UO 2 -BeO is one of the most promising accident tolerant nuclear fuels due to its excellent thermal conductivity compared to pure UO 2 fuel. Two different UO 2 -BeO fabrication methods have demonstrated the capability to fabricate enhanced thermal conductivity UO 2 -BeO composite fuels and improve fuel performance. In one method, BeO is continuously distributed around UO 2 grains while BeO is dispersed in the UO 2 phase in the other method. In the former type, BeO is considered as matrix and UO 2 as dispersed particle. However, in the later type, BeO is considered as dispersed particle and UO 2 as matrix. To calculate the thermal conductivity of UO 2 -BeO composite, Hasselman-Johnson model has been applied, which shows good agreement with experimental data. In this model, it includes the influence of thermal conductivity of matrix and particle, volume fraction of particle, radius of particle and the interfacial thermal conductivity between matrix and particle. To balance the improvement of thermal conductivity and enrichment of UO 2 , a UO 2 -BeO composite fuel with 10% volume fraction of BeO has been chosen. Besides, the grain size of matrix should be noticeable smaller than particle to relax thermal stress which may cause micro-cracks and destruction, leading to the grain sizes of UO 2 in two types being distinctive, and resulting in very big effect on fission gas release. In the paper the thermal conductivity has been intensively studied as well as fuel performance in two different types of UO 2 -BeO fuels, and the two fabrication methods have also been compared to assess their applications in commercial reactors.

Up to now, two kinds of filler metal with or without nickel element for submerged arc welding have been largely used in the reactor pressure vessel (RPV) manufacturing. In order to study the effect of nickel element on weld metal properties of SA-508 Gr.3 Cl.1, submerged arc welding material with nickel (AWS classification F8P4-EGN-F2N, F2 for short) and welding material without nickel (F8P4-EA3N-A3N, A3 for short) were used; and conventional mechanical properties, low-cycle fatigue test, and proton irradiation analysis of the two weld metals were studied. Results show that the mechanical properties of the two different weld metals are similar, except that the Charpy V-notch impact property of the weld metal with nickel is better than that without nickel; the micro-structures of F2 and A3 weld metals are both composed of ferrite base and granular bainite, but the columnar grain size of F2 weld metal is smaller relatively, which results in better impact property. In addition, the irradiated A3 weld metal has fewer dislocation loops than the irradiated F2 weld metal after the same proton irradiation dose; the irradiated weld metals both have higher micro-Vickers hardness than before.

In order to gain high strength, fine grain size, stronger anti-corrosion property, and especially low permeability, the material 022Cr19Ni10N was chosen to manufacture the Rod Travel Housing Forging (RTHF) for Control Rod Drive Mechanism (CRDM). But, cracks were found in some forgings failing to meet the requirements of ultrasonic testing (UT). The causes of the forging cracks of this austenitic stainless steel forging were investigated by means of metallography, scanning electron microscopy (SEM) and other experimental methodology. The results indicated that the second δ-ferrite phase leads to the forging cracks between γ-δ interface during the low temperature forging process, and finally leads to the forging failure. It’s found that the cracks are distributing along the stripe δ-ferrite, and almost distributing in the same area as the large size δ-ferrite by metallography & SEM microstructure observation. The δ-ferrite is firstly found in the electroslag ingot, and in which, the distribution and size is different from the case to the core. The largest size δ-ferrite is around the core area, and this characteristic passes on to the final forging microstructure, although the shape, quantity & distribution of the δ-ferrite changed during the manufacturing process. Most forging cracks were found around the core area of the forging by UT examination. In the final forging process, when the forging temperature drops to 750∼850°C, the δ-ferrite have been forged to stripe shape and hundreds-micron size while the plasticity of the austenite reduce. What’s more, there are large hot plasticity differences between the δ-ferrite and the austenite, so the forging cracks initiate between γ-δ interface and extend to the area around to be a long crack in the low temperature forging process. In order to avoid the forging cracks in the Rod Travel Housing Forging, it’s necessary to reduce the content of δ-ferrite or improve the final forging temperature. Improving the final forging temperature, to guarantee the plasticity of the δ-ferrite and austenite, is another process to reduce the cracks. But while the temperature improves, the grain size grows rapidly, and may form mixed structure. So the most effective mean to reduce the content of δ-ferrite is to redesign the chemical components, mainly by increasing the nitrogen content from 0.06 (wt, %) to 0.12(wt, %), which makes the low temperature forging process for fine grain size possible. In the high-nitrogen-content forging, the δ-ferrite distributed sporadically and no δ-ferrite strip is found. By increasing the austenite forming elements (especially nitrogen), the cracks during low temperature forging process are avoided. What’s more, owning to the optimization of chemical compositions and manufacturing processes, the Rod Travel Housing Forging got fine grain size, low relative permeability, and good comprehensive mechanical properties with the ultimate tensile strength up to 570MPa.

