bluuebunny/crossref_metadata_2025 huggingface dataset:
The Correlation Between Microstructures and in-BWR Corrosion Behavior of Highly Irradiated Zr-based Alloys
Authors: Takagawa Y , Ishimoto S , Etoh Y , Kubo T , Ogata K , Kubota O
Published: April 2008

The microstructures of four kinds of Zircaloy-2-type materials with neutron exposure up to about 15 × 1025 n/m2 (E>1MeV) were examined to investigate the microstructural evolution and to correlate it with the in-reactor corrosion behavior at very high fluences. The materials examined in this study included two advanced Zr-based alloys with good corrosion resistance, namely High Fe Zry and High FeNi Zry, and two other kinds of Zry-2 materials with slightly different elemental compositions and fabrication processes producing differing second phase particle sizes. All alloys showed good corrosion behavior and low hydrogen pickup when irradiated up to four cycles in dummy neutron source holders located outside of channel boxes. After six irradiation cycles, all four of the alloys had an increase in the rate of uniform corrosion and hydrogen pickup. The results of the microstructure examinations after six-cycle irradiation indicated two interesting points: (1) increased insight into the dissolution process for Zr-Fe-Cr particles, and (2) a correlation between the number density of Zr-Fe-Cr particles surviving through irradiation and the corrosion behavior at high fluences. From these results, it is suggested that increasing the initial particle size and/or controlling the alloy chemistry would be beneficial to improve the uniform corrosion performance and to lower hydrogen pickup, especially at very high fluences.
In-Core Tests of Effects of BWR Water Chemistry Impurities on Zircaloy Corrosion
Authors: Shimada S , Cheng B , Lutz D , Kubota O , Ichikawa N , Ibe H
Published: April 2008

There have been several cases of fuel failures resulting from localized corrosion acceleration of the Zircaloy-2 cladding in BWRs over the last 25 years [1–3]. Some of the failures resulted from localized crud deposition, but others appear to be related to accelerated corrosion of the cladding, possibly due to certain chemical attacks of the cladding. The corrosion variability of Zircaloy-2 in BWRs has been attributed to a synergistic effect of Zircaloy-2 material variability, water chemistry impurities, and in-core service duty. Through extensive research efforts, the susceptibility of Zircaloy-2 cladding to nodular corrosion has been mitigated through improvements in the thermo-mechanical processing and surface finish since the mid-1980s. However, isolated cases of enhanced nodular corrosion have been reported recently in US BWRs [3]. Since such corrosion variability of Zircaloy-2 is undesirable, particularly for high burnup fuel components, candidate new Zr-alloys that are expected to have better nodular corrosion resistance than Zircaloy-2 have been developed [4–6].
Identification of Crystalline Behavior on Macroscopic Response and Local Strain Field Analysis: Application to Alpha Zirconium Alloys
Authors: Gélébart L , Crépin J , Dexet M , Sauzay M , Roos A
Published: April 2008

The purpose of this paper is to present an identification method of the crystalline behavior of a material from a mechanical test performed on a polycrystalline sample. Because of the lack of knowledge about its crystalline behavior, this method is applied to a Zirconium alloy. This identification is based on a finite element modeling of the microstructure, and the results are compared to both the macroscopic and the microscopic experimental results. On the microscopic scale, the plastic strains are obtained using a micro-extensometry technique and the crystalline orientation using an EBSD technique. In order to validate the method, an identification is performed with only two free parameters: the evolutions of the macroscopic and microscopic errors appear to be regular and exhibit a well-defined minimum so that the parameters can be clearly identified.
Ductility of Zircaloy-4 Fuel Cladding and Guide Tubes at High Fluences
Authors: Yagnik SK , Hermann A , Kuo R-C
Published: April 2008

Zircaloy fuel cladding suffers progressive degradation of ductility as its neutron exposure and hydrogen uptake increase with burnup. The loss of ductility appears to be the key property governing the cladding integrity in service. We report ductility data of Zircaloy-4 fabricated in stress-relief annealed (SRA) and recrystallized (RXA) conditions, covering a range of fluence, hydrogen content, and irradiation and test temperature.

The testing was performed on unirradiated and irradiated Zircaloy-4 cladding and guide tubes in SRA and RXA conditions, respectively, in the temperature range 25–350°C. These materials had been exposed to an estimated neutron fluence of ̃ 8 to 10 × 1025 n/m2 (E > 1 MeV) over four cycles of PWR operation. Due to its RXA fabrication condition, the guide tube material was also believed to represent BWR cladding. The hydrogen contents in the irradiated cladding was in the range of ̃ 200–600 ppm, exhibiting typical radial distribution of circumferential hydrides. By comparison, the hydrogen content in the irradiated guide tube material was in the range of ̃ 250–1800 ppm, exhibiting fairly uniform through-wall distribution of circumferential hydrides. Tensile tests and hydraulic burst tests were conducted on 130–150 mm long tubular specimens. In addition, smaller specimens machined in the form of 55 mm long curvilinear dog-bone and 10 mm slotted semicircular arc were tested in plane stress and plane strain configurations. Corresponding unirradiated archive materials in as-received condition and with uniform hydrogen charging up to 1200 ppm were also tested by identical methods. In all tests the fracture mode was examined by SEM fractography.

