One of the most crucial challenges facing the world today is the growing demand for energy. Several nations are aiming to minimize their dependence on fossil fuels due to geopolitical instabilities and the inherent issues of pollution and decreasing availability, as well as rising costs and the risks linked to more sophisticated extraction technologies.
Nowadays, a range of energy sources such as nuclear energy, photovoltaics, biofuels, hydroelectric, and wind are being considered. In 2008, Baldev Raj, M. Vijayalakshmi, P.R. Vasudeva Rao and K. B. S. Rao from the Indira Gandhi Centre for Atomic Research in India published an outstanding article on the subject titled “Challenges in Materials Research for Sustainable Nuclear Energy.” They explained emerging technologies in the nuclear industry and the material challenges that have to be overcome to make these technologies a feasible reality for energy production.
The authors delineate five long-term goals for making nuclear energy a more appealing energy source:
- Improving thermal efficiency by transferring to high-temperature reactors
- Multiple industrial applications of high-temperature reactors (i.e. using the heat for commercial purposes, for example, desalinating water or heating buildings)
- Better fuels (for example, thorium) and coolants specifically in fast spectrum reactors
- Enhanced safety and minimum nuclear waste via in-situ incineration using accelerator-driven systems
- Commercially feasible fusion technology
The authors observed that increased global cooperation to realize these goals and expect commercial fusion energy production in the second half of the 21st century. However, an important move in fulfilling these goals lies in overcoming material challenges created by these sophisticated technologies.
High-temperature mechanical properties, radiation resistance, fabricability, coolant compatibility, and fuel compatibility are the common areas of concern. Usually, the core component materials experience more extreme conditions when compared to the out-of-core materials.
This article examines the materials challenges mentioned in the Raj et al. paper and underlines some of the regions where automated EBSD or Orientation Imaging Microscopy (OIM™) can play a key role in resolving the material issues associated with the next generation of nuclear power. It also offers a short review of some of the studies on nuclear materials that have been reported in the scientific literature.
The aim of much nuclear materials research for existing reactor technology is concentrated on obtaining higher “burn up” (energy production per unit quantity of fuel). Table 1 based on the Raj et al. paper illustrates some of the main components, the current materials employed, and the chief challenges involved.
Table 1. Materials and material issues for thermal and fast spectrum nuclear reactors for a few critical components
||pressure vessels, piping
||cost and corrosion
||turbines and steam generators
||low alloy steels, 12% Cr steels
||cost, corrosion, high temperature, and pressure
||clad and hexcon
||cold worked 316 stainless steel, stainless steel alloy D9 (15% Cr, 15% Ni-Ti stabilized), ferritic steels
||void swelling, irradiation creep, irradiation embrittlement, tensile strength, ductility and creep strength, sodium compatibility, fuel & fission products
||316 stainless steels or 316 L(N) stainless steels
||tensile strength, creep low cycle fatigue, creep-fatigue interactions, weldability, fabricability
||modified 9Cr-10Mo ferritic steels
||sodium compatibility, steam corrosion resistance, fretting and wear
While choosing materials for a thermal reactor, the key consideration is selecting raw materials with low absorption for neutrons. Zirconium fulfills this demand. Considerable research has been conducted to produce a broad range of zirconium-based alloys with chemical properties optimized for enhanced corrosion and irradiation behavior and with a negligible tendency toward hydrogen embrittlement due to thermomechanical processing.
New materials are required to overcome challenges associated with hydriding, irradiation growth, and embrittlement, so as to achieve higher burnup.
Fast reactors (or fast neutron reactors) sustain the fission chain reaction using fast neutrons. Fast reactors can be developed to generate more fissile material than they use up (breeders) or incinerate the fuel to yield energy and/or transform long-half-life waste into less troublesome isotopes (burners). Irradiation creep and void swelling are two key materials challenges in fast reactors.
The dimensional variations and density reduction induced by these factors bring in several engineering issues. In order to obtain higher burnup, these materials are being pushed to their boundaries. For instance, ferritic steels tend to turn brittle upon exposure to radiation. A study is ongoing to overcome this issue through grain boundary engineering (GBE).
Some short-term enhancements in thermal efficiency for fast reactors are accompanied by a rise in the outlet temperature of the coolant from 823 K to around 1123 K. Due to this, it is necessary to develop radiation-resistant materials with good high-temperature creep resistance.
