bluuebunny/crossref_metadata_2025 huggingface dataset:
Atomistic Investigations of Intrinsic and Extrinsic Point Defects in bcc Uranium
Authors: Beeler Benjamin , Deo Chaitanya , Baskes Michael , Okuniewski Maria
Published: January 2013

Metallic alloys of uranium show great potential as transmutation fuels that could be used to burn long-lived and high-heat-producing minor actinides and fission products in nuclear reactors. In fuels, fission and radiation damage result in the production of a large number of intrinsic point defects and extrinsic fission atoms. Radiation damage and diffusion (processes heavily dependent on point defects), as well as fission product behavior, are important to the understanding of the behavior of these metallic fuel alloys. Of the fission products, fission gases Xe, Kr, and the decay product He are of special importance, as they migrate and form bubbles detrimentally affecting fuel properties. In this work, several systems of body-centered-cubic (gamma) U are examined through a semi-empirical interatomic potential based on the modified embedded-atom method. The vacancy formation energy is analyzed as a function of pressure and is used to determine the stable vacancy formation energy at ambient conditions. The vacancy formation energy as a function of temperature is analyzed for high-temperature systems. Interatomic potentials are developed and implemented in the investigation of He, Xe, and Kr point defects in the gamma phase of uranium. For all fission gases studied, the most energetically favorable location is the substitutional position, with helium having the lowest formation energies of the species investigated.
Application of ECLs and Today's Legislation
Authors: Laemmle Patrick , Rohrbach Peter
Published: January 2013

Environmentally considerate lubricants (ECLs)—the so called “bio oils”—were introduced in the mid 80s to minimize impact of lubrication on the environment. In the early 1990s the German RAL (Blue Angel) and the Swedish Standard organisations defined corresponding specifications for ECLs, other Eco-Labels followed. In 2005 the European Community released the directive 2005/360 defining the toxicity and ecotoxicity profile for ECLs, which qualifies them for the Euro-Marguerite. For more than 20 years top tier ECLs were used in mobile and stationary hydraulic equipment. Well documented field experiences not only demonstrate these lubricants contribute to the fulfillment of the new laws but also meet technical requirements for this equipment. In this paper important EC Directives related to lubricants and the protection of the environment are discussed and how they are implemented into national laws. In addition the benefits of ECL in hydraulic equipment are addressed.
Characteristics of Base Fluid in Environmentally Acceptable Lubricants
Authors: Kusak Brian , Wright Greg , Krol Ron , Bailey Matt
Published: January 2013

The main component in environmentally acceptable lubricants and grease is the base fluid. Typically it comprises greater than 70% of grease and 90% of a lubricant. Many biodegradable fluids have specific characteristics, which make formulating finished products difficult. Through a stepwise approach these characteristics will be discussed. The fluids for testing will comprise natural and synthetic esters and some petroleum based synthetics. The testing will find the extremes of performance at high and low temperatures and the effects of common additives to improve performance. Testing will include pour point, elastomer compatibility, oxidation and thermal stability. Based on the testing a blended fluid will be made to offset the drawbacks of the individual fluids and the performance discussed. This blend will serve as the base fluid to formulate an environmentally acceptable lubricating grease.
Compatibility of Vegetable Oil Based Lubricating Greases With Different Mineral Oil Based Greases
Authors: Kumar Anoop , Humphreys Steve , Mallory Bill
Published: January 2013

Applications of lubricants and greases based on vegetable or animal sources route back to horary past. Until the beginning of the 19th Century, lubricants were reported to be primarily based on beef tallow and vegetable oils. However, advances in petroleum derived mineral oils led to the replacement of these greases with mineral oil based lubricants. The reasons for their dominance may basically be attributed to high performance compared to vegetable oil based lubricants and greases and also their cheaper cost. Environmental consciousness and the speeding race for green lubricants has once again revived vegetable oil based lubricants. Compatibility of greases is an important property of greases that plays a crucial role in some applications like centralized lube system. As the base oil composition of mineral/synthetic oils is significantly different than vegetable oils, it might considerably influence the compatibility of the greases. In view of this, compatibility studies of vegetable oil based greases with mineral oil based greases have been carried out. The vegetable oil based greases formulated with lithium complex, lithium–calcium, and aluminum complex thickeners have been studied to determine the compatibility with non-vegetable oil based greases having lithium, lithium complex, aluminum complex, polyurea, calcium sulfonate thickeners, etc. However, the various greases covered in this paper are just base greases and do not contain any performance additives considering the fact that additives may significantly influence the compatibility of finished grease with other greases under investigation. It is interesting to note that the vegetable oil based greases evaluated so far have either been found incompatible or just borderline compatible.
Luminescent Bacteria as an Indicator Species for Lubricant Formulation Ecotoxicity
Authors: Sander John , Smith Terry , Bilberry Patrick
Published: January 2013

