bluuebunny/crossref_metadata_2025 huggingface dataset:
Effect of Hydrogen on Irradiation Creep and Growth for ZIRLO Alloy and Zr-1.0Nb
Authors: Foster John Paul , Pan Guirong , Cai Lu , Atwood Andrew R.
Published: February 2018

The impact of hydrogen on the irradiation growth and creep of stress-relief annealed (SRA) ZIRLO® alloy and recrystallized annealed (RXA) Zr-1.0Nb cladding tubes is evaluated in this paper. The samples were charged with hydrogen in the range of approximately 160–720 ppm using the gaseous method. Biaxial in-reactor creep tests were performed after Cycle 1, Cycle 2, Cycle 3, and Cycle 4, on both as-received and precharged hydrogen cladding tubes. Outside diameter and axial length measurements were performed on the samples. The results showed that hydrogen had no effect on the axial irradiation creep but a relatively large effect on the axial irradiation growth. Increasing hydrogen decreased the axial irradiation growth in RXA Zr-1.0 Nb, which was opposite from the behavior of SRA ZIRLO cladding. This unique hydrogen effect on the irradiation axial growth of RXA Zr-1.0Nb could be due to the different intergranular stress resulting from the fabrication process or the absence of alloying elements, including tin. The total axial strain of both Zr-1.0Nb and SRA ZIRLO cladding increased with increasing fast fluence, and Zr-1.0Nb increased at a faster rate relative to SRA ZIRLO cladding. In the diameter direction, hydrogen had a minimal effect on the total diameter strain and the diameter irradiation creep strain for both SRA ZIRLO samples and the RXA Zr-1.0Nb sample. This finding from in-reactor test is contrary to the out-reactor tests results from the literature that have shown that hydrogen significantly decreases thermal creep. The total diameter strain and diameter irradiation creep behavior for the SRA ZIRLO samples and RXA Zr-1.0Nb were similar.
The Performance of NSF in BWR Operating Conditions
Authors: Cantonwine Paul E. , Lutz Dan R. , White David W. , Lin Yang-Pi
Published: February 2018

NSF is a zirconium alloy in the zirconium-tin-niobium-iron family. The nominal alloy content is 1 % tin, 1 % niobium, and 0.4 % iron. The initial targeted component for NSF in boiling water reactors (BWRs) is the fuel assembly channel. A BWR fuel channel encases the fuel bundle such that boiling is confined within the channel while the water remains solid outside the channel. The BWR fuel channel also defines the gap between the bundles where the cruciform control blade inserts and withdraws to control the criticality of the core. The key performance requirements of the BWR fuel channel are dimensional stability (including both shape and thickness) and adequate mechanical properties to withstand the applied stresses. The distortion mechanisms that affect dimensional stability are fluence gradient-induced bow, shadow corrosion-induced bow, and elastic and creep bulge. Although NSF is known to have improved irradiation growth characteristics compared with both Zircaloy-2 and Zircaloy-4 that translates to increased dimensional stability in fuel channels, the BWR corrosion performance, tensile properties, and irradiation creep performance are less well known. This paper reports on the physical metallurgy of NSF, oxide thickness and hydrogen content after in-reactor operation, mechanical properties (both irradiated and unirradiated), in-reactor creep response, and irradiation growth and bow of NSF channels.
Preliminary Irradiation Effect on Corrosion Resistance of Zirconium Alloys
Authors: Markelov Vladimir , Novikov Vladimir , Shevyakov Alexandr , Gusev Anatoly , Peregud Michael , Konkov Viktor , Eremin Sergey , Pokrovsky Alexander , Obukhov Alexander
Published: February 2018

In order to improve the composition and structure of zirconium alloys, it is important to know how changes in their structure-phase state under irradiation affect various properties, including corrosion. In this work, autoclave corrosion tests were carried out on fuel cladding samples made of zirconium-niobium and zirconium-niobium-tin-iron alloy systems. All samples were preliminarily irradiated in a BOR-60 nuclear reactor at a fluence of 2 × 1026 m−2 (E > 0.1 MeV). The autoclave tests of irradiated and unirradiated samples were performed at 350°C for 240 days in distilled water containing 10 ppm lithium and 1,600 ppm boron.

The irradiation of all alloys resulted in a significant change in their initial structure-phase state. The content of niobium in β-niobium–phase precipitates was reduced from 80 % to −90 % to 50 % to −60 %. Iron atoms from the Laves-phase particles were diffused into a matrix, and their lattice transformed from hexagonal close-packed to body-centered cubic. An average size of the transformed second-phase precipitates in the irradiated alloys increased relatively to an average size of the initial precipitates. Elongated radiation-induced precipitates (RIPs) with an average length up to 7 nm were formed in all alloys under irradiation. In comparison with the unirradiated state, irradiation reduced the corrosion of zirconium-niobium alloys, and this effect became more apparent with increased niobium content in the alloy.