Ferritic steel with dispersed nanoparticles of titania and yttria (ODS steel) is one potential cladding material. However, issues related to this class of steels include deformation texture, fabricability, and anisotropy in mechanical properties.
In order to obtain higher burnup, sophisticated radiation-resistant materials are being developed including quaternary and ternary alloys of Zr-Nb. The following statement in Raj et al. elucidates the difficulties of introducing new materials into the nuclear sector: “A complex interplay of chemistry, microstructure, out-of-core behavior, and in-reactor performance necessitates detailed evaluation and validation of many zirconium-based alloys before acceptance in the nuclear industry.” This indeed applies to any new material, because a candidate material must be validated to ensure that it satisfies the challenging operating conditions present in nuclear reactors.
A study on oxide dispersion strengthened (ODS) steel by Chen, Sridharan, Ukai, and Allen (2007) is one good example. The investigators used various characterization tools such as OIM™ and Energy Dispersive Spectroscopy (EDS) to analyze the corrosion behavior of these materials upon exposure to supercritical water, specifically the role of grain boundaries. The exceptional creep properties of ODS steels make them candidate materials for cladding applications in fast reactors.
By combining OIM™ with other analytical techniques such as EDS, a complete characterization of the microstructure is provided — crystallographic, as well as chemical aspects of the material can be collected. Such information is important in order to understand the “complex interplay” between the crystallographic and chemical aspects of the microstructure. As an example, Figure 1 demonstrates simultaneous OIM™/EDS results collected from an interface between alumina and copper oxide.
Figure 1. Blended elemental map (left), phase map (center), and orientation map (right) overlaid on a gray scale map based on the quality of individual EBSD patterns at each point in the scan. The maps are constructed from simultaneously collected EBSD and EDS data.
Improved Material Models
In order to overcome the material issues of the nuclear industry, the precise modeling of material behavior is important. It is usually complicated, if not impossible, to reproduce the challenging operating conditions present in the latest nuclear reactors in the laboratory setting.
Predicting the behavior of material requires accurate material models which in turn require a basic understanding of the correlation between microstructure and properties. With the introduction of OIM™ as a practical tool for characterizing the crystallographic orientation, it is now possible to improve the various materials models connecting microstructure with properties to include orientation information for more exact simulation of microstructural evolution and more accurate predictions of material properties under changing conditions.
The literature has provided a number of good examples. For instance, Wu, Pan, and Stubbins (2007) have demonstrated how the effects of irradiation-induced hardening on 316L stainless steel can be modeled using the information obtained from OIM™ data. They employed OIM™ to characterize the local misorientation (an indicator of plastic strain), crystallography, slip system activity, and twinning and this information were, in turn, used to enhance the finite element modeling (FEM) results. Figure 2 represents some sample EBSD and FEM results from this study.
Figure 2. OIM™ orientation map (left) and FEM simulated stress contours (right) for a 316L stainless steel notch sample. (Adapted from Wu, Pan, and Stubbins, 2007)
Another example of modeling is reported in safety studies by Medevielle, Hugon, and Dugne (1999) on the system U, Zr, and O (called the corium system). Corium is important as it includes the vital components involved in the fusion of the tube (Zircaloy) with nuclear fuel (UO2) during a nuclear incident. EBSD was used together with Wavelength Dispersive Spectroscopy (WDS) to explicitly identify the complexity of phases present in samples achieved by fusion of the corium elements at differing oxygen concentrations. The chemical (via WDS) and crystallographic (via EBSD) characterization of the resulting oxides allows thermodynamic calculations to predict the behavior of corium at high temperatures.
Grain Boundary Engineering (GBE)
Grain boundary engineering uses thermomechanical processing to alter the microstructure to support specific types of grain boundaries over others in order to improve a grain boundary sensitive material property. For example, some kinds of grain boundaries may be corrosion-resistant, while others may be more vulnerable. Increasing the fraction of the corrosion-resistant boundaries at the cost of the corrosion susceptible boundaries is a means of fabricating the material to enhance its performance.
For instance, Alexandreanu, Capell, and Was (2001) have demonstrated that GBE can enhance the corrosion resistance of alloy 600 steam generator tube material. Furthermore, they have demonstrated that the enhanced performance can be obtained by encouraging the presence of coincident site lattice (CSL) boundaries. CSL boundaries exist where atoms are shared between the crystallite lattices contiguous to the grain boundary and are assumed to be less resistant to corrosion when compared to random boundaries.