The investigations revealed a decrease in ductility of Zircaloy-4, mainly caused by irradiation and only partly by increasing hydrogen content. Unlike the elongation data, the strength data remained nearly constant with increasing hydrogen content in both materials at all test temperatures. The irradiated RXA material showed better ductility than irradiated SRA material at equivalent hydrogen levels, and exhibited a clearer correlation of ductility with hydrogen content, mainly due to its uniform hydrogen distribution. The paper will provide these and other quantitative data, e.g., those correlating ductility with the local (near fracture surface) hydrogen content. The paper synthesizes the experimental results and discusses their possible application to the criteria for hydrogen concentration and ductility limits in high burnup fuel.
Thermal Creep of Irradiated Zircaloy Cladding
Authors: Tsai H , Billone MC
Published: April 2008

As part of an effort to investigate spent-fuel behavior during dry-cask storage, thermal creep tests are being performed with defueled Zircaloy-4 cladding segments from two pressurized water reactors — Surry at ≈36 GWd/MTU burnup and H. B. Robinson at ≈67 GWd/MTU burnup, with corresponding fast (E > 1 MeV) fluence levels of 7×1025 and 14×1025 n/m2. The Surry rods are particularly relevant because they were stored in an inert-atmosphere (He) cask for 15 years. The Robinson rods were received after reactor discharge and pool storage. Commensurate with their high burnup, the Robinson cladding has significant waterside corrosion and hydrogen uptake. Test results to-date indicate good creep ductility for both claddings in the 360–400°C and 160–250 MPa (hoop-stress) regime. Partial recovery of radiation hardening may have occurred during the long tests at 400°C, which led to improved creep ductility. Creep-rate sensitivity is significant for stress and even more so for temperature. The higher hydrogen content in the Robinson material appears to have no detrimental effect on creep behavior at the test temperature. One Robinson sample, which ruptured in the weld region at 205°C during cooling from 400°C under stress (190 MPa), precipitated all visible hydrides in the radial direction.
Delayed Hydrogen Cracking Velocity and J-Integral Measurements on Irradiated BWR Cladding
Authors: Grigoriev V , Jakobsson R
Published: April 2008

Propagation of an axial crack has been characterized by means of the Pin-Loading Tension (PLT) test. The same experimental arrangement is used for the Delayed Hydrogen Cracking (DHC) and for the J-integral measurements. The DHC tests are performed under variable temperature and/or load conditions simulating a variable PCMI. Tests have been performed in air on unirradiated SRA Zircaloy-4 tubing and irradiated BWR Zircaloy-2 cladding containing from 30–350 ppm of hydrogen. Variable loading facilitates DHC: in the experiment at 300°C a crack velocity as high as 9.3∙10-7 m/s (80 mm/day) has been observed for irradiated cladding. V-K curves or diagrams of DHC crack velocity, V, as a function of stress-intensity factor, K, are suggested for different BWR claddings. Despite a limited amount of tests, the V-K curves appear to depend on the type of cladding. DHC does occur at much lower stress intensities compared to the Kc converted from the J0.2 values, confirming DHC to be the dominating secondary failure crack propagation mechanism in BWR cladding.
Overload Fracture of Flaw Tip Hydrides in Zr-2.5Nb Pressure Tubes
Authors: Shek GK , Cui J , Perovic V
Published: April 2008

CANDU Zr-2.5Nb pressure tubes are susceptible to a crack initiation mechanism known as Delayed Hydride Cracking (DHC), which is a process that involves hydrogen diffusion, hydride precipitation, hydride region formation, and fracture at a flaw tip. An overload occurs when the hydrided region is loaded to a stress higher than that at which this region is formed. Service-induced flaws are present in some pressure tubes, which can act as crack initiation sites. Most experimental data to assess DHC initiation are obtained under constant loading conditions in which hydride formation and fracture occur at the same load, and therefore they are not suitable to assess crack initiation under overload condition. A series of step-wise increasing load experiments was performed on unirradiated Zr-2.5Nb pressure tube samples to determine the fracture stress of hydrides formed at notches with 15 μm root radius under different hydride formation stresses and thermal histories. Crack initiation in the overload tests was detected by the acoustic emission technique. The notch tip hydride morphologies were examined by optical and scanning electron microscopy. Test results indicated that the resistance to overload fracture was dependent on the hydride formation stress and thermal histories, which affected the notch tip hydride size, density, and distribution. Overload tests were performed at different temperatures, and a transition temperature to high resistance to overload fracture was observed.
Effect of Irradiation on the Fracture Properties of Zr-2.5Nb Pressure Tubes at the End of Design Life
Authors: Pan ZL , St Lawrence S , Davies PH , Griffiths M , Sagat S
Published: April 2008