For zirconium-niobium-tin-iron alloys, irradiation led to more severe corrosion, mostly because of increasing tin content. Niobium content increased up to 2.37 %, and iron content decreased up to 0.18 % in the alloys, contributing to higher corrosion resistance. An interconnection between the niobium/(tin + iron) alloying parameter, quantity of nanometric precipitates formed therein under irradiation, and their corrosion performance has been suggested.
The Effect of Iron on Dislocation Evolution in Model and Commercial Zirconium Alloys
Authors: Topping Matthew , Harte Allan , Frankel Philipp , Race Christopher , Sundell Gustav , Thuvander Mattias , Andrén Hans-Olof , Jadernas Daniel , Tejland Pia , Romero Javier E. , Darby Edward C. , Dumbill Simon , Hallstadius Lars , Preuss Michael
Published: February 2018

Although the evolution of irradiation-induced dislocation loops has been well correlated with irradiation-induced growth phenomena, the effect of alloying elements on this evolution remains elusive, especially at low fluences. To develop a more mechanistic understanding of the role iron has on loop formation, we used state-of-the-art techniques to study a proton-irradiated Zr-0.1Fe alloy and proton- and neutron-irradiated Zircaloy-2. The two alloys were irradiated with 2-MeV protons up to 7 dpa at 350°C and Zircaloy-2 up to 14.7 × 1025 n · m−2, approximately 24 dpa, in a boiling water reactor at approximately 300°C. Baseline transmission electron microscopy showed that the Zr3Fe secondary-phase particles in the binary system were larger and fewer in number than the Zr(Fe,Cr)2 and Zr2(Fe,Ni) particles in Zircaloy-2. An analysis of the irradiated binary alloy revealed only limited dissolution of Ze3Fe, suggesting little dispersion of iron into the matrix, while at the same time a higher -loop density was observed compared with Zircaloy-2 at equivalent proton dose levels. We also found that the redistribution of iron during irradiation led to the formation of iron nanoclusters. A delay in the onset of -loop nucleation in proton-irradiated Zircaloy-2 compared with the binary alloy was observed. The effect of iron redistributed from secondary-phase particles because of dissolution on the density and morphology of  and  loops is described. The implication this may have on irradiation-induced growth of zirconium fuel cladding is also discussed.
Investigating the Effect of Zirconium Oxide Microstructure on Corrosion Performance: A Comparison between Neutron, Proton, and Nonirradiated Oxides
Authors: Garner Alistair , Baxter Felicity , Frankel Philipp , Topping Matthew , Harte Allan , Slater Thomas , Tejland Pia , Romero Javier E. , Darby Edward C. , Cole-Baker Aidan , Gass Mhairi , Preuss Michael
Published: February 2018

Scanning precession electron diffraction in the transmission electron microscope has been used to simultaneously map the phase, orientation, and grain morphology of oxides formed on Zircaloy-2 after three and six cycles in a boiling water reactor in unprecedented detail. For comparison, a region of a preoxidized autoclave-formed oxide was also proton-irradiated at the Dalton Cumbrian Facility. The proton irradiation was observed to cause additional stabilization of the tetragonal phase that was attributed to the stabilizing effect of irradiation-induced defects in the oxide. In the reactor-formed oxides, no extra stabilization of the tetragonal grains was observed under neutron irradiation, as indicated by the similar tetragonal phase fraction and transformation twin-boundary distributions between the nonirradiated and reactor-formed oxides. It is suggested that the damage rate is too low in the newly formed oxide to cause significant stabilization of the tetragonal phase. This technique also reveals that the oxide formed under reactor conditions has a more heterogeneous microstructure, and the growth of well-oriented columnar monoclinic grains is significantly reduced compared with a nonirradiated oxide. High-angle annular dark-field scanning transmission electron microscopy also revealed the development of extensive networks of intergranular porosity and eventually grain decohesion in the reactor-formed oxides. These results suggest that the tetragonal-monoclinic transformation is not responsible for the accelerated corrosion exhibited under reactor conditions. It is proposed that the usual out-of-reactor oxide growth and nucleation processes are significantly modified under reactor conditions, resulting in a more heterogeneous and randomly oriented oxide microstructure with reduced columnar grain growth. It is suggested that this disordered oxide microstructure allows for the formation of extensive intergranular porosity that could lead to accelerated in-reactor corrosion.
Texture Development during Rolling of α + β Dual-Phase ZrNb Alloys
Authors: Daniel Christopher S. , Honniball Peter D. , Bradley Luke , Preuss Michael , Quinta da Fonseca João
Published: February 2018

The texture evolution during high-temperature rolling (800°C) of two model zirconium-niobium alloys was studied using electron backscatter diffraction. The aim was to decouple the relative importance of deformation and phase transformation on the final texture in an effort to elucidate the origin of the strong α-transverse basal texture commonly found in the hot-rolled product. The analysis uses a reconstruction software based on the Burgers relation to reconstruct the high-temperature β orientations, with the indexed α as input. In the low niobium content alloy, the texture evolution of the { 112̅0 } < 101̅0 > component was consistent with deformation in both α and β phases, along with a variant selection mechanism that occurred during the β → α phase transformation. Variant selection was observed to depend on the orientation of the deformed prior β grains. In the high niobium content alloy, a lower α volume fraction during hot rolling gave rise to different β textures and promoted the formation of a Goss component. This contributed to the { 112̅0 } < 101̅0 > component upon cooling, with variant selection enhanced by anisotropic grain breakup.
Understanding Corrosion and Hydrogen Pickup of Zirconium Fuel Cladding Alloys: The Role of Oxide Microstructure, Porosity, Suboxides, and Second-Phase Particles
Authors: Hu Jing , Setiadinata Brian , Aarholt Thomas , Garner Alistair , Vilalta-Clemente Arantxa , Partezana Jonna M. , Frankel Philipp , Bagot Paul , Lozano-Perez Sergio , Wilkinson Angus , Preuss Michael , Moody Michael , Grovenor Chris
Published: February 2018