For a number of years, luminescent bacteria have been used as an indicator species for acute toxicity of substances in aquatic marine environments. Various other higher organisms have been used for this exact same purpose. Many of the higher organism test evaluations have been used successfully, but they have a few limitations, such as a low number of organisms exposed to toxicant during a test, less than desirable precision and the need for highly trained individuals to run the tests. One luminescent bacterial bioassay includes many organisms, can be run rapidly and requires training in sample dilution techniques. ASTM Subcommittee D02.12 has been working to develop an ecotoxicity method to evaluate the effects of lubricant formulations on luminescent bacteria. The theory describing this method, as well as the D02.12 method development activities, will be discussed.
Thermal Oxidative Stability of Vegetable Oils as Metal Heat Treatment Quenchants
Authors: de Souza Ester Carvalho , Belinato Gabriela , Simencio Otero Rosa L. , Adão Simêncio Éder Cícero , Augustinho Sylvana C. M. , Capelupi William , Conconi Charles , Canale Lauralice C. F. , Totten George E.
Published: January 2013

The potential use of vegetable oil derived industrial oils continues to be of great interest because vegetable oils are relatively non-toxic, bio-degradable, and a sustainably produced basestock alternative to petroleum oil, a non-renewable basestock. For the conservation of the environment, bio-mass materials, such as vegetable oils, are desirable as substitutes for petroleum oil in heat treatment. Therefore, it is expected that these basestock materials will continue to be of increasing interest in the heat treatment industry. However, the fatty ester components containing conjugated double bonds of the triglyceride structure of vegetable oils typically produce considerably poorer thermal oxidative stability than that achievable with petroleum base-stocks under typical use conditions. This is especially true when a vegetable oil is held in an open tank with agitation and exposure to air at elevated temperatures for extended periods of time (months or years). Furthermore, when used as quenchants, furnace loads of hot steel (850 °C) are typically rapidly immersed and cooled to approximately 50 °C to 60 °C bath temperatures for steel hardening applications. Clearly, for this application, reasonable thermal-oxidative stability is essential. This paper reviews the work completed thus far in screening various vegetable oils as potential steel quenchants both with and without antioxidants. Particular focus is placed, where possible, on comparing pressure differential scanning calorimetry as a potential screening method with the more commonly used (for this application) modified Indiana Stirring Oxidation Test. In addition, the general impact of oxidation on the quenching performance of the better vegetable oil candidates is shown.
The Development of Zr-2.5Nb Pressure Tubes for CANDU Reactors
Authors: Cheadle B. A.
Published: January 2013

The first reactor to use zirconium alloy tubes to contain hot pressurized water as a heat transport medium was the Hanford N reactor in 1962. For this reactor, three companies that had suitable extrusion and cold working equipment were given contracts to produce Zircaloy-2 tubes. All the companies were successful, and tubes from each company were installed in the reactor. When Canada decided to design and build a power reactor (nuclear power demonstration (NPD)), it elected to use the pressure tube concept and gave a contract to one of the companies (Chase Brass) to fabricate the Zircaloy-2 tubes. Douglas Point and Pickering Units 1 and 2 followed NPD, and all used similar Zircaloy-2 pressure tubes. A stronger tube was desired in order to thin the wall and improve the neutron economy. An alloy development program in the USSR had shown that the alloy Zr-2.5Nb looked very promising as a stronger alloy than Zircaloy-2, and both the USSR and Canada developed this alloy and subsequently used Zr-2.5Nb pressure tubes in their reactors. When both the Zircaloy-2 and Zr-2.5Nb pressure tubes were first installed in the reactors, several important properties and characteristics such as hydrogen ingress into the metal from corrosion in water, the enhancement of creep by neutron irradiation, shape change by neutron irradiation, reduction in fracture toughness by neutron irradiation, and delayed hydride cracking were not appreciated. These properties all led to problems during the service lives of the early tubes. Large research programs investigated these properties and showed that the α-grain size, shape, and crystal-lographic texture, the distribution of the β-phase, the dislocation type, and density as well as the micro-chemistry controlled the in-reactor properties of the tubes. This information enabled the design and operation of the reactors to be changed so that the tubes had satisfactory service lives. In addition, the information was used to fabricate tubes that had much improved properties and service lives.
Photoelectrochemical Investigation of Radiation-Enhanced Shadow Corrosion Phenomenon
Authors: Kim Young-Jin , Rebak Raul , Lin Yang-Pi , Lutz Daniel , Crawford Douglas C. , Kucuk Aylin , Cheng Bo
Published: January 2013

The photoelectrochemical investigation was performed by measuring the photoelectrochemical behavior of various alloys such as Zircaloy 2, 304 stainless steel (SS) and Alloy X-750 in 0.01M Na2SO4 at 25C or in high purity water at 300C under intense ultraviolet (UV) illumination. UV was selected because its photon energy (~5 eV) is similar to the energy gap of the electron-hole pairs in zirconium oxide. The data show that the photoexcitation of ZrO2 caused the corrosion potential to shift in the anodic direction and produced anodic photocurrents under the oxidizing water chemistry condition when Zircaloy 2 electrode was galvanically coupled with dissimilar electrodes, such as Alloy X-750, 304 SS, or Pt, causing accelerated corrosion of Zircaloy. It is thus postulated that the photoelectrochemical enhancement of surface reaction kinetics at the ZrO2 surface may be responsible for the radiation-enhanced corrosion on Zircaloy (i.e., shadow corrosion). In addition, it is revealed that no significant change in galvanic current between Zircaloy 2 and other dissimilar electrodes was measured in high temperature water containing only hydrogen that resulted in similar ECPs. These data thus clearly provide an explanation for why the radiation-enhanced corrosion of Zircaloy only occurs in BWRs but not PWRs. Based on galvanic corrosion, impedance, and transmission electron microscopy analysis, it is proposed that the defect structure of equiaxed grain layers may be responsible for the photoelectrochemical response of ZrO2 in high temperature water. The mechanism of shadow corrosion is still highly debated but it appears to be similar to a process of galvanic corrosion in connection with sufficiently conducting Zr oxide structure. The galvanic corrosion data suggests that a Zr coating on the fuel assembly spacer (Alloy X-750 or 304 SS) may mitigate the shadow corrosion in BWRs.
In Situ Studies of Variant Selection During the α—β—α Phase Transformation in Zr-2.5Nb
Authors: Mosbrucker P. , Daymond M. R. , Holt R. A.
Published: January 2013