To determine the fracture properties of Zr-2.5Nb pressure tubes irradiated until the end of design life, cantilever beam, curved compact toughness, and transverse tensile samples were prepared from a typical pressure tube and irradiated in the high flux reactor OSIRIS at CEA, Saclay, France. Experiments were conducted on two batches of samples mounted in two irradiation inserts. Each insert held sixteen samples of each type of specimen. The first insert was irradiated to a fluence corresponding to approximately half of the design life in a CANDU3 reactor. The experimental results were reported in [1]. Samples in the second insert were irradiated for 10.5 years in OSIRIS and received a maximum neutron fluence of 2.61 × 1026 n/m2 (E > 1 MeV), being equivalent to 2.98 × 1026 n/m2 (E > 1 MeV) in a CANDU reactor, i.e., corresponding to ∼30 years operation in CANDU reactors at 80 % capacity factor. The present report describes the results of tensile, fracture toughness, and Delayed Hydride Cracking (DHC) tests and XRD microstructure analysis from the second batch of specimens. A continuous and gradual evolution in tensile, fracture, DHC properties, and dislocation densities is demonstrated without any evidence of a sudden change following the initial transitient at very low fluence. In the whole high fluence range, there is a very slow rate of increase in c-component dislocation density, strength, and DHC velocity and a slow reduction in elongation and Nb concentration in the β-phase. The a-type dislocation density and fracture toughness remain approximately constant. The results from the second insert of specimens confirm that, following the initial transient at very low fluence, there is little further change in the fracture properties of Zr-2.5Nb pressure tube material. Therefore, material properties behave in a stable and predicable manner to the end of a 30 years design life for CANDU reactor pressure tubes.
Fretting-Wear Behavior of Zircaloy-4, OPTIN™, and ZIRLO™ Fuel Rods and Grid Supports under Various Autoclave and Hydraulic Loop Endurance Test Conditions
Authors: King SJ , Young MY , Guérout FM , Fisher NJ
Published: April 2008

High fuel duty, eighteen-month and longer cycle lengths, and higher burnups have put increased demands on pressurized water reactor (PWR) fuel during recent years. As a result of these demanding operating conditions, the fretting-wear performance of PWR fuel rods against structural grid supports is being challenged. To understand the in-reactor fretting-wear performance of PWR fuel, Westinghouse Electric Company (W) and Atomic Energy of Canada Limited (AECL) have performed investigations to study the fretting-wear behavior of zirconium alloy nuclear fuel rods and grid supports. Experimental work was carried out at the AECL Chalk River Laboratories (Canada) in autoclave impact fretting-wear test machines and at the W Columbia, South Carolina (USA) plant in the Westinghouse — Vibration Investigation Pressure-Drop Experimental Research (VIPER) hydraulic test loop.

Autoclave impact fretting-wear tests were conducted for several grid-support/fuel-rod designs using either ZIRLO™, Zircaloy-4, and/or OPTIN™ materials. Rod-to-grid relative displacements and contact forces were measured throughout each test. The wear volumes of each fret mark on the fuel rod specimens were measured via surface profilometry at the end of each test. From these wear volumes, wear rates and wear coefficients for the fuel rod in contact with either springs or dimples were determined.

The VIPER loop is used for full-scale nuclear fuel assembly vibration and wear testing. Endurance tests were performed with severe flow conditions to induce fuel fretting-wear for different grid support and assembly designs using either all ZIRLO™ or all Zircaloy-4 materials. The wear volumes of the most severe fret marks on the fuel rods were measured via surface profilometry at the end of each test. From the profilometry of each fret mark, and using the fretting-wear coefficient derived from autoclave wear tests, the available energy (i.e., work-rate) to cause the fretting-wear damage was determined.

Fretting-wear scars on ZIRLO™, Zircaloy-4, and/or OPTIN™ fuel rods from the autoclave impact fretting-wear tests and the VIPER hydraulic tests were examined using scanning electron microscopy (SEM) to compare the fretting-wear mechanisms for different grid-support designs. Fretting-wear mechanisms from both types of tests were shown to be comparable and related to the grid-support design. For a given material (i.e., ZIRLO™, Zircaloy-4, or OPTIN™) under similar test conditions, there was no difference in the fretting-wear performance.
Failure of Hydrided Zircaloy-4 Under Equal-Biaxial and Plane-Strain Tensile Deformation
Authors: Glendening A , Koss DA , Motta AT , Pierron ON , Damt RS
Published: April 2008

The fracture behavior of unirradiated Zircaloy-4 containing either solid hydride blisters or hydrided rims has been examined for the contrasting conditions of equal-biaxial and plane-strain tensile deformation at three temperatures (25°, 300°, and 375°C). Cold-worked and stress-relieved Zircaloy-4 sheet containing hydride blisters shows nearly identical failure strains in equal-biaxial and plane-strain tensile deformation for a wide range of blister or rim depths. In all cases, failure strains decrease rapidly with increasing hydride blister or rim thickness, especially in the ⩽100 μm range. Test temperature has a significant effect on ductility with failure strains at 300° and 375°C being much greater than at room temperature. The results indicate that the ductility of material containing hydride rims/blisters greater than ≈ 30–40 μm deep is limited by crack growth, which occurs in a mode I manner at 25°C but in a mixed mode I/II manner at ⩾300°C (and at higher failure strain levels).
Mechanical and Physical Properties of Titanium-12Molybdenum-6Zirconium-2Iron Beta Titanium Alloy
Authors: Murray NGD , Jablokov VR , Freese HL
Published: April 2008