We used a range of advanced microscopy techniques to study the microstructure, nanoscale chemistry, and porosity in zirconium alloys at different stages of oxidation. Samples from both autoclave and in-reactor conditions were available, including ZIRLO™, Zr-1.0Nb, and Zr-2.5Nb samples with different heat treatments. Scanning transmission electron microscopy (STEM), transmission Kikuchi diffraction (TKD), and automated crystal orientation mapping with TEM were used to study the grain structure and phase distribution. Significant differences in grain morphology were observed between samples oxidized in the autoclave and in-reactor, with shorter, less well-aligned monoclinic grains and more tetragonal grains in the neutron-irradiated samples. A combination of energy-dispersive X-ray mapping in STEM and atom probe tomography analysis of second-phase particles (SPPs) can reveal the main and minor element distributions respectively. Neutron irradiation seems to have little effect on promoting fast oxidation or dissolution of β-niobium precipitates but encourages the dissolution of iron from Laves-phase precipitates. An electron energy-loss spectroscopy (EELS) analysis of the oxidation state of niobium in β-niobium SPPs in the oxide revealed the fully oxidized Nb5+ state in SPPs deep into the oxide but Nb2+ in crystalline SPPs near the metal-oxide interface. EELS analysis and automated crystal orientation mapping with TEM revealed Widmanstatten-type suboxide layers in some samples with the hexagonal ZrO structure predicted by ab initio modeling. The combined thickness of the ZrO suboxide and oxygen-saturated layers at the metal-oxide interface correlated well to the instantaneous oxidation rate, suggesting that this oxygen-rich zone is part of the protective oxide that is rate limiting in the transport processes involved in oxidation. Porosity in the oxide had a major influence on the overall rate of oxidation, and there was more porosity in the rapidly oxidizing annealed Zr-1.0Nb alloy than in either the recrystallized alloy or the similar alloy exposed to neutron irradiation.
Alloying Effect of Niobium and Tin on the Zirconium Alloy Fuel Cladding Behavior at High Temperature Oxidation in Steam
Authors: Malgin Andrey , Markelov Vladimir , Gusev Anatoly , Nikulina Antonina , Novikov Vladimir , Shelepov Ivan , Donnikov Vladimir , Latynin Vyacheslav , Kosihina Julia
Published: February 2018

As the materials for fuel claddings for water-cooled reactors, binary and multicomponent zirconium alloys zirconium-niobium (Zr-Nb) (E110, M5), zirconium-tin-iron (Zr-Sn-Fe) (Zircaloy-2,-4), and zirconium-niobium-tin-iron (Zr-Nb-Sn-Fe) (E635, ZIRLO®) most commonly are used. Improvement extends, in particular, by varying their composition by niobium, tin, and iron. At the same time, importance is given to ensuring the safe operation of the fuel rods not only in normal conditions but also in emergency situations such as a loss of coolant accident. In this paper, we present the research results of studies of the influence of alloying elements niobium, tin, and iron on corrosion resistance and embrittlement at high-temperature steam oxidation of fuel claddings based on the alloy types Zr-xNb (x = 1.0 ÷ 2.5) and Zr-xNb-ySn-zFe (x = 0.6 ÷ 2.4; y = 0.24 ÷ 1.1; z = 0.18 ÷ 0.34) made by using zirconium sponge. We conducted high-temperature steam oxidation tests at temperatures of 1,000°C, 1,100°C, and 1,200°C with continuous measurement of the weight gain during the experiment and subsequent cooling in steam with rate ∼20°C/s. We conducted studies of kinetics of high-temperature oxidation in steam, structural-phase state changes, alloying elements distribution, absorbed hydrogen content, and residual ductility after rapid cooling of the oxidized specimens. We revealed the difference in the oxidation kinetics of the materials studied, which decreased as the oxidation temperature increased. Likewise, an increase of the tin content in the alloy influenced a specimen's breakaway oxidation intensification at 1,000°C. This shows that the mechanical properties of fuel claddings oxidized under the same conditions depend on their alloying composition, and the hydrogen fraction is absorbed by the specimen and formed by the oxidation of ZrO2, α-Zr(O), and ex-β layer structures, which in turn varies depending on the alloy composition. The increase of niobium, tin, and iron content in the zirconium alloy leads to a decrease in the residual ductility of fuel claddings after high-temperature steam oxidation. According to obtained results, among the materials studied, Zr-1.0Nb alloy is the most resistant to embrittlement during high-temperature steam oxidation at temperatures of 1,000–1,200°C.
Influence of Hydride Precipitation on the Corrosion Kinetics of Zircaloy-4: Effect of the Nanostructure and Grain Boundary Properties of the Zirconium Oxide Layer on Oxygen Diffusion Flux
Authors: Jublot Michael , Zumpicchiat Guillaume , Tupin Marc , Pascal Serge , Berdin Clotilde , Bisor Caroline , Blat-Yrieix Martine
Published: February 2018

