°С and with irradiation damage of 0.15 and 0.27dpa. Post-irradiation tests were carried out to assess the tensile properties and fracture toughness of the materials. The tests results show that CuCrNiSi has better strength and retains higher ductility after irradiation, but has somewhat lower crack resistance than CuCrZr. [Copyright &y& Elsevier]

Abstract: Stress corrosion cracking (SCC) is enhanced by cold work and causes many problems in components of the nuclear power plants. Besides, during manufacturing, installation, welding and service of the material, residual strains can be produced increasing the susceptibility to SCC. For this reason, it is important to characterize the degree of plastic strain due to dislocation accumulation in each crystal. Electron backscatter diffraction (EBSD), in conjunction with scanning electron microscope (SEM), has been a great advance in this field because it enables to estimate the plastic strain in a quick and easy way. Nevertheless, over the last few years, a lot of different mathematical expressions to estimate the plastic strain have appeared in the literature. This situation hinders the election of one of them by a novel scientist in this field. Therefore, in this paper some of the more common expressions used in the calculation of the angular misorientation have been presented and discussed in order to clarify their more important aspects. Then, using one of these expressions (average local misorientation), curves relating misorientation density with known levels of strain will be obtained for an austenitic stainless steel 304L and nickel base alloy 690, which have shown a linear behaviour that is in good agreement with results found in the literature. Finally, using curves obtained in previous steps, levels of plastic strain in a plate of nickel base alloy 600 welded with weld metal 182 were estimated between 8 and 10% for a high temperature mill annealing sample. [Copyright &y& Elsevier]

Abstract: This paper presents the results of molecular dynamics modeling of quasi-brittle microcrack propagation on the (001) plane of bcc iron. The nucleation and propagation of such cracks is accompanied by the emission of slip bands from the crack tips. The 3D nature of the model, in conjunction with a Voronoi decomposition post-processing technique, allows us to clarify the atomistic structure of slip bands and to follow the detailed kinetics of slip band expansion and interaction with other microstructural defects. The effects of simulation temperature and alternative defects (in particular, vacancies and micro-voids) on the crack propagation kinetics are discussed. [Copyright &y& Elsevier]

Abstract: The paper presents the results of molecular dynamics (MD) simulations of three-dimensional kinetics of micro-crack propagation in alpha-iron and the accompanying lattice transformations at the crack tips. We show that crack initiation on {001} planes in iron is preceded with the emission of compact slip bands from the pre-crack tips, in agreement with the predictions of the earlier quasi-two-dimensional simulations. The application of Voronoi decomposition technique for atomic short-range order processing has allowed us to clarify the kinetics of structural transformations at the tips of nucleating and propagating cracks for three most common systems of crack propagation in iron. It is demonstrated that the compact slip bands emanating from the crack tips not only accompany crack nucleation, but remain an essential feature of the crack propagation on {001} planes. Due to the strong coupling between the crack tip and slip band propagation, the crack propagation can be limited by slip band interaction with microstructural obstacles, abundantly created in ferritic–martensitic steels in radiation environment of nuclear facilities. [Copyright &y& Elsevier]

Abstract: Graphite will be used as a structural and moderator material in next-generation nuclear reactors. While the overall nature of the production of nuclear graphite is well understood, the historic nuclear grades of graphite are no longer available. This paper reports the virgin microstructural characteristics of filler particles and macro-scale porosity in virgin nuclear graphite grades of interest to the Next Generation Nuclear Plant program. Optical microscopy was used to characterize filler particle size and shape as well as the arrangement of shrinkage cracks. Computer aided image analysis was applied to optical images to quantitatively determine the variation of pore structure, area, eccentricity, and orientation within and between grades. The overall porosity ranged between ∼14% and 21%. A few large pores constitute the majority of the overall porosity. The distribution of pore area in all grades was roughly logarithmic in nature. The average pore was best fit by an ellipse with aspect ratio of ∼2. An estimated 0.6–0.9% of observed porosity was attributed to shrinkage cracks in the filler particles. Finally, a preferred orientation of the porosity was observed in all grades. [Copyright &y& Elsevier]