In situ phase transformation experiments have been carried out using neutron and synchrotron X-ray diffraction to monitor the texture evolution and establish the criteria for variant selection during the α → β → α phase transformation in Zr-2.5Nb. There is clear evidence of a strong variant selection occurring during the α → β transformation and a weaker variant selection during the β → α transformation. Further, comparing hot-worked specimens with those that have received additional cold-work revealed that cold-work introduced microstructural characteristics that inhibited the full development of the transformed texture seen in the hot-worked samples. A study of the effect of external biasing stresses during transformation was conducted. An external biasing stress was applied along the hoop axis during different components of the α → β → α transformation. An external stress during heating does not appear to significantly influence variant selection, at least for the stress magnitudes tested here, while an external stress during cooling did have some influence on the resultant texture.
Improved Zr-2.5Nb Pressure Tubes for Reduced Diametral Strain in Advanced CANDU Reactors
Authors: Bickel G. A. , Griffiths M. , Douchant A. , Douglas S. , Woo O. T. , Buyers A.
Published: January 2013

In an Advanced CANDU Reactor (ACR) (ACR is a registered trademark of Atomic Energy of Canada Limited), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure, and flux than those in a CANDU (CANDU is a registered trademark of Atomic Energy of Canada Limited) reactor, and accordingly require a thicker wall (6.5 mm for ACR versus 4.2 mm for CANDU). In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a study to model and forecast the performance of these thicker pressure tubes has been undertaken. One of the main requirements for the pressure tube is to have low diametral creep. Based on previous experience with CANDU reactor pressure tube performance and manufacture, an assessment of the grain structure and texture of the ACR pressure tubes indicates that the in-reactor creep deformation will be improved. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor.
Microstructural Studies of Heat Treated Zr-2.5Nb Alloy for Pressure Tube Applications
Authors: Saibaba N. , Jha S. K. , Tonpe S. , Vaibhaw Kumar , Deshmukh V. , Ramana Rao S. V. , Mani Krishna K. V. , Neogy S. , Srivastava D. , Dey G. K. , Kulkarni R. V. , Rath B. B. , Ramadasan E. , Anantharaman S. A.
Published: January 2013

Zr-2.5Nb alloy is used as pressure tube material in pressurized heavy water reactors (PHWR). Generally, these pressure tubes are used in the cold drawn condition. Heat treated Zr-2.5Nb alloy pressure tubes are used in reaktor bolshoy moshchnosti kanalniy and FUGEN type of reactors. In recent times, there has been a greater interest toward increasing the life of pressure tubes in advanced reactors. In the present work, fabrication parameters were optimized to manufacture heat treated Zr-2.5Nb alloy tube. A quenching dilatometer study was performed to establish the continuous cooling transus temperature for the alloy used in this study. Heat treatment under controlled condition in a dilatometer was performed to study microstructure at different soaking temperatures and cooling rates. In the dilatometer, during gas quenching, quenching rates were varied from 0.06 to 100C/s to assess the effect of cooling rate on resulting microstructures. Soaking temperature and cooling rates were varied to obtain martensitic microstructure with appropriate volume fractions of primary α. On the basis of the results obtained during controlled heat treatments performed in the quenching dilatometer, 883C was selected as the soaking temperature and water as the quenching medium for the α + β quenching operation for large dimension tubes. The α + β quenched microstructures, consisting of fine martensite phase along with 20%–25% primary α volume fraction, were used for further cold deformation and subsequent aging below the recrystallization temperature. Aging at 540C produced fully recovered and tempered structure consisting of βNb of equilibrium composition. Mechanical properties of the finished heat treated pressure tube (aged at 515C/24 h) produced with the present route were similar to the cold work pressure tube. Tube produced with 540C/24 h aging exhibited substantially higher yield strength value at reactor operating temperature (300C). The bulk texture at different stages of fabrication was evaluated. The volume fraction of the primary α phase significantly controls the final texture.
High Temperature Aqueous Corrosion and Deuterium Uptake of Coupons Prepared from the Front and Back Ends of Zr-2.5Nb Pressure Tubes
Authors: Nordin H. M. , Elliot A. J. , Bergin S. G.
Published: January 2013