Hydride precipitation at the oxide-metal interface is frequently proposed as causing the corrosion acceleration of Zircaloy-4 at high burnup in pressurized water reactors (PWRs). In order to identify the local mechanisms possibly involved, we studied the nanostructure of oxides formed on massive zirconium hydrides and reference Zircaloy-4 with an innovative grain mapping technique with the use of transmission electron microscopy (TEM). In autoclave PWR conditions, the presence of a precipitated hydride phase, previously formed by a cathodic charging technique at the surface of Zircaloy-4, clearly increased the corrosion rate, and a higher oxygen diffusion flux along oxide grain boundaries is observed compared with the reference Zircaloy-4. The texture, grain size, and adjacent grain misorientation in the oxide layer were studied in detail along the direction of the oxide growth using an automated crystal orientation mapping technique associated with TEM for both prehydrided and reference Zircaloy-4 samples. The texture in the growth direction is similar on prehydrided and reference samples, but the grain-to-grain misorientations showed differences. Indeed, on the prehydrided sample, the misorientation of 90° with respect to the [001] monoclinic axis is less probable than in the reference oxide, and more misorientations of 50–70° and 120–150°, corresponding to larger mismatches between neighboring grain boundaries, are observed. A smaller average diameter of the columnar monoclinic grains is also clearly revealed for the oxide grown on the prehydrided sample that leads to a larger number of diffusion paths for oxidizing species. These results are discussed and used for simulating oxygen diffusion flux through the polycrystalline microstructure of the oxide layer as a function of the grain size.
Understanding Irradiation Growth through Atomistic Simulations: Defect Diffusion and Clustering in α-Zirconium and the Influence of Alloying Elements
Authors: Christensen Mikael , Wolf Walter , Freeman Clive , Wimmer Erich , Adamson Ronald B. , Hallstadius Lars , Cantonwine Paul E. , Mader Erik V.
Published: February 2018

Irradiation-induced structural changes of α-zirconium alloys and in particular the effect of iron were investigated by molecular dynamics simulations using embedded atom potentials derived from first-principles calculations. The simulations revealed that at temperatures between 500 and 600 K self-interstitial atoms (SIAs) diffuse rapidly in a cooperative movement, preferably parallel to basal planes (a directions; ), forming nanoclusters with an extension in  and . Vacancies diffuse more slowly than SIAs and remain isolated for a longer period of time. Nanoclusters associated with SIAs cause a pronounced overall expansion in a directions, as well as local strains. Under compressive strain in the c direction, vacancy diffusivity increases in the c direction. In contrast, the diffusivity of SIAs increases in the c direction under a tensile strain in the c direction. SIA nanoclusters are highly mobile within basal planes. Vacancy clusters grow by merging, leading to a contraction in the a direction, compensating for the expansion caused by SIA nanoclusters and possibly contributing to the plateau in growth after the initial rapid expansion. At the onset of breakaway growth, possibly due to stress buildup, the vacancy nanoclusters can condense into c loops, thereby diminishing the compensation effect. The alloying elements iron, nickel, chromium, and niobium liberated from secondary phase particles under irradiation or already in solution are attracted to vacancies and SIAs and are found inside vacancy and SIA loops. The interaction of alloying elements with defect clusters is discussed, with a particular focus on iron. Iron has been found to promote cluster formation in zirconium, and the structures of zirconium-iron clusters have been analyzed. Tin is repelled by SIA clusters and only weakly attracted by vacancies. Niobium impedes the diffusion of SIAs (and therefore may increase annihilation rates with nearby vacancies) and does not destabilize vacancy or SIA clusters. Ab initio calculations of the dimensional and elastic coefficients of the intermetallic phases occurring in secondary phase particles, such as Zr2Fe and Zr3Fe, are presented, allowing an assessment of local strains in a zirconium matrix. Thus, novel results from extended molecular dynamics simulations provide new insights and contribute to a deeper understanding of the complex mechanisms causing irradiation-induced dimensional changes and the breakaway growth of zirconium alloys.
Mechanistic Understanding of Zirconium Alloy Fuel Cladding Performance
Authors: Motta Arthur T.
Published: February 2018

A review is presented of work performed in our group over the years in the areas of radiation damage, corrosion, hydrogen pickup, hydriding, and the mechanical behavior of zirconium alloy nuclear fuel cladding with the goal of developing a greater mechanistic understanding of cladding performance in service.
Dynamic J-lntegral Toughness and Fractographic Studies of Fast Reactor Irradiated Type 321 Stainless Steel
Authors: Little Edward A.
Published: February 2018

The dynamic J-lntegral toughness and associated strength changes in solution-treated Type 321 stainless steel have been characterized following fast reactor irradiation to displacement doses in the range 20 to 54 dpa (displacements per atom) at mean temperatures of ~230 and ~400°C, using instrumented impact test procedures.

Irradiation induces significant but not deleterious toughness degradation. Both the dose and test temperature dependence of the toughness changes are controlled by the irradiation temperature. Maximum toughness loss occurs at doses below 20 dpa for both irradiation temperatures; however, 400°C irradiation is associated with subsequent saturation of the toughness loss, whereas the 230°C irradiation condition exhibits further toughness degradation up to 54 dpa. The toughness response is consistent with fractographic observations that demonstrate that a predominantly ductile fracture mode is maintained after irradiation, but with significant changes in the scale of microvoid coalescence.
Fracture in Helium-Irradiated Type 316 Stainless Steel Microtensile Specimens
Authors: Gerke R. David , Jesser William A.
Published: February 2018

In an attempt to clarify the role of helium in promoting high-temperature intergranular failure in irradiated materials, in situ tension experiments have been conducted in a high-voltage electron microscope (HVEM) on 40-μm-thick foil microtensile specimensof Type 316 stainless steel irradiated with 80 keV helium ions to a fluence of about 1 × 1021 ions/m2. It is found for irradiated austenitic stainless steel that when tension testing is performed above 700°C, intergranular failure occurs; but if the tension test is performed below 550°C, transgranular failure occurs. In order to better examine the role of temperature in causing intergranular failure, in situ tension testing was initiated at 700°C, briefly halted while the temperature was lowered to just below 550°C, and resumed at the lower temperature.