Howmedica developed the Titanium-12Molybdenum-6Zirconium-2Iron beta titanium alloy (also known as TMZF® beta titanium alloy) in the late 1980s. Two patents were issued for this high strength, low modulus, ductile, biocompatible, titanium base alloy [1,2]. The mechanical properties for wrought machining and forging bar in ASTM F 1813 show higher tensile and yield strengths than Ti-6A1-4V ELI (ASTM F 136), the most widely used titanium grade for medical and surgical implant applications. According to the ASTM standard, the TMZF alloy also has improved ductility over Ti-6A1-4V ELI, which is to be expected when comparing beta and alpha+beta titanium alloys. TMZF alloy offers higher tensile strength and greater flexibility than Ti-6A1-4Valloy. Data are presented for semi-finished TMZF mill product forms in the solution-annealed condition. The influence of oxygen as an interstitial strengthening element is shown for Ti-12Mo-6Zr-2Fe and compared with common titanium alloys used in or being considered for use in the medical device industry.
Influence of Oxygen Content on the Mechanical Properties of Titanium-35Niobium-7Zirconium-5Tantalum Beta Titanium Alloy
Authors: Jablokov VR , Murray NGD , Rack HJ , Freese HL
Published: April 2008

Titanium-35Niobium-7Zirconium-5Tantalum beta titanium alloy (also known as TiOsteum® beta titanium alloy) is a new metallic biomaterial that was designed to have outstanding osseointegratabilty, good mechanical strength, improved ductility, very low elastic modulus, and good hot and cold workability [1]. In this study, mechanical property data are presented from several TiOsteum alloy production lots with ingot oxygen contents ranging from 0.05 to 0.68%. Additionally, mill annealed mechanical property data collected over the last 10 to 15 years for the more common alpha, alpha-beta, and beta titanium alloys are presented as a function of oxygen content and compared to TiOsteum alloy. The influence of oxygen as an interstitial strengthening element is well established in titanium alloys. This study examines this effect over a broad range of oxygen values, and has significance for the application of the new TiOsteum alloy and other beta titanium alloys being considered for use in medical and surgical devices.
Beta Titanium Alloy Processed for High Strength Orthopedic Applications
Authors: Marquardt B , Shetty R
Published: April 2008

The general material requirements for Ti-15Mo have been standardized for surgical implant applications in ASTM F 2066 [1]. This particular standard is currently limited to one microstructural condition that is produced by beta solution treating and rapidly quenching the material to avoid the formation of alpha phase. When the alloy is processed in this way, it maintains a very high level of ductility but relatively low tensile and bending fatigue strengths. Since beta titanium alloys can be strengthened substantially by thermal processes that introduce alpha phase, it may be possible to adjust the microstructure to achieve a balance of properties that will meet the requirements for more highly stressed applications. Tensile and rotating beam fatigue data for several dual phase material conditions of Ti-15Mo indicate that the alloy can be strengthened substantially without introducing tensile notch sensitivity. However, it was necessary to use an alpha/beta annealing process and avoid beta solution treatment prior to aging at 480 °C to achieve this balance of properties. The potential benefits of alpha/beta processing appear to warrant expansion of ASTM F 2066 beyond the current beta solution treated and quenched condition.
The Application of Ti-15Mo Beta Titanium Alloy in High Strength Structural Orthopaedic Applications
Authors: Jablokov VR , Nutt MJ , Richelsoph ME , Freese HL
Published: April 2008

Titanium-15Molybdenum beta titanium alloy is a generic alloy that has been around for a long time, but has never found wide use in the aerospace industry. Beta titanium alloys, in general, and corrosion-resistant biocompatible titanium alloys, in particular, attracted a lot of attention in the medical and surgical device industry in the late 1980s and early 1990s. Until that time, all of the titanium grades used for medical implant applications were based on ASTM “F” standards modeled from aerospace industry and AMS specifications. The CP titanium grades (CP-1, CP-2, and CP-4) and two α + β alloys, Ti-6A1-4V ELI and Ti-6A1-4V, were widely used in aerospace applications and found early acceptance to become the reference metallic materials in ASTM F 748. Two important factors held Ti-15Mo back from commercial applications and use: the physical metallurgy of the alloy and the inability of the then-current reactive metals melting and processing machinery to handle this unusual binary alloy. This paper will briefly review high-technology manufacturing advances and processing technology that have enabled the reliable production of large-scale lots of Ti-15Mo bar and rod product forms. Additionally, mechanical and metallurgical data for semi-finished high strength rod product suitable for highly stressed orthopaedic implant applications will be presented.
Effect of Surface Reaction Layer on Three-Point Flexure Bond Strength of Resin Composite to Cast Ti and Ti-6AI-7Nb
Authors: Watanabe I , Liu J , Saimoto A , Griggs J , Okabe T
Published: April 2008

This study investigated the effect of the α-case on the three-point flexure bond strength of resin composite to different thicknesses of cast CP Ti and Ti-6Al-7Nb. Cast plates (0.3, 0.5, and 0.7 mm thick) were prepared for each metal. Plates without α-case were also prepared for the 0.5 and 0.7 mm thick CP Ti and Ti-6Al-7Nb. After resin composite was polymerized on the plates, the resin/metal specimens were subjected to three-point bending testing. The fracture force (N) and deflection (mm) at fracture were recorded. The fracture force of the specimens increased with increasing specimen thickness. Specimens without α-case withstood greater fracture force than specimens of either metal with α-case. CP Ti and Ti-6Al-7Nb without α-case deflected more than with α-case. The results suggested that the bonding between the veneer composite and thin cast structures (<0.5 mm) is likely to fail and that the surface reaction layer (α-case) should ideally be removed.
Super Elastic Functional β Titanium Alloy with Low Young's Modulus for Biomedical Applications
Authors: Niinomi M , Akahori T , Hattori Y , Morikaw K , Kasuga T , Fukui H , Suzuki A , Kyo K , Niwa S
Published: April 2008