Abstract: A novel computational approach for simulation of intergranular cracks in a polycrystalline aggregate is proposed in this paper. The computational model includes a topological model of the experimentally determined microstructure of a 400μm diameter stainless steel wire and automatic finite element discretization of the grains and grain boundaries. The microstructure was spatially characterized by X-ray diffraction contrast tomography and contains 362 grains and some 1600 grain boundaries. Available constitutive models currently include isotropic elasticity for the grain interior and cohesive behavior with damage for the grain boundaries. The experimentally determined lattice orientations are employed to distinguish between resistant low energy and susceptible high energy grain boundaries in the model. The feasibility and performance of the proposed computational approach is demonstrated by simulating the onset and propagation of intergranular cracking. The preliminary numerical results are outlined and discussed. [Copyright &y& Elsevier]

Abstract: The paper describes the influence of the microstructure (coming from the extrusion shape, the chemical composition and the thermo-mechanical treatments) of (9–18%)Cr–W–Ti–Y2O3 ODS steels on their impact fracture properties. The extrusion shape plays a major role on the impact properties, materials extruded as a rod present a higher upper shelf energy (USE) and a lower ductile to brittle transition temperature (DBTT) compared to materials extruded as plates. The DBTT for the non-recrystallized 14%Cr ferritic steels was shifted towards higher temperatures compared to the 9%Cr tempered ferritic–martensitic steel. Increasing the W and Ti content in 9%Cr tempered ferritic–martensitic ODS steel leads to a USE and a DBTT shifted towards higher energies and higher temperatures respectively. Increasing the yttria content leads to a drop of the impact energy and a shift of the DBTT of ferritic ODS steel towards higher temperatures. The present study highlights extensive splitting of the fracture surfaces and a dependency of the impact energy on the fracture plane orientation according to the microstructure anisotropy. [Copyright &y& Elsevier]

Abstract: The presented paper summarizes the results of general corrosion and stress corrosion cracking (SCC) susceptibility tests in supercritical water (SCW), studied for austenitic stainless steel 316L, with the aim to identify maximum SCW temperature usability and specific failure mechanisms prevailing during slow strain-rate tensile (SSRT) tests in ultra-pure demineralized SCW solution with controlled oxygen content. The general corrosion tests clearly revealed the applicability of austenitic stainless steel in SCW to be limited to 550°C as maximum temperature as oxidation rates of austenitic stainless steels 316L increase dramatically above 550°C. The SSRT tests were performed using a step-motor controlled loading device in an autoclave at 550°C SCW. Besides the strain rate (resp. crosshead speed), the oxygen content was varied in the series of tests. The obtained results showed that even at the lowest strain rate, a serious increase of SCC susceptibility, as typically characterized by IGSCC crack growth, was not observed. The fractography confirmed that failure was due to a combination of transgranular SCC and transgranular ductile fracture. Based on fractographic findings a phenomenological map describing the SCC regime of SSRT test parameters could be proposed for AISI 316L. [Copyright &y& Elsevier]

Abstract: Weibull stress model represents a basic local approach model used in the ductile-to-brittle transition region for description and prediction of cleavage fracture for materials of both PWR and WWER reactor pressure vessels. In the Weibull stress model used most frequently until now , the parameters are determined by a calibration procedure using the fracture toughness values of high and low constraint specimens. In the present paper, the results of SEN(B) pre-cracked specimens of 10×20×120mm size, with deep and shallow cracks, are utilized. Specimens were made of material of WWER-1000 reactor pressure vessel, and were tested at Nuclear Research Institute Rez. Determination of Weibull stress was performed for both the case of including plastic strain correction into the Weibull stress formula and without it. [ABSTRACT FROM AUTHOR]

Abstract: Local approach to brittle fracture for low-alloyed steels is discussed in this paper. A bibliographical introduction intends to highlight general trends and consensual points of the topic and evokes debatable aspects. French RPV steel 16MND5 (equ. ASTM A508 Cl.3), is then used as a model material to study the influence of temperature on brittle fracture. A micromechanical modelling of brittle fracture at the elementary volume scale already used in previous work is then recalled. It involves a multiscale modelling of microstructural plasticity which has been tuned on experimental inter-phase and inter-granular stresses heterogeneities measurements. Fracture probability of the elementary volume can then be computed using a randomly attributed defect size distribution based on realistic carbides repartition. This defect distribution is then deterministically correlated to stress heterogeneities simulated within the microstructure using a weakest-link hypothesis on the elementary volume, which results in a deterministic stress to fracture. Repeating the process allows to compute Weibull parameters on the elementary volume. This tool is then used to investigate the physical mechanisms that could explain the already experimentally observed temperature dependence of Beremin’s parameter for 16MND5 steel. It is showed that, assuming that the hypothesis made in this work about cleavage micro-mechanisms are correct, effective equivalent surface energy (i.e. surface energy plus plastically dissipated energy when blunting the crack tip) for propagating a crack has to be temperature dependent to explain Beremin’s parameters temperature evolution. [ABSTRACT FROM AUTHOR]