These beneficial types of boundaries are incorporated into the materials via a range of thermomechanical treatments. As shown in Figure 3, increasing the CSL boundary fractions via these treatments results in a reduction in the fraction of boundaries cracked. Tan, Sridharan et al. (2006, 2008) have shown that analogous improvements in corrosion resistance can be obtained via GBE in Incoloy 800H and 617 (candidate alloys for Generation IV nuclear power systems).
Figure 3. Correlation of total cracked fraction of boundaries at 15% strain with the fraction of CSL boundaries as measured by OIM™. (Adapted from data in Alexandreanu, Capell and Was, 2001)
OIM™ is an excellent tool for GBE research due to its potential to characterize the crystallographic structure of grain boundaries quickly enough to, “facilitate probabilistic analyses where inferences on future component performance and reliability are possible” (Tan, Sridharan, and Allen, 2008).
OIM™ determines the crystal orientation at individual points in the microstructure and therefore it is a perfect tool for studying anisotropy and texture. The texture is defined as the statistical distribution of the lattice orientations of the constituent grains in a polycrystalline material. A majority of materials in their single crystalline form show some form of anisotropy (or directionality) properties with regard to the orientation of the crystal. For instance, a hexagonal crystal may have a different value in yield strength in a direction parallel to the basal pole direction or c-axis of the crystal corresponding to a direction normal to the c-axis.
The Young’s Modulus of zinc in the c-axis direction (28.7x10-12m2/N) is about three times greater than in a direction perpendicular to c (8.4x10-12m2/N) (Nye, 1985). Usually, it is believed that such directional changes in the constituent crystals in a polycrystal are averaged out over the bulk of the polycrystal. However, if the material has texture then usually it will also have anisotropy.
The effect that texture has on delayed hydride cracking (DHC) from deuterium uptake in Zr-Nb pressure tubes (Lehockey et al., 2007) is one nuclear application example of these concepts. This research found that higher deuterium (D) uptake takes place in microstructures with a wide distribution of basal poles (c-axes) in directions perpendicular to the longitudinal directions of the pressure tube as represented in Figure 4. According to the authors of this study, “the correlation between D uptake and texture measured by OIM™ may offer one means for optimizing processing/heat treatments during manufacturing to minimize DHC susceptibility. Alternatively, it may allow a methodology for screening samples to identify installed pressure tubes at risk of DHC.”
Figure 4. (0001) pole figures showing the texture of Zr-Nb tubes with (a) low deuterium uptake and (b) high deuterium uptake. (Adapted from Lehockey, Brennenstuhl, Pagan, Clark and Perovic, 2007)
Other instances of OIM™ based texture studies on zirconium alloys have been described by Une and Ishimoto (2006) on Zircaloy claddings with various thermomechanical processing, where the focus was on the link between texture and hydride precipitation and the study by Holt and Zhao (2004) on the evolution of texture in extruded Zr-2.5Nb tubes.
The key to the commercial development of nuclear power is the nuclear fuel cycle. A closed fuel cycle is desired due to the costs associated with the nuclear waste stream. Research is in progress to close the cycle through reprocessing and reuse of spent fuels in reactors. Nevertheless, until that technology is realized, safe storage of spent fuel in repositories is an important component of the fuel cycle. Several research studies are in progress in this field. Two examples are the study on the storage containers (Lehockey et al., 2007) and on the dissolution of the fuel itself after supposed contact with groundwater (Römer et al., 2003).
OIM™ has been employed in research to enhance the integrity of welds of copper waste containers being developed for long-term storage of used-up nuclear fuel. OIM™ was used for characterizing the local distribution of plastic strain within the weld, and also the affected zone of material surrounding the weld.
OIM™ has the ability to measure changes in orientation with resolutions of tenths of a degree at the sub-micron scale. These local changes in orientation provide a relative indication of strain in the material. Areas with high local orientation variations indicate highly strained regions in relation to unstrained regions where the orientation changes are negligible. Such information can be used, for instance, to observe the impact of local heating on the microstructure, such as that created during welding. OIM™ was used by Lehockey et al., (2007) to examine weld cracking in copper waste containers being developed for long-term storage of nuclear fuel - some sample results are demonstrated in Figure 5. The authors of this research stated that “OIM™ proved useful in recommending the development of a more homogeneous heat treatment” to avoid weld cracking. Welding of dissimilar metals also creates problems that can be tackled using OIM™ (Nelson, Lippold, and Mills, 2000).