The low modulus β type titanium alloy, Ti-29Nb-13Ta-4.6Zr, was designed, and then the practical level ingot of the alloy was successfully fabricated by Levicast method. The mechanical and biological compatibilities, and super elastic behavior of the alloys were investigated in this study. The mechanical performance of tensile properties and fatigue strength of the alloy are equal to or greater than those of conventional biomedical Ti-6A1-4V ELI. Young's modulus of the alloy is much lower than that of Ti-6A1-4V ELI, and increases with the precipitation of α phase or ω phase in the β matrix phase. The compatibility of the alloy with bone is excellent. Low modulus of the alloy is effective to enhance the healing of bone fracture and remodeling of bone. Super elastic behavior is observed in Ti-29Nb-13Ta-4.6Zr conducted with short time solution treatment after heavy cold working. Total elastic strain in that case is aroimd 2.8°. The mechanism of the super elastic behavior of Ti-29Nb-13Ta-4.6Zr is still unclear.
Comparative Evaluations of Surface Characteristics of cp Titanium, Ti-6AI-4V and Ti-15Mo-2.8Nb-0.2Si (Timetal® 21SRx)
Authors: Petersen DW , Lemons JE , Lucas LC
Published: April 2008

Commercially pure titanium (cpTi), Ti-6AI-4V (Ti64), and Ti-15Mo-2.8Nb-0.2Si (21SRx), with three unique atomic, alpha, alpha-beta, and beta grain structures, respectively, were subjected to three different surface treatments: cleaning, nitric acid passivation, and heat treatment. Experiments were conducted to determine the effects of the type of material and surface modifications on the substrate microstructure, surface oxide composition and thickness, and resultant corrosion behavior.

Metallography showed the cpTi groups were an equiaxed single alpha phase material, the Ti64 groups a dual-phase alpha-beta material, and the 21SRx groups an equiaxed beta material. The different surface treatments did not alter the substrate microstructures of any groups.

Spectroscopic (AES) results showed typical titanium and titanium alloy spectra with dominant Ti and O peaks for all sample groups, indicative of titanium dioxide. In addition, small Al and Mo peaks were detected throughout the surface oxides of the Ti64 and 21SRx specimens, respectively. AES depth profiling showed no significant difference in the oxide thickness between all the Cleaned and Passivated groups regardless of metal or alloy group. However, all the Heat Treated groups had significantly thicker oxides.

In general, corrosion results showed Passivated and Heat Treated groups to have similar corrosion properties and significantly improved corrosion resistances compared to the Cleaned groups. All impedance spectra fit into the Randies equivalent circuit model, and all sample groups exhibited near ideal capacitive behavior (ϕ 90°) expected for titanium and its alloys.
A Methodology to Fabricate Titanium and Stainless Steel Wear Debris for Experimental Use: A Comparison of Size, Shape, and Chemistry
Authors: Sprecher CM , Kunze J , Burian B , Villinger N , Jacobs JJ , Schneider E , Wimmer MA
Published: April 2008

It is well established that particulate debris can cause osteolysis. The current paper describes a simple procedure to generate titanium and stainless steel particulates for in vitro and in vivo use.

The ‘wear generator’ consisted of three pins out of titanium or stainless steel fixed onto a stir “bar.” The bar was rotated in a beaker filled with Ringer's solution against a disk of the same metal. The extracted particles were described using scanning electron microscopy, X-ray diffraction analyses, and inductively coupled plasma - optical emission spectroscopy. In addition, the chemical stability of the generated particles was tested.

The Equivalent Circle Diameter of stainless steel particles was smaller than that of titanium. Titanium particles contained pure titanium metal and titanium oxides. In contrast, stainless steel particles contained soluble corrosion products (e.g., nickel) and other elements from the salty lubricant.
Post-Irradiation Tensile Behavior and Residual Activity of Several Ferritic/Martensitic and Austenitic Steels Irradiated in Osiris Reactor at 325°C up to 9 dpa
Authors: de Carlan Y , Averty X , Brachet J-C , Bertin J , Rozenblum F , Rabouille O , Bougault A
Published: April 2008

An experimental irradiation, named “Alexandre,” has been carried out in the Osiris experimental reactor to perform a generic study on the mechanical behavior after irradiation at 325°C of different kinds of steels suitable for use as irradiated components in a nuclear reactor [1]. The irradiated steels were austenitic stainless, martensitic (conventional and reduced activation), and ferritic-martensitic Oxide Dispersion Strengthened steels in various initial metallurgical conditions. The final dose was 9 dpa, which represents nearly a “saturation” dose for the hardening/embrittlement of both austenitic and martensitic steels. At this dose, the Yield Strength and the Ultimate Tensile Strengths are almost equal, and strong localization of the plastic deformation is often observed.

After irradiation, as-quenched martensitic steels exhibit very large tensile strengths, and some of them show ductility parameters comparable with those of the tempered martensitic steels. At 9 dpa, the behavior of cold-worked steels tends to be similar to that of tempered steels with a more pronounced localization of the deformation. This indicates that the recovery of the cold-worked microstructure is not achieved at 9 dpa. Certain martensitic steels (MANET II and HT9) show tendencies to brittle behavior and exhibit a considerable degradation of their ductility parameters after irradiation. The behavior of the most chromium-rich ferritic steel (ODS-MA957) is quite surprising. It exhibits moderate hardening and good post-irradiation ductility parameters.