Specimens tested under these dual temperature tension test conditions exhibited intergranular failure at 700°C and, after the tension test was resumed at 550°C, changed to a transgranular failure mode. This result agrees with a growing body of data that supports the idea that when helium atoms are trapped in bubbles they are benign but if the helium atoms are released from their traps and can migrate to the grain boundary and reside there as lattice helium, that is, dissolved helium, then the role of helium is to reduce ductility and promote intergranular failure.
Tensile Properties and Swelling of 20% Cold-Worked Type 316 Stainless Steel Irradiated in a High-Flux Isotope Reactor (HFIR)
Authors: Klueh R. L. , Grossbeck M. L.
Published: February 2018

Immersion density and elevated-temperature tensile properties were determined on 20% cold-worked Type 316 stainless steel irradiated in the High-Flux Isotope Reactor (HFIR) at approximately 285, 370,470, 560, and 620°C to fluences of 1.8 to 6.2 × 1026 neutrons (n)/m2 (>0.1 MeV), which resulted in displacement damage levels of 16 to 54 displacements per atom (dpa) and helium concentrations of 900 to 3300 atomic parts per million (appm). Tension tests were done at temperatures near the estimated irradiation temperatures. Swelling increased with increasing irradiation temperature to a maximum of 2.1% at 620°C. Irradiation at the lowest temperatures (285 and 370°C) increased the strength. At the higher irradiation temperatures, the strength decreased during irradiation. Ductility generally reflected the strength behavior: an increase in strength resulted in a decrease in ductility. When the present data are compared with previously published data, there is good agreement with one exception. Previous experiments showed a large decrease in ductility when irradiated at 600°C and tested at 575°C that was not observed in the present tests. There was also good agreement between HFIR-irradiated steel and literature data for Experimental Breeder Reactor-II (EBR-II)-irradiated steel.
In-Reactor Creep Rupture Experiment in the Materials Open Test Assembly (MOTA)
Authors: Puigh Raymond J. , Schenter Robert E.
Published: February 2018

The capabilities of the Materials Open Test Assembly (MOTA) irradiation vehicle and the Fast Flux Test Facility (FFTF) cover gas monitoring system are described in relation to the creep rupture experiment during FFTF Cycle 2 operation. The creep rupture specimen matrix in MOTA includes both U. S. and foreign alloys at irradiation temperatures of 575, 605, 670, and 750°C. For exposure times less than 2400 h and fluences less than 3.2 × 1022 neutrons (n)/cm2 (E > 0.1 MeV), neutron irradiation does not impact the rupture times for the U. S. alloys irradiated in MOTA.
Results of the International Atomic Energy Agency (IAEA) Coordinated Research Programs on Irradiation Effects on Advanced Pressure Vessel Steels
Authors: Steele Lendell E. , Davies L. Myrddin , Ingham Terence , Brumovsky Milan
Published: February 2018

Final results of a multinational program to study the possible levels of embrittlement of steels of recent production are reviewed with clear evidence of high radiation resistance when specifications are carefully controlled. Steels from France, Germany, and Japan were studied in selected laboratories of eight nations. Besides comparative results on these steels as a function of the level of neutron exposure, results were obtained on effects of different temperatures, on fracture mechanics test results, and on mechanisms of radiation embrittlement. A third phase of the effort, which was approved in December 1983, is aimed at enhancing surveillance programs and is described.
Charpy Toughness and Tensile Properties of a Neutron-Irradiated Stainless Steel Submerged-Arc Weld Cladding Overlay
Authors: Corwin William R. , Berggren Reynold G. , Nanstad Randy K.
Published: February 2018

The possibility of stainless steel cladding increasing the resistance of an operating nuclear reactor pressure vessel to extension of surface flaws is highly dependent upon the irradiated properties of the cladding. Therefore, weld overlay cladding irradiated at temperatures and fluences relevant to power reactor operation was examined. The cladding was applied to a pressure vessel steel plate by the submerged-arc, single-wire, oscillating electrode method. Three layers of cladding were applied to provide a cladding thickness adequate for fabrication of test specimens. The first layer was Type 309, and the upper two layers were Type 308 stainless steel. There was considerable dilution of the Type 309 in the first layer of cladding as a result of excessive melting of the base plate. Specimens for the irradiation study were taken from near the base plate/cladding interface and also from the upper layers of cladding. Charpy V-notch and tension specimens were irradiated at 288°C to neutron fluences of 2 × 1023 neutrons (n)/m2 (E > 1 MeV).