Pressure tubes for CANada Deuterium Uranium (CANDU) reactors are extruded from billets of Zr-2.5Nb, at a temperature of ∼815C, and then cold drawn to give a final length of ∼6 m. The manufacturing process often results in a variation of properties along the length of a tube including grain structure, texture, dislocation density, and phase distribution. This variation affects the mechanical and deformation properties as well as the aqueous oxidation and deuterium uptake behavior along the installed pressure tube. The orientation of the installed pressure tube in the reactor, with its axial variation of properties, is an important factor in the effective optimization of its service life. This work reports on the differences in aqueous oxidation and deuterium uptake between the extruded front- and back-end sections of a number of pressure tubes. The corrosion tests were conducted in heavy water in static autoclaves at Chalk River Laboratories and in a heavy water re-circulating loop in the Halden Boiling Water Reactor. The test conditions, such as water chemistry and temperature, were similar to those in the primary heat transport system of a CANDU reactor. The results indicate that under some exposure conditions, the deuterium uptake may beup to 40 % lower for back-end coupons compared to front-end coupons. Several microstructural factors including texture, grain size, and concentrations of alloying elements may cause the observed differences in deuterium uptake. The results will be discussed within the current mechanistic understandings of Zr-2.5Nb corrosion and deuterium ingress.
In Situ Scanning Electron Microscope Observation and Finite Element Method Analysis of Delayed Hydride Cracking Propagation in Zircaloy-2 Fuel Cladding Tubes
Authors: Kubo Toshio , Muta Hiroaki , Yamanaka Shinsuke , Uno Masayoshi , Ogata Keizo
Published: January 2013

The objective of the present research is to build a modeling method for delayed hydride cracking (DHC) of zirconium alloys. DHC tests were carried out on Zircaloy-2 cladding tubes in the chamber of a scanning electron microscope to directly observe the crack propagation and measure the crack velocity in the radial direction. These in situ observations showed that a sharply tipped crack propagated at a relatively high rate, while the velocity decreased when the crack tip was blunted, supporting the occur-rence of intermittent crack propagation that could be expected from the DHC mechanism. V-KI curves or diagrams of crack velocity, V, versus stress intensity factor at a crack tip, KI, were obtained as a function of 0.2 % offset yield stress, hydride orientation, and pre-crack depth. The steady state crack velocity and the threshold stress intensity factor for the onset of the crack propagation tended to increase or decrease, respectively, with an increase in the 0.2 % offset yield stress. Analyses of stress distribution and hydrogen diffusion around a crack tip were made using a finite element computer code. The analyses showed that a strong hydrostatic pressure field was generated concentrically around the crack tip and hydrogen diffused towards the crack tip according to the hydrostatic pressure gradient. The crack velocity was estimated from the calculated hydrogen flux rate assuming the critical hydrogen quantity for the crack propagation. There was good agreement between the experiments and the calculations regarding the crack velocity and its dependency on KI. Calculations showed that the increase in the 0.2 % offset yield stress would accelerate the crack propagation by increasing the hydro-static pressure at the crack tip.
Study on the Role of Second Phase Particles in Hydrogen Uptake Behavior of Zirconium Alloys
Authors: Yao M. Y. , Wang J. H. , Peng J. C. , Zhou B. X. , Li Q.
Published: January 2013

In an effort to better understand the role of second phase particles (SPPs) in the hydrogen uptake of zirconium alloys, four alloys and four heat treatments for each alloy were chosen to prepare specimens with different SPPs size distributions and area fractions. The hydrogen uptake performance of these specimens was investigated after autoclave testing in 400C/10.3 MPa steam. Results show that the hydrogen uptake is not always in a strict corresponding relationship with the corrosion resistance among the specimens, but it is closely related to the size, area fraction, and compositions of the SPPs. In the case of Zry-2 and Zry-4, the hydrogen uptake fraction (HUF) increased with increasing size and area fraction of the SPPs. The dependence was more notable for the Zry-2 than the Zry-4. In the case of N36 and N18, the HUF had only a slight variation with the size and area fraction of the SPPs. No matter which heat treatment was employed, the corrosion resistance of the N18 specimens was superior to the N36 specimens, but the HUF of the former was larger than that of the latter. These results clearly demonstrate that the effect of the size and area fraction of SPPs on the hydrogen uptake depends on the SPP compositions. Pressure-composition-temperature and kinetics of absorbing and desorbing hydrogen tests were conducted on Zr(Fe,Cr)2, Zr2(Fe,Ni), Zr(Nb,Fe)2, and β-Nb alloys (which may be found as SPPs in the four zirconium alloys tested) as well as on pure zirconium. Results show that Zr(Nb,Fe)2, Zr2(Fe,Ni), and Zr(Fe,Cr)2 alloys have a stronger reversible ability for hydrogen absorption and desorption than β-Nb alloy and pure zirconium. Based on the testing results, a model correlating the hydrogen uptake performance to the reversible ability of the SPPs to absorb and desorb hydrogen is proposed. The model can successfully explain the results.
Hydride Platelet Reorientation in Zircaloy Studied with Synchrotron Radiation Diffraction
Authors: Colas Kimberly B. , Motta Arthur T. , Daymond Mark R. , Kerr Matthew , Almer Jonathan D.
Published: January 2013