Abstract: The aim of this work is a second reply to Puls’s comments on the author’s first reply to the paper published in J. Nucl. Mater. 393 (2009) 350–367. The Dutton–Puls model indicates that the CGR is governed not by the stress gradient but by the ΔC that results from a decrease in the crack tip solubility due to the stress when compared to the bulk solubility, demonstrating that Puls’s defense of the Dutton–Puls model is inconsistent and invalid. Given the fact that DHC involves three consecutive processes such as nucleation, growth and cracking of hydrides at the crack tip, Puls’s claim that DHC is simply a diffusion-controlled process and the CGR is governed solely by the rate of hydride growth is incorrect, yielding many of the unsolved issues related to DHC. It is confirmed that Kim’s criticism that the Dutton–Puls model for the crack growth rate (CGR) is established based on a faulty thermodynamic basis is correct. [Copyright &y& Elsevier]

Abstract: The aim of this work is a reply to Puls’s paper entitled “Review of the thermodynamic basis for models of delayed hydride cracking (DHC) rate in zirconium alloys in J. Nucl. Mater. 393 (2009) 350–367” claiming that the thermodynamic basis of Dutton and Puls’s model termed the first version of the old models is valid when compared to Kim’s new model. The critical defect of the first version old model is to assume that the bulk hydrides are the source of diffusible hydrogen and the stress would decrease the crack tip concentration in solution even without hydride precipitation, creating the difference in hydrogen concentration or ΔC between the crack tip and the bulk. The latter assumption leads to predict DHC even at high temperature above 300°C without a thermal cycle, demonstrating that the first version old model is unrealistic. The second version old DHC model assumes that the stress gradient is driving hydrogen to the crack tip, increasing the crack tip concentration over the bulk concentration, which is faulty because it violates the thermodynamic principle that tensile stresses lower the chemical potential of hydrogen or the hydrogen solubility. Despite the old models’ claim that the driving force for DHC is the stress gradient, their analytical equations indicate that the crack growth rate (CGR) is governed by the ΔC, demonstrating that the driving force for the DHC is the ΔC, not the stress gradient. [Copyright &y& Elsevier]

Abstract: High-purity zone refined zirconium carbide has been fast neutron irradiated in the dose and temperature range of 1–10×1025 N/m2 (E >0.1MeV) and 635–1480°C, respectively. Non-irradiated and as-irradiated properties were measured including the lattice parameter, hardness and elastic modulus as determined through nano-indentation, thermal conductivity, and indentation fracture toughness. The effects of neutron irradiation on the microstructure were also determined though using transmission electron microscopy. The general finding of this paper, limited to this particular zone refined ZrC of nominal C/Zr ratio of 0.93, is that this ceramic is quite stable under neutron irradiation in the temperature and dose range studied. Measurement of lattice parameter before and after irradiation indicated a lack of significant crystalline strain due to irradiation. Only modest changes were observed in the mechanical properties of hardness, elastic modulus, and indentation fracture toughness. The thermal conductivity underwent a slight reduction near 1000°C irradiation, though was essentially unchanged for 1300–1480°C irradiation. Transmission electron microscopy revealed black-spot-type defects (likely Frank or other small loops) for irradiation at 670°C, maturing to faulted Frank loops at 1280°C. As the irradiation temperature increased from 1280°C to the highest irradiation temperature, of 1496°C, a transition to prismatic loops occurs. [Copyright &y& Elsevier]

Abstract: Modelling of advanced materials has become a main tool in materials development. This is particularly the case for present developments in advanced nuclear materials. This paper reviews the current status of modelling fuels and structural materials in the EU and discusses future programs. At the microscale, fundamental physics issues such as magnetism are being included to build a robust modelling scheme and to understand microstructural stability. At the meso-scale, dislocation interactions with defect clusters, void/bubbles and dispersoids are shown to be the main cause of hardening. How such issues, as well as model validation experiments are emerging to verify these modelling schemes is presented. [Copyright &y& Elsevier]