When irradiated, both Types 308 and 309 cladding showed a 5 to 40% increase in yield strength accompanied by a slight increase in ductility in the temperature range from 25 to 288°C. All cladding exhibited ductile-to-brittle transition behavior during impact testing. The Type 308 cladding, microstructurally typical of that in reactor pressure vessels, showed very little degradation in either upper-shelf energy or transition temperature due to irradiation. Conversely, the impact properties of the specimens containing the highly diluted Type 309 cladding, microstructurally similar to that produced during some off-normal welding conditions in existing reactors, experienced significant increases in transition temperature and drops of up to 50% in upper-shelf energy. The impact energies of the Charpy specimens containing the Type 309 layer of cladding strongly reflected the amount of the Type 309 actually comprising the specimen, falling into two distinct high- and low-energy populations with the low-energy population corresponding to a higher fraction of Type 309 in the specimen.
Post-Irradiation Annealing Recovery of Commercial Pressure Vessel Steels
Authors: Macdonald Bruce
Published: February 2018

A correlation was found to describe the residual shift in the 41-J Charpy transition temperature after post-irradiation annealing of commercial pressure vessel steels and weld deposits. The resulting trend equation has a 11°C standard deviation in residual shift.
Chemical Composition of Nuclear Reactor Vessel Welds
Authors: Moore Kenneth E. , Heller A. S. , Lowe Arthur L.
Published: February 2018

The chemical compositions of Babcock & Wilcox Nuclear 177 fuel assembly (B&W 177 FA) reactor vessel beltline welds were characterized. Emphasis was placed on copper concentration and variability due to the effect of copper on the fracture and impact toughness properties of irradiated weld metals. Fourteen full section and surveillance program welds were examined. These welds (unique filler wire heats and flux lots) were made with 8 of the 13 heats of filler wire and a majority of the flux lots used in the B&W 177 FA reactor vessel beltline welds. The concentrations of 21 elements were measured using emission spectrographic analysis. A total of 290 through-weld-thickness analyses were performed on these welds that provided a measure of copper concentration and variation. The mean copper concentration of the 14 welds ranged from 0.18 to 0.42% by weight with standard deviation ranging from 0.005 to 0.07% by weight. The mean copper concentration of the 290 analyses was 0.29% by weight. Variability of the other 20 elements was found to be small within each weld.

Electron microprobe analysis was used to assess the micro-variations in copper concentration in a typical reactor vessel weldment. The measured degree of segregation was low.

Additional analyses were performed to determine the quantity of copper available on the weld filler wire that was transferred to the weld deposit during the weld process. The results of these analyses indicate copper concentration equivalence between the filler wire and the weld deposit.
Relationship of Yield Strength and Upper-Shelf Impact Energy Changes for Irradiated Reactor Vessel Steel
Authors: Byrne Stephen T. , Pirek Ronald C.
Published: February 2018

Radiation-induced changes on the strength and toughness properties of reactor pressure vessel steels are known to be dependent on the chemical content and, to a lesser extent, on the product form. The precise relationship between key variables, however, is yet to be established. Emphasis is currently being placed on the development of mechanistic models to provide coherence between the variety of empirical observations presented to date.

Results are presented of an investigation of radiation damage trends exhibited by changes in the uniaxial tension and Charpy upper-shelf energy properties. Regression analysis techniques were used to determine trends based on chemistry and neutron fluence. These trends were evaluated in terms of proposed mechanistic models toward developing a broader understanding of radiation damage mechanisms.

Radiation-induced transition temperature shift in reactor vessel steels is known to be dependent on the copper and nickel content of the material. In this investigation, Charpy upper-shelf impact energy and ultimate tensile strength changes were found to exhibit similar dependencies on the copper content and neutron irradiation exposure, but neither property change was significantly influenced by nickel content. Based on the apparent difference in radiation behavior trends between transition temperature shift and shelf energy/tensile property changes, a multiple defect radiation damage model was postulated. The first defect type, dependent on copper content, increased the yield strength through a precipitation hardening mechanism. A second defect type, dependent on both copper and nickel content, affected the microcleavage fracture stress through a process such as grain boundary embrittlement. A third defect type, dependent on neutron irradiation only, affected the yield strength by a process such as vacancy hardening. This postulated model provides a reasonable basis for explaining observed differences in radiation behavior.
Analysis of Charpy V-Notch Impact Toughness of Irradiated A533-B Class 1 Plate and Four Submerged-Arc Welds
Authors: Berggren Reynold G. , Hawthorne J. Russell , Nanstad Randy K.
Published: February 2018

Studies of the effects of neutron irradiation on Charpy V-notch impact properties of steels have, in general, included a minimum number of tests for each combination of material and irradiation parameters. The present study attempts to apply statistical analyses with multiple testing at selected temperatures to assess the accuracy and reliability of results.

Charpy V-notch impact test specimens were irradiated in the Bulk Shielding Reactor at Oak Ridge National Laboratory at 288°C to target neutron fluences of 2 × 1023 neutrons (n)/m2 (>1 MeV). The materials were ASTM A533-B Class 1 plate (Heavy-Section Steel Technology Plate 02) and four submerged-arc welds representing current nuclear pressure vessel fabrication practice. Both unirradiated and irradiated specimens were tested by two separate groups and multiple tests were conducted at several selected temperatures. Statistical analyses permitted determination of material and test variability and an interlaboratory comparison.