Hydrogen ingress into zirconium alloy fuel cladding in light water reactors can degrade cladding performance as a result of the formation of brittle hydrides. In service, hydrides normally precipitate in the circumferential direction and are homogeneously distributed through the cladding thickness in ideal cases. However, temperature and stress gradients in the cladding can promote hydrogen redistribution. This hydrogen redistribution is responsible for the formation of hydride rims, dissolution, and reorientation of hydride precipitates and for the formation of brittle hydrides at stress concentration locations, all of which can reduce cladding resistance to failure. Thus, it is crucial to understand the kinetics of hydride dissolution and precipitation under load and at temperature. Studies of hydrogen behavior in zirconium alloys are normally performed post facto, which causes them to suffer both from a scarcity of data points and from the confounding effects of studying hydrides at room temperature that might be dissolved at higher temperature. In the current study, we have used synchrotron radiation diffraction to study the kinetics of hydride precipitation and dissolution in situ (under load and at temperature). Samples of hydrided Zircaloy-4 were examined in transmission by using 80 keV synchrotron radiation while undergoing heating and cooling in a furnace. Temperatures ranged from 20 to 550C, and loads from 75 to 100 MPa were applied. The hydrides dissolved and reprecipitated in a different orientation when sufficiently high loads were applied. Through careful study of the intensities and full-width half maxima of the diffraction peaks as a function of time, load, and temperature, it was possible to identify the characteristic diffraction patterns for the reoriented hydrides so that the kinetics of dissolution, reprecipitation, and orientation of the hydrides could be followed. The analysis of the diffraction patterns allowed a detailed understanding of the kinetics of hydride evolution under temperature and stress, as presented in this work.
The Effect of Microstructure on Delayed Hydride Cracking Behavior of Zircaloy-4 Fuel Cladding—An International Atomic Energy Agency Coordinated Research Program
Authors: Coleman C. , Grigoriev V. , Inozemtsev V. , Markelov V. , Roth M. , Makarevicius V. , Kim Y. S. , Liaqat Ali Kanwar , Chakravartty J. K. , Mizrahi R. , Lalgudi R.
Published: January 2013

The rate of delayed hydride cracking (DHC) has been measured in Zircaloy-4 fuel cladding in several metallurgical conditions using the pin-loading tension technique. In light water reactor (LWR) cladding in the cold-worked and cold-worked and stress-relieved conditions, the cracking rate followed Arrhenius behavior up to about 280 °C, but at higher temperatures the rate declined with no cracking above 300 °C. Non-LWR cladding appeared to behave in the same manner. In LWR cladding in the recrystallized condition, the cracking rate was highly variable because it depended on KI within the test range up to 25 MPa√m, whereas in the other LWR claddings, cracking rate was independent of KI, indicating that KIH was below 11 MPa√m. The main role of microstructure was to control the material strength; the cracking rate increased as the strength increased. Although all the claddings had a radial texture, it did not protect the cladding from DHC. The DHC fracture surface consisted of flat broken hydrides, often in arcs, but no striations were observed, except in one specimen subjected to thermal cycles.
Study of the Initial Stage and Anisotropic Growth of Oxide Layers Formed on Zircaloy-4
Authors: Zhou B. X. , Peng J. C. , Yao M. Y. , Li Q. , Xia S. , Du C. X. , Xu G.
Published: January 2013

An in situ investigation of the epitaxial oxide layer formed on a thin specimen heated in a transmission electron microscope was carried out. Some dot-like grains about 10 nm in size were formed on the surface of a relatively thin area. The dot-like grains are monoclinic zirconium oxide and have an orientation relationship of (001)m//(01¯11)α-Zr, (001)m//(11¯01)α-Zr, with the α-Zr matrix. Some long strip-like grains, probably a new kind of zirconium suboxide, were formed on the surface of a relatively thick area. The strip-like grains have a bcc structure with a lattice parameter a=0.66 nm and have an orientation relationship of (110)bcc//(10110)α−Zr, [1110]bcc//[0001]α−Zr with the α-Zr matrix. The relationship between the thickness of oxide layers and the grain orientations of the α-Zr matrix was studied with coarse-grained Zircaloy-4 specimens through autoclave corrosion tests at 500 and 400°C in superheated steam, and at 360°C in both lithiated and deionized water for long time exposure. The results show that the anisotropic growth of oxide layers on the grain surface with different orientations is considerable. However, the relationship between the thickness of oxide layers and the grain orientations of the α-Zr matrix varies with corrosion temperature and water chemistry. The largest variation of oxide thickness developed during corrosion tests at 500°C. The thickest oxide layers were formed on those grains whose surface orientations were distributed around the planes from (011¯0) to (1¯21¯0). The thicker oxide layers on these grains were further developed into nodular corrosion. When the specimens were corroded at 360°C in lithiated water, the thickest oxide layers formed on those grains, whose surface orientations tilted from 15 to 30° away from the (0001) plane. When the specimens were corroded at 400°C in superheated steam and at 360°C in deionized water, the difference between the thickness of oxide layers on different grain surfaces was less prominent.
Hydrogen Solubility and Microstructural Changes in Zircaloy-4 Due to Neutron Irradiation
Authors: Vizcaíno P. , Flores A. V. , Bozzano P. B. , Banchik A. D. , Versaci R. A. , Ríos R. O.
Published: January 2013