The thermal conductivity is one of the most important properties for UO 2 . The influences of microstructure are especially important for UO 2 due to the severe structural changes under irradiation conditions. In this study, we have investigated the thermal conductivity of UO 2 with different microstructures using Finite Element Method. The thermal conductivity increases with increasing grain size. The grain size distribution has obvious influence on the thermal conductivity especially when there are pores in the polycrystal. The influences of porosity and pore size are very sensitive to the position of the pores. The results obtained in this study are useful for prediction of property changes of UO 2 fuel in pile and important to gain some design guidance to tune the properties through the control of the microstructure.

This paper reviews the role of fundamental factors such as stress, temperature and material microstructure in contributing to the PWSCC susceptibility in Alloy 600 components of operating PWRs. The effect of residual stresses from fabrication, cold work and welding as well as cyclic work hardening are considered. The influence of fabrication practice and the resulting microstructural parameters such as grain size, carbide distribution and ghost boundaries are considered. The paper considers the development of deterministic and probabilistic PWSCC susceptibility models to predict failures in the components and explain how these are applied to operating units. Investigation of service failures in the base metal and welds are considered.

During its lifetime in the core of a nuclear reactor, the fuel undergoes significant changes in its physical, chemical and morphological characteristics. In outer regions of the fuel pellets the so called “high burn-up” or “rim” structure may form. In this region UO 2 grains, with an original size of about 10 μm in fresh dense fuel, are reorganized into a porous structure with grain size 0.1–0.3 μm and porosity fraction up to 20 %. The mechanical, thermal and fission product retention properties of the high burn-up structure have encouraged further interest and attempts to mimic this morphology in fresh fuel. The JRC-ITU has studied various techniques for the synthesis of uranium and thorium dioxide in aqueous or nonaqueous media. Such nanoparticles can serve as starting material for production of material having similar characteristics as the high burn-up structure, as has been proved using Zr(Y)O 2 nanoparticles. Recently, efforts have been focused on the compaction of the nanoparticle powders. A spark plasma sintering device (SPS, FCT Systeme GmbH) has been commissioned in the JRC-ITU and tested using various nonradioactive materials. The present study is oriented on pressing and sintering of Hf(Y)O 2 and ZrO 2 into nanostructured pellets. Final products have been characterized by optical and electron microscopy, X-ray powder diffraction and density measurements.

Li 2 TiO 3 has been recognized as one of the most promising tritium breeding materials for D-T fusion reactor blanket. In this study, ultra-fine Li 2 TiO 3 powder was prepared rapidly by microwave-induced solution combustion synthesis (MSCS) using nitrates of lithium and titanate as raw materials, citric acid as fuel. The as-synthesized Li 2 TiO 3 powder exhibits an average crystalline size as small as 20 nm with uniform distribution. Computer-assisted 3D printing technology was employed to fabricate Li 2 TiO 3 ceramic breeding pebbles. This computer-assisted process is precise, efficient and controllable, which offers an alternative for the mass production of Li 2 TiO 3 pebbles. The pebbles exhibit good sphericity, relatively small grain size, preferable sinterability and crushing load strength.

In this paper, the low temperature sintering technology to UO 2 pellets has been introduced, and we have studied the high temperature creep properties of the pellets which manufactured by low temperature sintering. The sintering temperatures are 1073K, 1273K, 1473K and 1673K, sintering time are 1 hour, 2 hours and 3 hours respectively. We obtained the highest sintering density of pellets at 1673K with 3 hours, which is 10.41g/cm 3 (94.98% theoretical density). The grain size of pellets which prepared by low temperature sintering technology and traditional technology are 9.0μm and 23.8μm respectively. So the high temperature creep properties of the two kinds of pellet must be studied. They were performed at 20–50 MPa, 1673K and 1773K respectively, under a nitrogen atmosphere to shorten the experimental time. According to the results, the creep rates of sintered UO 2 with the grain sizes of 9.0μm and 23.8μm under the load of 10MPa are almost the same. The creep process is controlled by both Nabarro-Herring creep and Hamper-Dorn creep for uranium dioxide with grain size of 9.0μm; while Hamper-Dorn creep is the dominant mechanism for uranium dioxide with grain size of 23.8μm.