Measurements of the dose rate have been performed after several cooling-times, of up to 53 months after unloading, to study the radioactive decay of the steels. Experimental data show that the reduced activation steels exhibit the lowest residual activity and the highest rate of relative activity decrease (90 %) during the cooling period.
Development of Fuel Clad Materials for High Burn-up Operation of LWR
Authors: Cho HS , Kimura A , Ukai S , Fujiwara M
Published: April 2008

Oxide dispersion strengthening (ODS) steels have been considered to be very promising for fuel clad materials for high burn-up operation of light water reactor and super critical water reactor. For improving the corrosion resistance, increasing chromium concentration, which often resulted in degradation of mechanical properties and thermal embrittlement, has been required. In this work, the effects of aluminum addition on the mechanical properties of high chromium ODS steels have been investigated. Various ODS steels which have 14 ∼ 22 wt% Cr concentrations were made by mechanical alloying method. In order to evaluate the mechanical properties and the effect of thermal aging, a miniaturized Charpy V notch test and tensile test were performed before and after aging. 1) The addition of Al resulted in the decrease in anisotropy of tensile properties, although it reduced the tensile stress. 2) The impact properties of 19Cr-ODS steel are improved by the addition of Al.
Small Angle Neutron Scattering Study of Irradiated Martensitic Steels: Relation Between Microstructural Evolution and Hardening
Authors: Mathon MH , de Carlan Y , Averty X , Alamo A , de Novion CH
Published: April 2008

Martensitic/ferritic steels (containing 7–13 % Cr) are candidate materials for internal structures in pressurized water, fast breeder, and fusion reactors. Approval for use requires verification of structural stability under neutron irradiation in relation to the evolution of mechanical properties. In this context, several conventional and Reduced Activation (RA) martensitic materials were neutron irradiated at 325°C up to 6 dpa. They were investigated by Small Angle Neutron Scattering (SANS) under a magnetic field after various doses.

It was shown that when the Cr content of the b.c.c. ferritic matrix was larger than a critical threshold value (∼ 7.2 at.% at 325°C), the ferrite separated under neutron irradiation into two isomorphous phases, Fe-rich (α) and Cr-rich (α′). The kinetics of phase separation is much faster than under thermal aging. The quantity of precipitated α′ phase increases with the Cr content and the irradiation dose. In the case of steel with the lowest Cr content (F82H) irradiated at 5.6 dpa at 325°C, the α′ phase does not form, and the SANS signal suggests a small contribution due to vacancy clusters. It was believed that these could contribute to the “black dots” observed by TEM. Furthermore, we studied the microstructural features responsible for the secondary hardening phenomenon detected in the as-quenched F82H martensitic steel during irradiation or annealing.

In addition, the microstructural evolution of the Oxide Dispersion Strengthened (ODS) steel MA957, which presents an excellent hardening/ductility compromise after irradiation, has been also characterized. The stability of the oxides has been elucidated, and an important α′ volume fraction has been detected.

The contribution of α′ to the irradiation-induced hardening was assessed. This, although not negligible, is not the critical factor in normalized and tempered or cold-worked steels. However, it may be the main contribution to hardening in MA957.
Microstructural and Mechanical Characterization of Radiation Effects in Model Reactor Pressure Vessel Steels
Authors: Ulbricht A , Böhmert J , Viehrig HW
Published: April 2008

This paper presents results of radiation-induced nanoscale microstructural changes measured by Small-Angle Neutron Scattering (SANS) on ASTM-type reactor pressure vessel (RPV) steels. Five different base metals and one weld metal were investigated. The irradiation was performed in a VVER-type power reactor to three different dose levels, maximally 0.14 dpa, at an irradiation temperature of 255°C. The volume content of the micostructural features measured is correlated with the irradiation hardening and embrittlement.

In every case clear radiation-related hardening, embrittlement, and microstrucural effects are proven. Radiation produces nanoscale scattering defects of a radius between 0 < R < 3 nm with the maximum near R = 1 nm. Fluence and chemical composition do not or hardly change the shape of the size distribution but strongly influence the volume fraction. The radiation sensitivity is mainly influenced by the copper content. Copper also increases the A-ratio. There are good correlations between volume fraction of the nanoscale radiation defects and the radiation hardening or embrittlement.

Annealing at 475°C produces almost full recovery of the mechanical properties. However, the original microstructure of the unirradiated state is not completely reproduced.
Effects of Neutron Irradiation on Precipitation in Reactor Pressure Vessel Steels
Authors: Faulkner RG , Lu Z , Ellis D , Williams TJ
Published: April 2008