Figure 5. Variation in strain magnitude (as indicated by varying levels of local misorientation) at various locations in a copper waste container weld. (Adapted from Lehockey, Brennenstuhl, Pagan, Clark, and Perovic, 2007).
Using EBSD combined with electrochemical atomic force microscopy (AFM), Römer, Plaschke, Beuchle, and Kim (2003) discovered that dissolution rates of uranium oxide are strongly anisotropic. Figure 6 illustrates that the dissolution rate of the grain with a (001) face almost in parallel to the surface (in red) is more rapid than those with near (111) orientations. Such understanding is important for guaranteeing accurate modeling of dissolution as required for a valid safety assessment of a nuclear waste repository.
Figure 6. SEM micrograph of corroded uranium-oxide showing grains of different dissolution levels and the corresponding EBSD patterns and orientation schematics. (Adapted from Römer, Plaschke, Beuchle, and Kim, 2003).
Metallic uranium alloys are candidate materials that can be used as fuel. Uranium hydrides can be used as preparatory materials to produce reactive uranium powder. Uranium hydride can be produced by heating uranium metal in the temperature range of 250 to 300 °C in the presence of hydrogen. OIM™ has been used to determine the effect of texture on hydride initiation and growth (Bingert, Hanrahan, Field, and Dickerson, 2004). Some results from this work are shown in Figure 7.
Figure 7. OIM™ orientation map showing deformation twins in hydride α-Uranium and bulk textures shown in gray intensities in inverse pole figure form overlaid with the location of key orientation components within the orientation map. (Adapted from Bingert, Hanrahan, Field and Dickerson, 2003).
All of the following longer-term developments need continuous research into structural materials for the reactor that can function under the strain, radiation, and thermal conditions that are present in these reactors. Conversely, other materials challenge specific to each of these reactor developments must be addressed as well. OIM™ will most surely play a role in finding ways to solve these problems.
High Temperature, Multiuse Reactors
High-temperature reactor technology requires a selection of new materials, for example, composites, advanced coatings, and refractory alloys. “Materials behaviors such as microstructural stability; mechanical properties such as creep, fatigue, and toughness; and chemical properties, such as corrosion and compatibility need to be understood in the new domains of higher temperatures and irradiation levels.” (Raj et al., 2008)
Models that associate properties with microstructure have been created for most of these materials at less harsh operating conditions. However, the validity of these models at the more extreme service conditions has to be confirmed, which needs properties, as well as microstructural data at data points within these more severe irradiation and temperature regimes. The inclusion of OIM™ data offers a more comprehensive explanation of the microstructure, which can help in enhancing these models.
Long-term goals in this development field include alternate fuels such as nitrides, carbides, or metal alloys. Although the cladding and structural materials for systems with these fuels are comparable to those already in use, reprocessing the waste products of these fuels may need advanced materials because the reprocessing creates highly hostile environments. (Raj, Ramachandran, and Vijalakshmi, 2009). Raj et al. indicate the use of some of the technologies previously mentioned to fulfill the requirements of fast-reactors. This involves grain boundary engineering to increase inter-granular corrosion resistance, the addition of minor elements to enhance void swelling behavior, and enhancement of models for improved predictive potentials.
The uranium-molybdenum system is one example of research in metal alloy fuels. Medevielle, Hugon, and Dugne (1999) were able to characterize the solidification structure and eventually observed dendrites, with the help of EBSD measurements. These dendrites denoted a liquid phase reaction during solidification as illustrated in Figure 8. At times, the microstructure can be identified before the phase transformation from the orientation maps of the post-transformation microstructure (Cayron, 2009).
Thorium Fuel Cycle
For thorium-based fuels, a comparatively high concentration of fluoride had to be used in reprocessing. This induces a need for widespread research in materials resistant to corrosion in environments rich in fluoride. In concert with the development of sophisticated materials to fulfill this requirement, grain boundary engineering is an established tool for increasing corrosion resistance in certain systems and may also play a role in thorium-based fuel development. The ability to measure chemical composition and its spatial distribution in materials makes EDS a perfect tool for conducting research into a sophisticated material’s performance in a fluoride-rich environment and due to the ability of EBSD to characterize grain boundaries, it is a vital component of any grain boundary engineering work.