Hot deformation characteristics of forged and β-quenched Zr-1.0Sn-0.3Nb-0.3Fe-0.1Cr (N18 alloy) in the temperature range 625–950°C and in the strain rate range 0.005–5 s −1 have been studied by uniaxial compression testing of Gleeble 3500. For this study, the approach of processing maps has been adopted and their interpretation done using the Dynamic Materials Model (DMM). Based on a series of true stress-true strain curves on various temperatures and strain rates, the flow stress has been summarized and both the strain rate sensitivity index (m) and deformation activation energy (Q) have been calculated by the constitutive equations that flow stress and the relationship of Z parameter and flow stress have been established subsequently. Furthermore, the efficiency of power dissipation (⬜) given by [2m/(m+1)] and improved by Murty has been plotted as a function of temperature and strain rate to obtain different processing maps at different true strain rates ranging from 0.1–0.7. Subsequently, the microstructures of the specimens after compression testing were characterized by electron channeling contrast (ECC) imaging techniques used an FEI Nova 400 field emission gun scanning electron microscopy (FEG-SEM). The results showed that: (i) The hyperbolic sine constitutive equation can describe the flow stress behavior of zirconium alloy, and the deformation activation energy and flow stress equation were calculated under the different temperature stages which insists that the deformation mechanism is not dynamic recovery. (ii) The hot processing maps and its validation were analyzed, which indicated that the DMM theory was reliable and could be adopted as useful tool for optimizing hot workability of Zr. The optimum parameters for extrusion and hammer forging were revealed on the processing maps of 830–950°C, 0.048–2.141 s −1 and 916–950°C, 2.465–5 s −1 . (iii) The microstructure of the ingot exhibits a typical lamellar Widmanstatten structure. Under the different strain rates, the grains formed by dynamic recrystallization existed normally in the central zone of the compression samples while the no uniformity of grain size increased with the increasing of strain rate. Meanwhile, due to the dynamic recrystallization as a thermal activation process, the grains size and uniformity increased with the increasing of temperature. In brief, microstructure analysis showed that continuous dynamic recrystallization and geometric dynamic recrystallization operated concurrently during the isothermal compressive deformation.

The very high temperature gas-cooled reactor (VHTR), with dual capacities of highly efficient electricity generation and thermochemical production of hydrogen, is considered as one of the most promising Gen-IV nuclear systems. The primary candidate materials for construction of the intermediate heat exchanger (IHX) for the VHTR are Alloy 617 and Alloy 230. To have a better understanding of the degradation process during high temperature long-term service and provide practical data for engineering design of the IHX, aging experiments were performed on Alloy 617 and Alloy 230 at 900 and 1000°C. Mechanical properties (hardness and tensile strength) and microstructure were analyzed on post-aging samples after different aging periods (up to 3000 h). Both alloys attained increased hardness during the early stages of aging, and dramatically soften after extended aging times. Microstructural analysis including Transmission Electron Microscopy (TEM), Scanning Electron Microscope (SEM), Energy Dispersive X-ray spectroscopy (EDS), and Electron Backscatter Diffraction (EBSD) were carried out to investigate the microstmcture evolution during aging. A carbide particle precipitation, growth and maturing process was observed for both alloys, which corresponds to the changes of materials’ mechanical properties. Few changes in grain boundary character distribution (GBCD) and grain size distribution were observed after aging. In addition, high temperature corrosion studies were performed at 900 and 1000°C for both alloys. Alloy 230 exhibits much better corrosion resistance at elevated temperature compared to Alloy 617.