Abstract: The degree of embrittlement of the reactor pressure vessel (RPV) limits the lifetime of nuclear power plants. Therefore, neutron irradiation-induced embrittlement of RPV steels demands accurate monitoring. Current federal legislation requires a surveillance program in which specimens are placed inside the RPV for several years before their fracture toughness is determined by destructive Charpy impact testing. Measuring the changes in the thermoelectric properties of the material due to irradiation, is an alternative and non-destructive method for the diagnostics of material embrittlement. In this paper, the measurement of the Seebeck coefficient () of several Charpy specimens, made from two different grades of 22 NiMoCr 37 low-alloy steels, irradiated by neutrons with energies greater than 1MeV, and fluencies ranging from 0 up to 4.5×1019 neutronspercm2, are presented. Within this range, it was observed that increased by ≈500nV/°C and a linear dependency was noted between and the temperature shift ΔT 41J of the Charpy energy vs. temperature curve, which is a measure for the embrittlement. We conclude that the change of the Seebeck coefficient has the potential for non-destructive monitoring of the neutron embrittlement of RPV steels if very precise measurements of the Seebeck coefficient are possible. [Copyright &y& Elsevier]

Abstract: In this paper, the authors present the results of investigations into the mechanical characteristics after irradiation of GlidCopAl25/316L(N) and CuCrZr/316L(N)-type joints produced by two joining technologies. Specimens of the joints were irradiated in the RBT-6 reactor in the dose range of 10−3–10−1 dpa at Т irr =200°С and 300°С. Irradiation at Т irr =200°С causes strengthening of the joints specimens (by about 100MPa at the maximum dose). The uniform elongation drops from 8% in the initial state to 2–3%. But the total elongation remains at a relatively high level of ∼7%. Irradiation at 300°С does not practically change the ultimate strength of CuCrZr/316L(N) joints, but the total and uniform elongation of the joints decreases monotonously with a dose. But in this case, the level of elongation remains considerably higher than that when irradiated at 200°С. [Copyright &y& Elsevier]

Abstract: Advanced nanostructured ferritic alloys (NFAs) containing a high density of ultra-fine (2–5nm) nanoclusters (NCs) enriched in Y, Ti, and O are considered promising candidates for structural components in future nuclear systems. The superior tensile strengths of NFAs relative to conventional oxide dispersion strengthened ferritic alloys are attributed to the high number density of NCs, which may provide effective trapping centers for point defects and transmutation products produced during neutron irradiation. This paper summarizes preliminary tensile and fracture toughness data for an advanced NFA, designated 14YWT, currently being developed at Oak Ridge National Laboratory. For this study, an alloy designated 14WT was manufactured using the same production parameters used to produce 14YWT but without the Y2O3 addition during ball milling required for NC formation in order to quantify the effect of the NCs on mechanical properties. Tensile specimens produced from both alloys were irradiated at 300, 580, and 670°C to 1.5 displacements per atom (dpa), while 14YWT fracture toughness specimens were irradiated at 300°C to 1.5dpa. Tensile strengths for 14YWT were found to be about two times greater than 14WT for both irradiated and unirradiated conditions, with yield strength for 14YWT decreasing from ∼1450MPa at 26°C to ∼700MPa at 600°C. Moderate radiation-induced hardening (50–200MPa) and reduction in ductility was observed for 14YWT for all irradiation conditions and test temperatures. In contrast, 14WT exhibited significant hardening (∼250MPa) for the 300°C irradiated specimens, while almost no hardening was observed for the 580 and 670°C irradiated specimens. Fracture toughness results showed 14YWT in the unirradiated condition had a fracture toughness transition temperature (FTTT) around −150°C and upper-shelf K JIc values around 175MPa√m. Results from irradiated 14YWT fracture toughness tests were found to closely mirror the unirradiated data and no shift in FTTT or decrease in K JIc values were observed following neutron irradiation to 1.5dpa at 300°C. [Copyright &y& Elsevier]

Abstract: The presence of micro-cracks at the surface of a ferritic–martensitic steel is known to favour its embrittlement by liquid metals and thus decrease the mechanical properties of the structural materials. Unfortunately, conventional fracture mechanics methods cannot be applied to tests in liquid metal environment due to the opaque and conducting nature of the LBE. Therefore new methods based on the normalization technique for assessment of plain strain fracture toughness in LBE were examined. This paper discusses the assessment of the plain strain fracture toughness of T91 steel in liquid lead bismuth environment at 473K, tested at a displacement rate of 0.25mmmin−1 and makes the comparison with results obtained in air at the same temperature and displacement rate. Although there is a decrease of the fracture toughness by 20–30% when tested in LBE, the toughness of the T91 steel remains sufficient under the tested conditions. [Copyright &y& Elsevier]