Several Charpy curve-fitting methods were used and results are compared with predictions from several “trend curve” expressions.
Effects of Neutron Irradiation on Fracture Toughness of A533-B Class 1 Plate and Four Submerged-Arc Welds
Authors: Menke Blaine H. , McGowan John J. , Berggren Reynold G. , Nanstad Randy K. , Miller Kevin C.
Published: February 2018

An irradiation program was carried out with four submerged-arc welds and one plate of A533-B Class 1 pressure vessel steel. The welds were made by commercial vendors using current welding practice and contained relatively low copper levels. The target fluence was 2 × 1023 neutrons (n)/m2 (E > 1 MeV). The program was administered by Oak Ridge National Laboratory (ORNL) through the Heavy Section Steel Technology Program Office, and the testing was shared by ORNL and Materials Engineering Associates, Inc. (MEA). The results of the test program indicated that the welds had generally very good properties (tension, Charpy V-notch, and fracture toughness) prior to irradiation and with the exception of tearing modulus all five materials were relatively insensitive to irradiation on the upper shelf. The data further indicated that shifts of the transition region to higher temperature with irradiation is directly related to the copper and nickel content.
Measured Radiation Environment at the Clinton P. Anderson Los Alamos Meson Physics Facility (LAMPF) Irradiation Facility
Authors: Davidson Dorothy R. , Greenwood Lawrence R. , Reedy Robert C. , Sommer Walter F.
Published: February 2018

A foil activation dosimetry experiment was conducted at the Clinton P. Anderson Los Alamos Meson Physics Facility (LAMPF) beam stop area to characterize the radiation environment and to identify techniques for spallation radiation measurements. The measured total spallation neutron flux outside the primary proton beam was 5.5 × 1017 neutrons/m2/s/mA ±7%. A previous Monte Carlo calculation gave a total spallation neutron flux in approximately the same location as the dosimetry package of 6.0 × 1017 neutrons/m2/s/mA ±3%. Comparison between the measured and calculated neutron fluxes shows good agreement. To extend the characterization, additional spallation cross sections for energies up to 800 MeV and the use of foils sensitive to the epithermal energy range are needed.
A Modified Tritium Trick Technique for Doping Vanadium Alloys with Helium
Authors: Braski David N. , Ramey Dan W.
Published: February 2018

A modified tritium trick technique was used to implant three different levels of 3He in V-15Cr-5Ti (% by weight) and Vanstar-7 specimens before irradiation in the Fast Flux Test Facility (FFTF). The modifications include: (1) wrapping of the specimens with tantalum foil to minimize oxygen contamination, and (2) a 400°C decay-time treatment to prevent vanadium tritide formation and to produce a 3He bubble distribution similar to that produced during elevated temperature irradiation. Preliminary results show that both modifications were successful. However, the tritium removal step at 700°C, especially at higher helium levels, allowed large 3He bubbles to form in the grain boundaries and severely embrittled the V-15Cr-5Ti alloy. A better microstructure for simulation purposes can probably be produced by lowering the tritium removal temperature. Vanstar-7 specimens consistently absorbed about half as much tritium, and subsequently contained half as much 3He as V-15Cr-5Ti. Implanting 3He in vanadium alloys via the tritium trick offers a convenient technique to study the mechanism of helium embrittlement without irradiation and should provide a rapid screening method to help develop embrittlement-resistant vanadium alloys.
A Transmission Electron Microscopy Study of the Structure and Annealing Behavior of Copper-Krypton Deposits
Authors: Evans John H. , Williamson Robert , Whitmell Derek S.
Published: February 2018

A combined implantation and sputtering method has been developed at Harwell to entrap high (5%) atomic concentrations of krypton in a metal matrix. The work has been aimed at providing a safe and effective method for storing the active krypton-85 resulting from fuel reprocessing. In the present paper, the results of a transmission electron microscopy investigation into the structures of copper-krypton deposits will be presented together with the effects of annealing behavior. The overall impression of the as-deposited material is one of a very fine grain size, high dislocation density, and an extremely large concentration of small bubbles in the range 1.5 to 2.0 nm diameter. Taking into account additional results of macroscopic density and lattice parameter measurements, it is argued that the bubbles contain krypton at very high pressures; nevertheless, an appreciable fraction of krypton must be present in bubbles that are submicroscopic in size.

On annealing, large changes in bubble substructures became apparent at temperatures greater than half the melting point when the main effect was one of bubble coarsening as a result of coalescence and grain boundary bubble growth. These effects were reflected in the specimen surface structure as observed by scanning electron microscopy. The observations of the as-deposited material and its evolution with temperature should provide a useful background to the understanding of the material stability and high temperature gas release behavior.
Neural Determinants of Task Performance during Feature-Based Attention in Human Cortex
Authors: Jigo Michael , Gong Mengyuan , Liu Taosheng
Published: February 2018

Studies of feature-based attention have associated activity in a dorsal frontoparietal network with putative attentional priority signals. Yet, how this neural activity mediates attentional selection and whether it guides behavior are fundamental questions that require investigation. We reasoned that endogenous fluctuations in the quality of attentional priority should influence task performance. Human subjects detected a speed increment while viewing clockwise (CW) or counterclockwise (CCW) motion (baseline task) or while attending to either direction amid distracters (attention task). In an fMRI experiment, direction-specific neural pattern similarity between the baseline task and the attention task revealed a higher level of similarity for correct than incorrect trials in frontoparietal regions. Using transcranial magnetic stimulation (TMS), we disrupted posterior parietal cortex (PPC) and found a selective deficit in the attention task, but not in the baseline task, demonstrating the necessity of this cortical area during feature-based attention. These results reveal that frontoparietal areas maintain attentional priority that facilitates successful behavioral selection.
Rescue of Learning and Memory Deficits in the Human Nonsyndromic Intellectual Disability Cereblon Knock-Out Mouse Model by Targeting the AMP-Activated Protein Kinase–mTORC1 Translational Pathway
Authors: Bavley Charlotte C. , Rice Richard C. , Fischer Delaney K. , Fakira Amanda K. , Byrne Maureen , Kosovsky Maria , Rizzo Bryant K. , Del Prete Dolores , Alaedini Armin , Morón Jose A. , Higgins Joseph J. , D'Adamio Luciano , Rajadhyaksha Anjali M.
Published: February 2018