The microstructure and texture evolution of N18 sheets are studied after a series of fabrication processes which contain hot rolling at 780°C, cold rolling with the total deformation of 10% and recrystallized annealing at 580°C for 1.5h. The results show that grain size and the density of low angle grain boundaries of the fabricated sheets increase with the increase of hot rolling pass, and there is a little difference of grains for cold rolling and recrystallized sheets that deformed grain could be observed on both of them, which indicates that dynamic recrystallization occurs during hot rolling, but the cold deformation was so limited that the sheets could not be fully recrystallized because of lack of recrystallization driving force. Second phase precipitated particles (SPPs) in as-received N18 sheets are smaller than that in the fabricated sheets. It may because that part of SPPs are dissolved during hot rolling which lead to the volume decrease and supersaturation elements separate out to form SPPs again or primary SPPs grow up during recrystallization which lead to volume extending. The N18 sheets mainly have (0001) texture before and after fabrication and the density of basal pole which was along normal orientation on rolling plane increases as rolling pass increasing but decreases after cold rolling and increase again after recrystallization. After fabrication, the value of f ND increases for N18 sheets, but there were a little difference among different rolling pass. EBSD analysis shows that the grains of rolling sheets mainly display orientations with <10 1¯ 0>//RD and {0001} <10 1¯ 0> texture was stronger, and the grains of recrystallized sheets mainly display orientations with <11 2¯ 0>//RD made {0001} <11 2¯ 0> texture increasing.

Nuclear scientists and engineers should consider adopting a more operational approach for the purpose of selecting their future materials. For each type of nuclear power generating reactor, for each coolant (water, helium or liquid metal), the next generation of specialists and decision-makers will need to choose and optimise the iron or nickel alloys, steels, ODS (oxide dispersed strengthened steels) and ceramics that are going to be used. It may well be considered that either each reactor type has its own, specific materials, or, in a complementary manner, that the efforts for improvements should be shared. At high temperatures, as found on fuel-cladding liners, heat exchangers or even tubes or tube liners, different types of steels and alloys may be envisaged. It is considered that austenitic steels provide a better creep resistance at high temperature but they must be stabilized by nickel, thereby becoming more expensive. Ferrite steels could be better as far as swelling, mechanical strength and thermal behaviour are concerned. To withstand corrosion, chromium or aluminium, ODS steels could turn out to be good solutions, if they can comply with stringent criteria. Concerning heat exchangers, choices must be made between iron and nickel alloys, according to proposed operating conditions. In the case of sodium-cooled rapid neutron reactors (RNRs), ferritic-martensitic alloys with 9%–12% chromium or chromium ODS steels could prove suitable, especially if we judge by their specific mechanical behaviour, up to at least 700°C. Nevertheless, behaviour of these steels — with respect to ageing, anisotropy, radiation induced segregation, radiation induced precipitation, reduction of activation products and welding — needs be better understood and qualified. Sodium heat exchanger materials should be carefully chosen since they have to withstand corrosion arising from the primary flow and also from the secondary or tertiary flow (either sodium or molten salts, gas or water); therefore, experimental loops are necessary to gain improved understanding and assessment of the designs envisioned. One way to improve alloys is through thermal, mechanical treatments or by surface treatments. A better way could, however, be to improve the nanostructure and mesostructure of the materials chosen at the drawing-board stage, for instance by nano-size cluster dispersion and grain size controls; experimental tests, microscope and spectroscope observations, multi-scale modelling and thermodynamics computing could also help calibrate and implement these improvements. Large, experimental databases and codes will be the keystone to defining more operational knowledge bases that will then allow us to determine terms of reference for the new materials. Failing this, time will be running out — within the next twenty years — to design and develop nuclear prototypes consistent with the criteria laid down for “Generation IV” reactors.