The high number density of stacking fault tetrahedra (SFTs) observed in irradiated fcc metals suggests that they should contribute to radiation-induced hardening and, therefore, taken into account when estimating mechanical properties changes of irradiated materials. The central issue is describing the individual interaction between a moving dislocation and an SFT, which is characterized by a very fine size scale, ∼100 nm. This scale is amenable to both in situ TEM experiments and large-scale atomic modelling. In this paper we present results of an atomistic simulation of dislocation–SFT interactions using molecular dynamics (MD). The results are compared with observations from in situ deformation experiments. It is demonstrated that in some cases the simulations and experimental observations are quite similar, suggesting a reasonable interpretation of experimental observations. [Copyright &y& Elsevier]

EUROFER97 is the European candidate RAFM steel for use as a structural material in fusion energy systems. It is presently under investigation by several European laboratories, within the long term programme of EFDA (European Fusion Development Agreement). This paper presents the outcome of the mechanical characterization of this steel that has been carried out at SCK • CEN (Belgian Center for Nuclear Studies), in the unirradiated condition and after irradiation during three different campaigns in the BR2 reactor (300 °C, doses between 0.3 and 1.6 dpa). Tensile, Charpy impact, and fracture toughness specimens have been irradiated and tested, in order to obtain an experimental assessment of the main effects of neutron exposure on tensile and toughness properties; namely, irradiation hardening (increase of mechanical resistance and loss of ductility) and irradiation embrittlement (shift of ductile-to-brittle transition temperature and degradation of upper shelf energy). Comparisons with another well-known IEA reference RAFM steel, F82H, are provided. [Copyright &y& Elsevier]

Weldability of irradiated stainless steel (SS) has been studied to develop the technical guideline regarding the repair-welding of reactor internals. Type 304 and 316L SSs were irradiated at ambient temperature in the US Advanced Test Reactor. The multi-pass bead-on-plate TIG (GTA) and YAG laser welding with heat input levels less than 1 MJ/m were performed on specimens containing helium up to 18 appm. In this paper, results of cross-sectional micrograph observations of the heat affected zone were considered in light of helium bubble properties. The tendency for weld crack formation of irradiated Type 316L SS was compared with that of irradiated Type 304 SS. [Copyright &y& Elsevier]

Neutron irradiation in the core of a nuclear power plant can change the material properties of stainless steel significantly. In this paper cold worked AISI 316 baffle bolts retrieved from Doel 1 PWR were investigated with the aim to assess the structural degradation due to neutron irradiation by analyzing microstructure, mechanical behavior and stress corrosion susceptibility. The baffle bolts had received neutron doses ranging from 6 to 59 dpa covering a major part of a nuclear power plant's lifetime. First, microstructural investigations were carried out to reveal the defect and dislocation density and the amount of void swelling. Then mechanical properties were determined using tensile tests and fracture toughness tests. The irradiation assisted stress corrosion cracking (IASCC) susceptibility was investigated by slow strain rate testing and crack growth rate testing. Results were compared with literature values and found to be in good agreement. Main conclusions were that void swelling was small even for the highest dpa values and that the fracture toughness values saturated above 15 dpa. The material was susceptible to IASCC as shown by SSRT and crack growth rate measurements exhibiting intergranular fracture morphology. [ABSTRACT FROM AUTHOR]

F82H is the Japanese reference reduced-activation ferritic-martensitic (RAFM) steel for fusion blanket applications. The harsh environment of a fusion reactor, such as neutron irradiation and He/H damage, can result in significant degradation of F82H fracture toughness. Therefore, understanding the fracture toughness behavior of F82H in the fusion environment is critical to ensure the long-term safe operation of the fusion reactor. In this paper, we summarize seven irradiation campaigns of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) covering five variants of F82H steels, including F82H IEA, F82H Mod3, F82H doped with 1.4% natural Ni, F82H doped with 1.4% 58Ni, and F82H doped with 1.4% 60Ni. The irradiation temperatures covered the range from 220 °C to 530 °C and the neutron irradiation dose spanned 4 dpa to 70 dpa. The effects of neutron irradiation temperature, dose, materials composition, Ni doping, and He production on F82H fracture toughness are discussed. Our results showed that irradiation embrittlement monotonically decreased with increasing irradiation temperature until 400 °C for F82H IEA and F82H Mod3. F82H Mod3 showed better fracture toughness than F82H IEA both before and after neutron irradiation. We determined that 1.4% Ni alloying can be applied to F82H for simulating He effect in a fission reactor without jeopardizing the fracture toughness of the material. However, more studies are needed to understand the effect of high dose (>20 dpa) and He production on F82H fracture toughness. [ABSTRACT FROM AUTHOR]