Several MnMoNi pressure vessel steels with varying Ni, Cu, and Mo contents have been investigated in the irradiated and unirradiated condition. Examination using analytical transmission electron microscopy was performed in order to understand more fully the nature of both the inter- and intra-granular precipitation in these alloys. In particular, attention has been paid to the carbide population, which exists in the material in addition to the copper precipitates, which have received detailed study elsewhere. The most dramatic revelations concern the discovery that fine, intra-granular (10 nm diameter) Fe3C particles are present in these steels, in addition to the normally expected copper-rich precipitates of a similar size. Neutron irradiation increases the volume fractions of inter- and intra-granular precipitation of all phases (Fe3C, Cu-rich, and Mo2C). Neutron irradiation has the least effect in increasing volume fraction in high Ni; high Cu alloys (WV). The effects of neutron irradiation on precipitation rates are greatest in low Ni; high Cu alloys (SG (SH)). Inter-granular precipitation is most prevalent in the high Ni; low Cu material (WF). It is significant that this material contains the least intra-granular fine-scale precipitation. This is probably because the Cu precipitation is minimal. Mo and C have an important influence on the amount of Mo2C phase formed, with large amounts being observed in high Mo-C, irradiated material (SH) (0.48 % Mo). These results are discussed in terms of various models for inter-and intra-granular precipitate nucleation and growth.
Assessment of Neutron Irradiation-Induced Grain Boundary Embrittlement by Phosphorous Segregation in a Reactor Pressure Vessel Steel
Authors: Kimura A , Shibata M , Kasada R , Fujii K , Fukuya K , Nakata H
Published: April 2008

The materials used were two sorts of reactor pressure vessel steels (RPVSs), which contain different amounts of impurity phosphorous (P) and copper (Cu). The specimens for Charpy V-notch impact tests and Auger electron spectroscopy (AES) were irradiated in the Japan Materials Test Reactor (JMTR) at 290°C up to fluences of 6 × 1021 and 1 × 1024 n/m2 using a multi-division temperature control capsule which enables removal of a part of the sub-capsules during operation of the reactor.

Neutron irradiation resulted in a significant shift in the ductile-brittle transition temperature (DBTT) accompanied by a large irradiation hardening in the high P and high Cu steel. The AES measurements following the irradiations revealed that almost no phosphorous segregation occurred at grain boundaries. The DBTT shifts by neutron irradiation were reasonably interpreted in terms of a so-called hardening mechanism. The SEM observations of the fractured surface indicated that a very small amount of grain boundary fracture was induced at the irradiation conditions, and resultantly no grain boundary embrittlement was observed.

Based on the results of irradiation experiments as well as long-term thermal aging experiments beyond 16 000 h, neutron irradiation-induced grain boundary embrittlement is considered to rarely happen to the RPVS that contains 110 wppm of phosphorous.
Mechanical Property Changes in Reactor Vessel Materials Thermally Aged for 209000 H at 282 °C
Authors: Gunawardane HP , Hall BJ , Rosinski ST
Published: April 2008

Surveillance materials consisting of a SA-508 Class 2 forging, a Mn-Mc-Ni Linde 80 submerged-arc weld, and an SA-533, Grade B, Class 1, correlation monitor material were thermally aged on a commercial reactor pressure vessel. The materials were exposed to a thermal environment of 282°C for 209 000 h. This temperature is below the range (minimum of 370°C) where the effects of long-term thermal aging are typically considered relevant. Charpy impact, master curve transition temperature, upper-shelf fracture toughness, and tensile testing were conducted to evaluate the long-term thermal aging changes in material properties. Small changes in the impact properties were observed for all the materials, but were generally within the 95 % confidence bounds for typical Charpy data. Upper-shelf energy also showed small variations, but a general decrease for all materials was not seen. Fracture toughness testing at the upper shelf indicated that the upper-shelf toughness had increased, however the data is scattered. Master curve T0 testing in the transition region showed little change in the forging and plate results; however an improvement in the transition temperature of the weld metal was measured.
Recent Surveillance Data and a Revised Embrittlement Correlation
Authors: Server WL , Lott RG , Rosinski ST , English CA
Published: April 2008

The mechanistically-guided embrittlement correlation model adopted in ASTM E 900-02 was based on US, surveillance results current through calendar year 1998. There now exists an extensive amount of new surveillance data that includes a large amount of boiling water reactor (BWR) results from a supplemental surveillance program designed to augment the plant-specific BWR surveillance programs. These recent data allow a statistical check of the ASTM E 900-02 embrittlement correlation, as well as the NRC correlation model currently being used in the pressurized thermal shock (PTS) re-evaluation effort and the older Regulatory Guide 1.99, Revision 2 correlation. Even though the ASTM E 900-02 embrittlement correlation is a simplified version of the NRC model, a comparison of the two embrittlement correlation models utilizing the new database has proven to be revealing. Based on the new BWR data, both models are inadequate in their ability to predict BWR results. Other aspects of the two models are presented, as well as plans to develop a revised embrittlement correlation.
Dynamic Finite Element Modeling of Fracture in Charpy V-Notch Specimens of Weld Material 72W
Authors: McCoy JH , Kumar AS
Published: April 2008

Dynamic finite element modeling of the fracture behavior of fatigue-precracked and V-notch Charpy specimens was performed using ABAQUS Explicit. The purpose of this modeling effort was to predict the upper shelf energies of full-size precracked and Charpy V-notch full-size and sub-size specimens based on the upper shelf energy (USE) of precracked specimens. To achieve this end, finite element model prediction of the USE of a precracked Charpy specimen was calibrated to the corresponding experimental data. Based on this calibration, a material model incorporating a stress-strain curve and fracture strain properties was developed for the nuclear reactor pressure vessel 72 material. The calibrated material model was then applied to the modeling of crack extension and propagation for predicting the USE of both full-size and sub-size Charpy precracked and V-notch specimens. Finite element modeling of Charpy V-notch models has not been performed previously using this methodology. It was found that the predicted USE of standard and sub-size specimens in both notched-only and precracked conditions based on the USE of third-size specimens were in reasonable agreement with the corresponding experimental data. However, when full-size precracked USE was used as the calibration point, full-size notched USE matched well, half and third-size precracked USE were within normal scatter and half and third-size notched USE did not acceptably match with the corresponding experimental values.
Correlated Formation and Stability of SIA Loops and Stacking Fault Tetrahedra in High Energy Displacement Cascades in Copper
Authors: Voskoboinikov RE , Osetsky YN , Bacon DJ
Published: April 2008