Figure 8. SEM micrograph of a molybdenum solidification structure (top). Corresponding (001) pole figure (bottom). (Adapted from Medevielle, Hugon, and Dugne, 1999).
Accelerator Driven Systems
The material for the window that isolates the reactor from the accelerator is the greatest material challenge for these systems. The window material must be resistant to corrosion, irradiation, and embrittlement of both liquid metal and helium, and should also have excellent thermo-physical properties. OIM™ can deliver information to enhance performance through microstructure optimization.
More developments are required on structural materials with regards to all nuclear technologies. However, plasma-facing materials are the main area of materials research for fusion. These materials are exposed to high neutron radiation and also subjected to strong mechanical, thermal, and electromagnetic loading.
Vanadium alloys, advanced ferritic steels, tungsten-based refractory alloys, and silicon-carbide composite ceramics are being developed to satisfy these stringent demands. In addition, materials research has focused on coating technology to eliminate hydrogen embrittlement, which is a specific issue with the steels and vanadium alloys. Previously, materials behavior has been proven to be highly sensitive to exposure conditions. Therefore, extrapolation of the response of a material in one system to another system is usually inaccurate.
This eliminates the need for complete characterization of material microstructure for more precise predictive abilities. OIM™ and EDS play a vital role in understanding the effects of the chemical and crystallographic aspects of material on material behavior. For instance, OIM™ has been used to analyze the fatigue and creep-fatigue behavior on P91 martensitic steel (Fournier, Sauzay, Renaut, Barcelo, Pineau, 2009). The 9%–12% Cr martensitic steels have been chosen as candidate materials for structural components in future fusion reactors.
Figure 9. Orientation map (left) of the as-received microstructure and the average mean misorientation per block of laths after various cyclic loadings at 823K (right), where Δξf is the prescribed fatigue strain range. (Adapted from Fournier, Sauzay, Alexandra, Barcelo, and Pineau, 2009).
“Nuclear materials researchers and technologists have gained rich experience studying the behavior of present-generation materials, such as zirconium-based alloys and special steels, in the past three decades. The present trends suggest that these materials will undergo incremental changes in the immediate future, to increase the burnup of the fuel and the lifetime of the reactors. However, sustained R&D on a wide spectrum of materials such as refractory alloys, composites, ceramics, low-activation steels, and coatings and the related processing technologies is essential to meet the demands of emerging nuclear technologies” (Raj et al., 2008). OIM™ can have an important role to play in these R&D requirements—four common areas are:
- In areas where grain boundaries have a role to play, OIM™ should be regarded as a main component of the characterization toolbox. Corrosion damage and embrittlement usually occur along grain boundaries. Grain boundary engineering can play an important role in enhancing the performance of a material in extreme environments mainly related to in-core applications. As a recognized tool for characterizing grain boundary distributions in polycrystalline materials, OIM™ has helped make grain boundary engineering a real opportunity for customizing the microstructure of the material to optimize its performance predominantly in reducing cracking, corrosion, or inter-granular attack at surfaces subjected to aggressive environments innate to nuclear applications.
- Nearly all fabricating processes impart texture to some extent into materials. Insight into the texture evolution during well-characterized and well-controlled processing is a major input into the precise development of materials models. Accurate simulation of the response of a material is key to improving a development process in order to produce a part with a microstructure particularly customized to meet the in-service requirements of the part. Due to the ability of OIM™ to offer statistically pertinent and spatially specific orientation data, it is a perfect tool for supplying the scientists with key orientation data required by sophisticated finite element-based modeling.
- Besides texture, components developed through deformation processes like extrusion, rolling, or forging also include a degree of residual strain. At times, this is alleviated through heat treatment. However, at other times the material is left in a strained condition. Furthermore, the material in a component may also be strained in the operating condition. Usually, the strain is not evenly distributed across the component but differs locally at the micron scale. OIM™ offers a way to characterize these local variations in strain with its potential to measure local small variations in orientation. Information like that can be very useful for optimizing thermomechanical forming paths, and also for understanding induced stress-states at the time of service.
- With integrated systems OIM™ and EDS can be used together to provide insight into the interplay of crystallographic structure and chemical composition as new materials or systems of materials are developed. Such an ability can be helpful for characterizing the spatial distribution of different phases within the microstructure of multiphase systems or even in the identification of phases. Information on bulk phase transformations or second phase inclusions can offer key insight into the local conditions experienced by these materials during service with regards to, irradiation, strains, or temperatures. Such information is important to precisely model the behavior of these materials in the adverse service environments in which they are placed.