The present work deals with two essential degradation mechanisms of zirconium alloys: Hydrogen behavior and microstructural transformations due to neutron irradiation. The studies were carried out by using samples taken from Zircaloy-4 components, which remained for 14 years in the Atucha I pressurized heavy water reactor, accumulating neutron fluences of up to 1022 neutrons/cm2. A hydrogen/deuterium solubility increase was observed in both dissolution and precipitation processes. With the aim to follow this evolution with radiation damage recovery, the irradiated samples were annealed for different periods at 350 and 400°C in a differential scanning calorimeter (DSC). These temperatures were chosen to reproduce those reached by spent fuels in the early stages of the dry storage (DS). Between intervals, the terminal solid solubility temperatures in dissolution (TTSSd) and in precipitation (TTSSp) were determined. They were also determined after additional annealing at 500 and 600°C, trying to simulate long periods at DS. The hydride morphology and the metallurgical state of the Zr(Fe, Cr)2 second-phase particles (SPPs) were followed between annealing by optical microscopy, analytical electron microscopy, synchrotron light X-ray diffraction4 (SXRD), and DSC. As a brief summary, we found from the first to the last run of the annealing sequence an increase of 55°C in TTSSd and 40°C in TTSSp on average. The optical and transmission electron microscopies (TEM) indicate that in the material in the irradiated condition, exists a hydride size distribution of four orders of magnitude, from 10 nm to 100 μm, but biased to sizes smaller than 5 μm. A complete amorphisation of the SPPs was observed in high fluence samples by TEM and SXRD. Similar TEM observations made in irradiated samples annealed above 400°C show a re-precipitation process of nano-SPPs forming clusters. The re-precipitation is affected by kinetics: The SPPs crystallization temperature, Tc, rises from 450 to 500°C as the heating rate increases from 5 to 20°C/min. The measured heat of crystallization was 27.3±2.1 kJ/mol (Fe + Cr).
Advanced Zirconium Alloy for PWR Application
Authors: Garde A. M. , Comstock R. J. , Pan G. , Baranwal R. , Hallstadius L. , Cook T. , Carrera F.
Published: January 2013

Westinghouse is evaluating several advanced zirconium-based alloys, designated collectively as AXIOM™, to achieve improved performance for more demanding fuel management schemes. There are five candidate AXIOM alloys currently being evaluated by Westinghouse. The in-pressurized water reactor (PWR) performance of one candidate alloy, X5A, is reviewed in this paper. The irradiation performance of X5A (previously identified as Alloy A) cladding that was fabricated using high-temperature processing (HTP) was published in 2002. Since then, the fabrication process of X5A was optimized by the use of a low-temperature process (LTP). Cladding tubes with the improved processing have been irradiated in two commercial PWRs (PWR A and PWR B) and in two test reactors (Test Reactor C and Test Reactor D). The irradiation performance of both versions (HTP X5A and LTP X5A) is reviewed in this paper with a primary emphasis on the current LTP cladding. After achieving an intermediate burnup in the range of 48–54 GWd/Metric Ton of Uranium (MTU) in PWR A, the average maximum oxide layer thickness for LTP X5A was about 23 μm or about 27 % lower than the oxide thickness on ZIRLO® clad fuel rods. In addition, the fuel rod axial growth strain for LTP X5A was about 50 % of ZIRLO rod growth. Lead fuel rod irradiation of LTP X5A in PWR B with burnup in the range of 47–53 GWd/MTU showed about 30 % lower corrosion relative to ZIRLO rods and a 7 % lower axial rod growth strain than ZIRLO. An irradiation experiment in Test Reactor C was designed to study breakaway irradiation growth (in the absence of waterside oxidation) of several alloys. LTP X5A cladding showed a growth strain of about 20 % that of ZIRLO cladding at a fluence of 16 × 1025n/m2. In Test Reactor D, at a burnup of about 44 GWd/MTU, HTP X5A had the same oxide thickness as ZIRLO. However, the post-irradiation hydrogen pick-up was 35 % lower for HTP X5A compared to ZIRLO. In addition to the irradiation experience, the supplemental out-reactor autoclave evaluation of X5A welds indicates adequate weld corrosion resistance. While additional in-PWR exposures are required along with post-irradiation examination, the results to date demonstrate that X5A is a promising alloy for future PWR application.
Radiation Damage of E635 Alloy Under High Dose Irradiation in the VVER-1000 and BOR-60 Reactors
Authors: Kobylyansky G. P. , Novoselov A. E. , Obukhov A. V. , Ostrovsky Z. E. , Shishov V. N. , Peregud M. M. , Markelov V. A.
Published: January 2013