Abstract: The present study is mainly aimed to provide the primary damage (displacements per atom, generation of solid transmutants and gas production rates) of structural materials irradiated in the high and medium flux test modules of the International Fusion Materials Irradiation Facility (IFMIF). We have investigated if the change of the composition during the irradiation time has effect on the prediction of the atomic displacements. The effect of the activation cross section uncertainties in the assessment of both solid transmutants and hydrogen and helium production is also analyzed. The results are provided element-by-element, so that the primary damage of any material irradiated in such neutron environments can be easily assessed; in this paper, we have predicted the primary damage of the low activation steel Eurofer. [Copyright &y& Elsevier]

This paper describes molecular dynamics simulations of fracture along a grain boundary (GB) in nickel (Ni) decorated with helium (He)-filled bubbles. We find that the quantity governing embrittlement of the GB is its areal coverage by He bubbles. Indeed, there is a threshold coverage—approximately 18%, for the GB studied here—above which the GB undergoes brittle fracture. During brittle fracture, intergranular cracks advance by the growth and coalescence of He-filled cavities ahead of the crack front. This process is enabled by plastic deformation of inter-bubble ligaments. In models with lower He bubble coverage, dislocation activity initiates in the adjoining grains prior to cavity growth and coalescence, relaxing crack tip stresses and averting brittle crack propagation. We conclude that the ductile-to-brittle transition observed in this study is governed by a competition between dislocation emission ahead of the crack tip and localized deformation confined to the inter-cavity ligaments. • Molecular dynamics simulations of fracture along a grain boundary in nickel decorated with He bubbles is performed. • When the fraction of boundary area covered by He bubbles exceeds approximately 18%, the GB undergoes brittle fracture. • During brittle fracture, intergranular cracks advance by the growth and coalescence of He bubbles ahead of the crack front. [ABSTRACT FROM AUTHOR]

Probably the most important range of materials for consideration as the blanket material for the tokamak design for fusion reactors ITER and DEMO is the high alloy Fe–9Cr oxide dispersion strengthened ferritic steels. Ferritic steels possess exceptional thermal conductivity and low thermal expansion and are resistant to void swelling. Their main drawback is high ductile to brittle transition temperatures, particularly in the oxide dispersion strengthened versions. This paper describes attempts to reduce the DBTT in an un-irradiated ferritic steel by a novel heat treatment procedure. New batches of high alloy Fe–9Cr oxide dispersion strengthened (Eurofer) ferritic steel have been produced by a powder metallurgy route, and relatively homogeneous material has been produced by hot isostatic pressing (HIP). Mini-Charpy test specimens were made from materials which had been subjected to a matrix of heat treatments with varying solution treatment temperature (ST), cooling rate from the ST temperature, and tempering treatment. The initial DBTT was in the range of 150–200°C (423–473K). Downward shifts of up to approximately 200°C (473K) have been observed after solution treatment at 1300°C (1573K) followed by slow cooling. This paper describes the microstructure of this material, and discussion is made of the likely microstructural factors needed to produce these DBTT downward shifts. [ABSTRACT FROM AUTHOR]

The aim of this paper is to assess the fracture behavior of tungsten plate material. Plate material with a thickness of 1mm and 4mm was investigated, considering samples with dimensions of 1×3×27mm3 without notch, 3×3×27mm3 without notch, and 3×4×27mm3 with a 1mm notch (KLST). The plate material was in either as-received or recrystallized condition (annealed for 1h at 2000°C (2273K)). All samples were characterized by means of Charpy impact tests, which were performed up to 1000°C (1273K) in a vacuum. The results show that plate material is anisotropic and that the brittle-to-ductile transition temperature (BDTT) strongly depends on the orientation of the specimen. A comparison of the results of notched and unnotched samples shows that delamination behavior only appears on notched samples and that unnotched samples show the classical brittle-to-ductile transition (BDT). Furthermore, it is shown that recrystallized tungsten plate material is extremely brittle and that the BDT does not even appear at a test temperature of 1000°C (1273K). [ABSTRACT FROM AUTHOR]