A homozygous nonsense mutation in the cereblon (CRBN) gene results in autosomal recessive, nonsyndromic intellectual disability that is devoid of other phenotypic features, suggesting a critical role of CRBN in mediating learning and memory. In this study, we demonstrate that adult maleCrbnknock-out (CrbnKO) mice exhibit deficits in hippocampal-dependent learning and memory tasks that are recapitulated by focal knock-out of Crbn in the adult dorsal hippocampus, with no changes in social or repetitive behavior. Cellular studies identify deficits in long-term potentiation at Schaffer collateral CA1 synapses. We further show thatCrbnis robustly expressed in the mouse hippocampus andCrbnKOmice exhibit hyperphosphorylated levels of AMPKα (Thr172). Examination of processes downstream of AMP-activated protein kinase (AMPK) finds thatCrbnKOmice have a selective impairment in mediators of the mTORC1 translation initiation pathway in parallel with lower protein levels of postsynaptic density glutamatergic proteins and higher levels of excitatory presynaptic markers in the hippocampus with no change in markers of the unfolded protein response or autophagy pathways. Acute pharmacological inhibition of AMPK activity in adultCrbnKOmice rescues learning and memory deficits and normalizes hippocampal mTORC1 activity and postsynaptic glutamatergic proteins without altering excitatory presynaptic markers. Thus, this study identifies that loss ofCrbnresults in learning, memory, and synaptic defects as a consequence of exaggerated AMPK activity, inhibition of mTORC1 signaling, and decreased glutamatergic synaptic proteins. Thus,CrbnKOmice serve as an ideal model of intellectual disability to further explore molecular mechanisms of learning and memory.

SIGNIFICANCE STATEMENTIntellectual disability (ID) is one of the most common neurodevelopmental disorders. The cereblon (CRBN) gene has been linked to autosomal recessive, nonsyndromic ID, characterized by an intelligence quotient between 50 and 70 but devoid of other phenotypic features, making cereblon an ideal protein for the study of the fundamental aspects of learning and memory. Here, using the cereblon knock-out mouse model, we show that cereblon deficiency disrupts learning, memory, and synaptic function via AMP-activated protein kinase hyperactivity, downregulation of mTORC1, and dysregulation of excitatory synapses, with no changes in social or repetitive behaviors, consistent with findings in the human population. This establishes the cereblon knock-out mouse as a model of pure ID without the confounding behavioral phenotypes associated with other current models of ID.
The Effect of Locomotion on Early Visual Contrast Processing in Humans
Authors: Benjamin Alex V. , Wailes-Newson Kirstie , Ma-Wyatt Anna , Baker Daniel H. , Wade Alex R.
Published: February 2018

Most of our knowledge about vision comes from experiments in which stimuli are presented to immobile human subjects or animals. In the case of human subjects, movement during psychophysical, electrophysiological, or neuroimaging experiments is considered to be a source of noise to be eliminated. Animals used in visual neuroscience experiments are typically restrained and, in many cases, anesthetized. In reality, however, vision is often used to guide the motion of awake, ambulating organisms. Recent work in mice has shown that locomotion elevates visual neuronal response amplitudes (Niell and Stryker, 2010; Erisken et al., 2014; Fu et al., 2014; Lee et al., 2014; Mineault et al., 2016) and reduces long-range gain control (Ayaz et al., 2013). Here, we used both psychophysics and steady-state electrophysiology to investigate whether similar effects of locomotion on early visual processing can be measured in humans. Our psychophysical results show that brisk walking has little effect on subjects' ability to detect briefly presented contrast changes and that co-oriented flankers are, if anything, more effective masks when subjects are walking. Our electrophysiological data were consistent with the psychophysics indicating no increase in stimulus-driven neuronal responses while walking and no reduction in surround suppression. In summary, we have found evidence that early contrast processing is altered by locomotion in humans but in a manner that differs from that reported in mice. The effects of locomotion on very low-level visual processing may differ on a species-by-species basis and may reflect important differences in the levels of arousal associated with locomotion.

SIGNIFICANCE STATEMENTMice are the current model of choice for studying low-level visual processing. Recent studies have shown that mouse visual cortex is modulated by behavioral state: primary visual cortex neurons in locomoting mice tend to be more sensitive and less influenced by long-range gain control. Here, we tested these effects in humans by measuring psychophysical detection thresholds and electroencephalography (EEG) responses while subjects walked on a treadmill. We found no evidence of increased contrast sensitivity or reduced surround suppression in walking humans. Our data show that fundamental measurements of early visual processing differ between humans and mice and this has important implications for recent work on the links among arousal, behavior, and vision in these two species.