Alloy E635 is used as a material for some parts of the water-water energy reactor (VVER)-1000 fuel assemblies (FAs). The evolution of structural components and redistribution of the elementary composition that occurred between the solid solution and phase precipitates in the E635 alloy were analyzed. The correlation between these changes and the irradiation-induced deformation of FA parts was determined. The common features of the E635 alloy's irradiation damageability was established as a result of the irradiation of model samples in the test reactor BOR-60 and the operation of products in the VVER-1000 reactor. The peculiarities of E365 alloy's irradiation damageability under high dose irradiation were revealed. The Laves phase (Zr(Nb,Fe)2) particles were found to be the main type of secondary phase precipitates observed in the E635 alloy products with completely or partially recrystallized structures. Under irradiation, iron released from this phase causes a transformation of its crystalline structure, i.e., it changes from hexagonal-close-packed into body-centered-cubic with an Fe-enriched β-Nb phase. The Fe content in the particles decreases as the dose rises. The above transformations are not observed on those FA areas where the damage dose is low. Some bigger (up to 0.50 μ) precipitates of the (Z,Nb)2Fe phase with face-centered-cubic lattices were found in the material structure. Irradiation up to high damage doses results also in the appearance of secondary fine-dispersed (up to ~5 in size) irradiation-induced precipitates. Practically no niobium is observed in the matrix, while all tin is in the solid solution and the Fe fraction in the matrix rises as the fluence becomes higher. The generation and formation of 〈c〉-dislocations occur only near the β-Nb precipitates (former Laves phases); 〈a〉-dislocations in the form of dislocation loops 10 nm in size and black dots are observed over the whole volume of recrystallized grains. On the whole, the changes in the structural-phase state of the FA parts tested in the VVER-1000 reactor correspond to the ideas about irradiation-induced damage in E635 alloy gained in experiments on irradiation of model samples in the BOR-60 reactor.
REFLET Experiment in OSIRIS: Relaxation under Flux as a Method for Determining Creep Behavior of Zircaloy Assembly Components
Authors: Carassou S. , Duguay C. , Yvon P. , Rozenblum F. , Cloué J. M. , Chabretou V. , Bernaudat C. , Levasseur B. , Maurice A. , Bouffioux P. , Audic K.
Published: January 2013

A stress relaxation on bent-beam specimen irradiation campaign was performed in the French material testing reactor OSIRIS in order to screen different materials according to their in-flux behavior and to derive constitutive laws that are used to describe the in-reactor behavior of assembly components under axial stress. This paper presents the methodology used, with particular emphasis on the validation of the different hypotheses. The methodology is illustrated with the results obtained on an industrial Zircaloy-4 (Zy-4) alloy used in guide thimbles. As an illustration of the capability of the method, the parameters of simple creep behavior laws are adjusted to the relaxation results. The creep-law predictions are compared to results obtained previously in a creep experiment.
Characterization of Oxygen Distribution in LOCA Situations
Authors: Duriez C. , Guilbert S. , Stern A. , Grandjean C. , Bělovský L. , Desquines J.
Published: January 2013

Oxygen distribution in the rod cladding tubes during a Loss of Coolant Accident (LOCA) situation is known to play a key role regarding the post-quench mechanical properties of fuel rods. Analytical solutions of the diffusion equations in the complex multilayer and multiphase systems describing the cladding material tube during high temperature oxidation are limited to stationary isothermal conditions. To address the problem of non-stationary transient LOCA situations, a numerical tool (named DIFFOX) has been developed. In the first part, the main assumptions and main features of the code are described. In parallel to the code development, an experimental program has been launched to provide a validation database as well as to input data for improvement of the code models. Zircaloy-4 cladding samples have been pre-oxidized at low temperature then vacuum annealed or steam oxidized at high temperature. Oxygen distribution profiles are investigated by Electron Probe Microanalysis. The ability of the code to describe the chemical reduction of the pre-oxide layer and oxygen diffusion into the metal during high temperature annealing is demonstrated. Regarding high temperature steam oxidation, the pre-oxide layers are observed to be either protective or not on a same given sample. While a good code to experimental data agreement is obtained in the first case, the growing of a high temperature oxide underneath the pre-oxide is not well reproduced by the code. The effect of hydrogen has also been investigated by performing small scale oxidation tests on 600 wppm hydrogen charged Zy-4 cladding. Samples have been long-term annealed for homogenization after oxidation. The main purpose of these experiments is to provide quantitative data on the modification of the α/α+β and α+β/β transus of the Zy4-O pseudo binary phase diagram induced by the presence of hydrogen. A variation of the transus locations deduced from these experiments is proposed.
Effect of Hydrides on Mechanical Properties and Failure Morphology of BWR Fuel Cladding at Very High Strain Rate
Authors: Nakatsuka Masafumi , Yagnik Suresh
Published: January 2013

The data on the cladding mechanical response under rapid deformation are crucial for the safety assessment of LWR fuel rods for reactivity initiated accident (RIA) and for spent fuel shipping and storage cask-drop accidents. The reported data on the mechanical properties of the irradiated cladding under high strain rates, however, are very limited, and conventional axial tensile data measured at slow strain rates may not be directly applicable to RIA or drop impact analyses. The objective of this work was to obtain basic mechanical properties such as the stress-strain relationship, ductility, and the critical strain energy density (SED) of fuel cladding with various hydride morphologies at high deformation rate. A unique rapid burst test method was employed on open-end tube specimens, achieving a strain rate of ~1 s−1 and specimen rupture in ~10 ms. Unirradiated and irradiated Zry-2 fuel cladding specimens were tested from room temperature up to 350C. The mechanical properties data were assessed in terms of SED and failure strain, and failure morphologies were examined by optical and scanning electron microscopy. The tests performed at room temperature on unirradiated cladding with circumferential hydrides revealed that the failure elongation was unchanged or increased slightly with an increase the in hydrogen content up to ~100 wt ppm but decreased drastically at ~400 wt ppm to values approaching 0 %. A drastic reduction in failure strain occurred at a lower hydrogen concentration of 100 wt ppm for radial hydrides, compared to ~200–400 wt ppm for the circumferential hydride distribution. The fraction of specimen thickness occupied by the accumulated length of hydrides in radial direction was a better predictive indicator of specimen failure and ductility reduction than the total hydrogen content. In addition, the high strain rates did not seem to seriously impact stress-strain behavior when hydrogen content is >400 wt ppm. The data analyses revealed smaller values of the strain hardening exponent (n) compared to those from the conventional data for the slow strain rates, indicating that the plastic instability theory simply is not appropriate to evaluate the failure strain under RIA conditions.
Simulation of Outside-in Cracking in Boiling Water Reactor Fuel Cladding Tubes under Power Ramp
Authors: Sakamoto Kan , Nakatsuka Masafumi , Higuchi Toru
Published: January 2013