Ferritic steels commonly used for pressure vessels and reactor supports in light water reactors (LWRs) exhibit dynamic strain aging (DSA) resulting in decreased ductility and toughness. In addition, recent work indicated decreased toughness during reverse-cyclic loading that has implications on reliability of these structures under seismic loading conditions. We summarize some of our recent work on these aspects along with synergistic effects, of interstitial impurity atoms (IIAs) and radiation induced point defects, that result in interesting beneficial effects of radiation exposure at appropriate temperature and strain-rate conditions. Radiation-defect interactions were investigated on pure iron, Si-killed mild steel, A533B, A516, A588 and other reactor support and vessel steels. In all cases, DSA is seen to result in decreased ductility accompanied by increased work-hardening parameter. In addition to mechanical property tests, fracture toughness is investigated on both A533B and A516 steels. While dips in fracture toughness are observed in A533B steel in the DSA region, A516 steel exhibited at best a plateau. The reasons could lie in the applied strain-rates; while J1c tests were performed on A533B steel using 3-point bend tests on Charpy type specimens, CT specimens were used for A516 steel. However, tensile and 3-point bend tests on similar grade A516 steel of different vintage did exhibit distinct drop in the energy to fracture. Load-displacement curves during J1c tests on CT specimens did show load drops in the DSA regime. The effect of load ratio (R) on J versus load-line displacement curves for A516 steel is investigated from +1 to −1 at a fixed normalized incremental plastic displacement of 0.1 (R = 1 corresponds to monotonic loading). We note that J-values are significantly reduced with decreasing load ratio. The work-hardening characteristics on the fracture surfaces were studied following monotonic and cyclic loading fracture tests along with the stress-field analyses. From the hardness and the ball-indentation tests, it was shown that decreased load ratio (R) leads to more strain hardening at the crack tip resulting in decreased fracture toughness. From the stress field analysis near the crack tip of a compact tension fracture toughness test specimen, a cycle of tensile and compressive loads is seen to result in tensile residual stresses (which did not exist at the crack tip before). These results are important to evaluations of flawed-structures under seismic loading conditions, i.e. Leak-Before-Break (LBB) and in-service flaw evaluation criteria where seismic loading is addressed. In addition, studies on fast vs total (thermal+fast) neutron spectra revealed unexpected results due to the influence of radiation exposure on source hardening component of the yield stress; grain-size of pure iron plays a significant role in these effects.

Computational algorithms aiming at modeling and visualization of the initiation and growth of intergranular stress corrosion cracks (e.g., in the steam generator tubes) on the grain-size scale have already been proposed. Main focus of the paper addresses the influence of randomly oriented anisotropic elasto-plastic grains on the microscopic stress fields at crack tips. The limited number of calculations indicate that the incompatibility strains, which develop along the boundaries of randomly oriented grains, influence the local stress fields (J-integrals) at crack tips significantly.

This paper deals with the correlation length estimated from a mesoscopic model of a polycrystalline material. The correlation length can be used in some macroscopic material models as a material parameter that describes the internal length. It can be estimated directly from the strain and stress fields calculated from a finite-element model, which explicitly accounts for the selected mesoscopic features such as the random orientation, shape and size of the grains. The crystal plasticity material model was applied during the finite-element analysis. Different correlation lengths were obtained depending on whether the strain or the stress field was used. The correlation lengths also changed with the macroscopic load. While the load is below the yield strength the correlation lengths are constant, and of the order of the average grain size. Increasing the load above the yield strength creates shear bands that temporarily increase the values of the correlation lengths calculated from the strain fields. With a further load increase the correlation lengths decrease slightly below the average grain size. The correlation lengths calculated from the stress field are smaller than the ones calculated from the strain field. However, with the exception of the load region where significant shear bands appear, both seem to follow the same qualitative rules.

The complete understanding of the incubation and growth of microstructurally short cracks is still somewhat beyond the present state-of-the-art explanations. A good example is the intergranular stress corrosion cracking of Inconel 600 in high-temperature water. An effort was therefore made by the authors to construct a computational model of the crack growth kinetics at the grain-size scale. The main idea is to divide continuum (e.g., polycrystalline aggregate) into a set of sub-continua (grains). Random grain structure is modelled using Voronoi-Dirichlet tessellation. Each grain is assumed to be a monocrystal with random orientation of the crystal lattice. Elastic behaviour of grains is assumed to be anisotropic. Crystal plasticity is used to describe (small to moderate) plastic deformation of monocrystal grains. Explicit geometrical modelling of grain boundaries and triple points allows for the development of the incompatible strains along the grain boundaries and at triple points. Finite element method (ABAQUS) is used to obtain numerical solutions of strain and stress fields. The analysis is currently limited to two-dimensional models. Numerical examples illustrate analysis of about one grain boundary long transgranular cracks. In particular, the dependence of crack tip displacements on the random orientation of neighbouring grains is studied. The limited number of calculations performed indicates that the incompatibility strains, which develop along the boundaries of randomly oriented grains, significantly influence the local stress fields and therefore also the crack tip displacements. First attempts are also made to quantify the preferential growth directions of cracks crossing the discontinuities (e.g., grain boundary).