- B. Alexandreanu, B. Capell and G.S. Was (2001) “Combined Effect of Special Grain Boundaries and Grain Boundary Carbides on IGSCC of Ni–16Cr–9Fe–xC Alloys”, Materials Science and Engineering A 300, 94-104.
- J.F. Bingert, R.J. Hanrahan Jr., R.D. Field and P.O. Dickerson (2004) “Microtextural investigation of hydride a-Uranium”, J. Alloys and Compounds 365, 138-148.
- C. Cayron (2009). “Crystallographic Reconstruction Methods to Study Phase Transformations by EBSD.” Microscopy and Microanalysis 15: 396-397.
- Y. Chen, K. Sridharan, S. Ukai, T.R. Allen (2007) “Oxidation of 9Cr Oxide Dispersion Strengthened Steel Exposed in Supercritical Water”, J. Nuclear Materials 371, 118-128.
- B. Fournier, M. Sauzay, A. Renault, F. Barcelo, and A. Pineau, “Microstructural Evolutions and Cyclic Softening of 9% Cr Martensitic Steels”, J. Nuclear Materials 386-388, 71-74.
- R.A. Holt and P. Zhao (2004) “Micro-texture of extruded Zr-2.5Nb Tubes”, J. Nuclear Materials 335, 520-528.
- E.M. Lehockey, A.M. Brennenstuhl, S. Pagan, M.A. Clark and V. Perovic (2007) “New Applications of Orientation Imaging Microscopy (OIM) for Characterizing Nuclear Component Failure Modes, Reliability Assessment, and Fitness-for-Service”, Proc. 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems, ed. T.R. Allen et al., Toronto, ON, Canada: Canadian Nuclear Society.
- A. Medevielle, I. Hugon and O. Dugne (1999) Electron Backscatter Diffraction: Applications for Nuclear Materials”, J. Microscopy 195, 233-238.
- T.W. Nelson, J.C. Lippold and M.J. Mills (2000) “Nature and Evolution of the Fusion Boundary in Ferritic-Austenistic Dissimilar Metal Welds – Part 2: On-Cooling Transformations”, Supplement to the Welding Journal 79, 267-277.
- J.F. Nye (1985) Physical Properties of Crystals, Clarendon Press: Oxford.
- B. Raj, M. Vijayalakshmi, P.R.V. Rao, and K.B.S. Rao (2008) “Challenges in Materials Research for Sustainable Nuclear Energy”, MRS Bulletin 33: 327-337.
- B. Raj, D. Ramachandran and M. Vijayalakshmi (2009) “Development of Cladding Materials for Sodium-Cooled Fast Reactors in India”, Transactions of the Indian Institute of Metals 62, 89-94.
- J. Römer, M. Plaschke, G. Beuchle and J.I. Kim (2003) “In Situ Investigation of U(IV)-Oxide Surface Dissolution and Remineralization by Electrochemical AFM”. J. Nuclear Materials 322, 80-86.
- L. Tan, K. Sridharan and T.R. Allen (2006) “The effect of grain boundary engineering on the oxidation behavior of INCOLOY alloy 800H in supercritical water”, J. Nuclear Materials 348, 263-271.
- L. Tan, K. Sridharan and T.R. Allen (2008) “Microstructural Influence on the Corrosion Behaviour of Structural Materials for Nuclear Power Systems”, Proc. Structural Materials of Innovative Nuclear Systems (SMINS), OECD, Nuclear Energy Agency, 183-191.
- L. Tan, K. Sridharan, T.R. Allen, R.K. Nanstadt, D.A. McClintock (2008) “Microstructure Tailoring for Property Improvements by Grain Boundary Engineering”, J. Nuclear Materials 374, 270-280.
- K. Une, S. Ishimoto (2006) “EBSP Measurements of Hydrogenated Zircaloy-2 Claddings with Stress-Relieved and Recrystallized Annealing Conditions”, J. Nuclear Materials 357, 147-155.
- X. Wu, X. Pan, J. Stubbins (2007) “Analysis of Notch Strengthening of 316L Stainless Steel with and without Irradiation- Induced Hardening using EBSD and FEM”, J. Nuclear Materials 361, 228-238.
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