In power ramp tests, some high burn-up boiling water reactor (BWR) fuels failed by outside-in cracking. After detailed post irradiation examinations, it was deduced that fracture of radially oriented hydrides precipitated at the outer surface had resulted in the formation of incipient radial cracks that propagated towards the inner surface. Since crack propagation is a key process that dominates the outside-in failure, experimental data of outside-in cracking are obviously required. The objective of the present study is to evaluate the threshold conditions and the velocity of the outside-in cracking during a power ramp to assess the integrity of BWR fuels during a power ramp. In the experiments, by using unirradiated Zry-2 fuel cladding tubes, the outside-in cracking was examined both with and without the thermal gradient. Without the radial thermal gradient, there was no radial delayed hydride cracking in all experimental conditions examined. On the other hand, with the radial thermal gradient, the velocity of the outside-in cracking was significantly higher than that expected from the data obtained under the isothermal condition in our previous study. The threshold depth of the incipient crack was found to be ~0.1 mm at the hoop stress of 300 MPa. It was suggested that the outside-in cracking during the power ram\p is strongly dependent on the distribution of dissolved hydrogen as a result of the thermal diffusion.
In Search of Standard Reference Materials for ASTM E05 Fire Standards
Authors: Alvares Norman J. , Hasegawa Harry K.
Published: January 2013

The first paragraph of E-691, Standard Practice for Conducting an Inter-laboratory Study to Determine the Precision of a Test Method [ASTM E691-05, 2005, “Standard Practice for Conducting an Inter-laboratory Study to Determine the Precision of a Test Method,” ASTM Book of Standards, ASTM International, West Conshohocken, PA.] states: “Tests performed on presumably identical materials in presumably identical circumstances do not, in general, yield identical results. This is attributed to unavoidable random errors inherent in every test procedure; the factors that may influence the outcome of a test cannot all be completely controlled.” Further in the same paragraph the factors are identified: “Many different factors (apart from random variations between supposedly identical specimens) may contribute to the variability in application of a test method, including: a. the operator, b. equipment used, c. calibration of equipment, and d. environment.” The primary subjects of both the first and second paragraphs are: “presumably (supposedly) identical materials”. If, in fact, “identical materials” were available, one of the variables of testing would be eliminated because the performance of the material would be a known, which could be used for calibration procedures. Thus, any variations in the test results would be caused by a, b, or d. The Holy Grail of ASTM fire standards is the precision and bias section. ASTM regulations require precision statements in all test methods in terms of repeatability and reproducibility. However, most E05 standards do not provide precision and bias data. In fact, only 33% of E05 standards have conducted an inter-laboratory study or round robin at some time. A path forward to promote more systematic calibration procedures for E05 fire test methods and to facilitate more frequent round robin studies, is to explore the identification or development of Standard Reference Materials (SRM's) with defined properties for calibration of ASTM E05 fire tests and to validate operational performance for the same test method conducted at different facilities. Surveys of a wide range of ASTM test standards show that some of them use well-characterized materials to provide specific output data as part of their calibration procedures. In fact, SRM's have been developed for test standards; E162, E648 and E662. This paper summarizes the properties of the materials used for test calibration and their potential use as reference standards or for identifying properties important to formulating a standard material(s) for ASTM E05 fire test methods.
Heat Flux Measurements and Their Uncertainty in a Large-Scale Fire Test
Authors: Lam Cecilia S. , Weckman Elizabeth J.
Published: January 2013

Heat flux data from a series of controlled experiments involving a 2 m diameter, wind-blown pool fire are examined to highlight the difficulties involved in conducting heat flux measurements in a realistic, large-scale, hydrocarbon-fueled fire. Data were taken at several locations along the ground near the fire. At each location, three different heat flux sensors were positioned together: a Gardon gage, a directional flame thermometer (DFT) and a Sandia heat flux gage (HFG). Methods were first developed to correct measured values of heat flux for the slight differences in gage location relative to the fire. The remaining discrepancies between the values of heat flux measured by the different gages were then used to highlight uncertainties in heat flux measurements due to differences in gage surface temperature, in gage thermal response to the inherent modes of heating involved in the large hydrocarbon fire environment, and in conduction losses from the gage sensor plates. The importance of these sources of discrepancy varied depending on the magnitude of the measured heat flux and on whether the gages were located in a radiation-dominated or mixed radiative-convective environment within the fire.