Atomistic modeling was conducted for an investigation of primary damage creation, self-interstitial and vacancy clusters formation, and their stability in high energy displacement cascades in copper. The simulations were carried out for a wide range of temperatures (100 K ≤ T ≤ 900 K) and primary knock-on atom (PKA) energies 5 keV ≤ Epka ≤ 25 keV. This study of over 400 cascades is the largest yet reported for this metal. At least 20 cascades for each (Epka, T) pair were simulated in order to ensure statistical reliability of the results. The number of surviving point defects for each cascade and the mean value for cascades at the same temperature and PKA energy were found. The corresponding fraction of self-interstitial atoms (SIA) in dislocation loops and vacancies in stacking fault tetrahedron (SFT)-like clusters was calculated. Strong spatial and size correlation of SFTs and SIA clusters at low temperatures were established.

In the context of high dose irradiation and the spatial overlap of displacement cascades, the stability of SFTs and dislocation loops inside an overlapping cascade region was investigated. It was observed that an SFT destroyed in the collision phase by a cascade is always recreated. On being completely enveloped by the region of displaced atoms, both SFT and SIA dislocation loops are destroyed with corresponding decrease of the number of residual point defects, whereas partial overlapping leads to increase in size of both types of cluster.
Flow Localization Processes in Austenitic Alloys
Authors: Wu X , Pan X , Li M , Stubbins JF
Published: April 2008

Austenitic alloys are widely used for structural component applications in high irradiation environments. In general, they are more resistant to embrittlement than other classes of structural alloys, particularly at ambient and intermediate temperatures. Nevertheless, this class of materials suffers from highly localized flow when irradiated to moderate dose (∼ 1 to 5 dpa) at temperatures between 150 and 400°C. The loss of ductility is normally exhibited by very low values of uniform elongation in tensile tests. The processes that lead to plastic instability are examined here for several face centered cubic materials and alloys. It is found that there is a critical stress level at which necking initiates. This critical stress level is not influenced by irradiation exposure. However, irradiation exposure, which increases material yield strength, does result in proportional reductions in uniform elongation. Most of the materials examined here exhibit a bilinear strain hardening behavior. This leads to direct correlation between the material yield strength and the uniform elongation.
Plutonium-238 Alpha-Decay Damage Study of A Glass-Bonded Sodalite Ceramic Waste Form
Authors: Frank SM , DiSanto T , Goff KM , Johnson SG , Jue JF , Barber TL , Noy M , O'Holleran TP , Giglio JJ
Published: April 2008

An accelerated alpha-decay damage study of a glass-bonded sodalite ceramic waste form has been completed recently. The study was designed to investigate the physical and chemical durability of the waste form after exposure to 238Pu alpha decay. The alpha-decay dose at the end of the four year study was approximately 1.0 × 1018 decays/gram of material. The ceramic waste form (CWF) was developed to immobilize fission products accumulated during the treatment of spent nuclear fuel from the Experimental Breeder Reactor II performed at Argonne National Laboratory-West in Idaho. Small quantities of actinide elements are also found in the waste form. The CWF is currently undergoing qualification testing and characterization for acceptance by the Office of Civilian Radioactive Waste Management for geologic disposal. Methods used to monitor the 238Pu-loaded CWF material in this study included: immersion density determination to measure possible macroscopic swelling, chemical durability by leach testing, microstructural analysis by scanning and transmission electron microscopy, and phase composition and stability by powder x-ray and electron diffraction.

After four years of exposure to alpha decay, the investigation observed little alteration to the CWF. Specifically, the 238Pu-loaded CWF maintained its physical integrity, the density remained constant, no cracking or phase debonding was observed by microscopy, and the material's chemical durability did not change significantly over the duration of the study. The only significant measured change was an increase of the unit-cell lattice parameters of the plutonium oxide and sodalite phases of the material, but these were very small and did not lead to any loss of waste form durability.
Mechanical Properties of Cubic Silicon Carbide after Neutron Irradiation at Elevated Temperatures
Authors: Katoh Y , Snead LL
Published: April 2008

Mechanical properties of high-purity polycrystalline cubic SiC was characterized after neutron irradiation. Bar samples were irradiated in target position capsules in the High Flux Isotope Reactor (HFIR, Oak Ridge, Tennessee) to nominal fluence levels up to 7.7 dpa at temperatures of 300, 500, and 800°C. A decrease in Young's modulus was observed after irradiation, and its irradiation temperature dependence qualitatively agreed with the calculated modulus change due to point defect swelling. Irradiation caused a significant modification of statistical flexural strength but caused only minor increase in nano-indentation hardness. The irradiation effect on fracture initiation through an enhanced cleavage resistance likely is primarily responsible for the major change in flexural strength properties.