Analysis of EBSD image quality related to microstructure evolution in zirconium–niobium cladding to quantify the degree of recrystallization J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-15 Tae-Sik Jung, Hoon Jang, Yong-Kyoon Mok, Jong-Sung Yoo
In industries, the properties of metallic materials are enhanced through improvement of their microstructure via heat treatment after processing, and the effect of heat treatments on the microstructure can usually be quantified on the basis of the degree of recrystallization. Zirconium alloy claddings used in nuclear power plants, when subjected to final annealing after cold-pilgering, exhibit a recrystallization degree that renders them suitable for in-pile applications. Because the degree of recrystallization can influence fuel performance, its accurate and simple measurement is important from the perspective of producing nuclear fuel cladding. Recently, methods for calculating the recrystallization degree using information obtained from electron backscatter diffraction (EBSD) measurements have been reported, including a calculation method based on the distribution of the image quality values. However, no studies have been reported on determining the recrystallization degree of zirconium using image quality. To apply this method to zirconium, we first analyzed the correlation between microstructural changes and image quality of zirconium cladding. Specifically, we found that the image quality value was increased by the recrystallization and that the mode, which is the most frequent image quality value, was observed at a higher position with increasing heat treatment temperature. In addition, larger recrystallized grains led to higher kurtosis, which is a measure of the degree of sharpness of the image quality distribution. In other words, the appearance of newly formed microstructures was observed as a change in the image quality frequency distribution, and its shape was analyzed using various results obtained with the aid of EBSD software (OIM) and previously reported details of the recrystallization behavior of zirconium. On the basis of this information, a new calculation formula suitable for recrystallization calculations for a zirconium–niobium cladding tube was derived and subsequently compared with the calculation formula presented in a previous study.
Effects of the short-range repulsive potential on cascade damage in iron J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-15 J. Byggmästar, F. Granberg, K. Nordlund
Recent work has shown that the repulsive part of the interatomic potential at intermediate atomic separations strongly affects the extent and morphology of the damage produced by collision cascades in molecular dynamics simulations. Here, we modify an existing embedded atom method interatomic potential for iron to more accurately reproduce the threshold displacement energy surface as well as the many-body repulsion at intermediate and short interatomic distances. Using the modified potential, we explore the effects of an improved repulsive potential on the primary damage production and the cumulative damage accumulation in iron. We find that the extent of the damage produced by single cascades, in terms of surviving Frenkel pairs, directly correlates with the change in threshold displacement energies. On the other hand, the damage evolution at higher doses is more dependent on the formation and stability of different defect clusters, defined by the near-equilibrium part of the interatomic potential.
Deuterium permeation behavior and its iron-ion irradiation effect in yttrium oxide coating deposited by magnetron sputtering J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-15 Takumi Chikada, Hikari Fujita, Jan Engels, Anne Houben, Jumpei Mochizuki, Seira Horikoshi, Moeki Matsunaga, Masayuki Tokitani, Yoshimitsu Hishinuma, Sosuke Kondo, Kiyohiro Yabuuchi, Thomas Schwarz-Selinger, Takayuki Terai, Yasuhisa Oya
Tritium permeation through structural materials is a critical issue in fusion reactors from the viewpoints of sufficient fuel balance and radiological hazard. Ceramic coatings have been investigated as tritium permeation barrier for several decades; however, irradiation effects of the coatings on permeation are not elucidated. In this work, yttrium oxide coatings were fabricated on reduced activation ferritic/martensitic steels by radio frequency magnetron sputtering, and their microstructures and deuterium permeation behaviors were investigated before and after iron-ion irradiation at different temperatures. An as-deposited coating had a columnar structure and transformed into a granular one after annealing. An amorphous layer formed near the coating-substrate interface of irradiated coatings, and its thickness became thinner with increasing irradiation temperature. Voids of approximately 20 nm in diameter also formed in the irradiated coatings. Deuterium permeation flux of the sample irradiated to 1 dpa at room temperature was the lowest among the unirradiated and irradiated samples, and a permeation reduction factor indicated up to 390. The amorphous layer disappeared after deuterium permeation measurements due to damage recovery, while the voids remained and aggregated. The irradiation damage would accelerate nucleation of the crystal, resulting in a decrease of the permeation flux.
Lithium-lead corrosion behavior of erbium oxide, yttrium oxide and zirconium oxide coatings fabricated by metal organic decomposition J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-15 Moeki Matsunaga, Seira Horikoshi, Jumpei Mochizuki, Hikari Fujita, Yoshimitsu Hishinuma, Kanetsugu Isobe, Takumi Hayashi, Takayuki Terai, Yasuhisa Oya, Takumi Chikada
Tritium permeation through and corrosion of structural materials are critical issues in fusion reactor liquid lithium-lead blanket concepts from the viewpoints of an efficient fuel cycle, higher operation rates, and radiological safety. In this study, lithium-lead compatibility of ceramic coatings has been investigated for the development of tritium permeation barriers with corrosion protection. Erbium oxide, yttrium oxide and zirconium oxide coatings were fabricated by a metal organic decomposition method on reduced activation ferritic/martensitic steel F82H substrates. Corrosion and delamination of the coatings were accelerated under a higher oxygen concentration. Under a lower oxygen concentration, zirconium oxide coatings had the best lithium-lead compatibility among three coating materials from surface and cross-sectional observations. However, the zirconium oxide coating after lithium-lead immersion at 550 °C for 500 h showed higher deuterium permeability in comparison to the sample without immersion. Formation of a chromium oxide layer on the surface of the substrate before fabricating the coatings drastically improved the lithium-lead compatibility of erbium oxide and yttrium oxide coatings. Degradation of the coatings was mainly caused by corrosion and delamination depending on immersion temperature, test duration, and impurity concentration.
Modeling of hydrogen behavior in spent fuel claddings during dry storage J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-15 M. Kolesnik, T. Aliev, V. Likhanskii
A new hydride reorientation model is presented for spent fuel dry storage conditions. The model takes into account the coupled influence of external stresses, alloy texture, temperature cycling and cooling rate. The model has been tested successfully against experiments in a wide range of conditions. The features of the model are the use of kinetic equations and the simulating of such processes as hydride memory effect and possible hydrogen redistribution between hydrides of different orientations. Current version of the model will be used for safety justification of dry storage conditions.
Atomistic modeling of α’ precipitation in Fe-Cr alloys under charged particles and neutron irradiations: Effects of ballistic mixing and sink densities J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-15 Frédéric Soisson, Estelle Meslin, Olivier Tissot
Implementation of a multilayer model for measurement of thermal conductivity in ion beam irradiated samples using a modulated thermoreflectance approach J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-15 M. Faisal Riyad, Vinay Chauhan, Marat Khafizov
Laser based modulated thermoreflectance (MTR) technique is an attractive method for measuring thermal conductivity in ion irradiated samples. Unlike laser flash analysis, traditionally used for measuring thermal conductivity in nuclear materials, the MTR method is sensitive to damage that is only a few micrometers thick. This allows MTR to detect damage resulting from a few MeV ion beam irradiation. MTR combined with tailored ion beam irradiations offers a promising opportunity for validation of lower-scale thermal conductivity models developed for irradiated materials. In this technique, a harmonically modulated laser pump heats the irradiated sample and a probe beam measures the temperature induced changes in the sample's reflectivity. Interpretation of the measured temperature profiles in ion irradiated samples is complicated by the nonhomogeneous damage profile and the presence of an additional thin metal transducer layer. In this work, we present a detailed analysis of the experimentally measured thermal wave profiles in UO2 samples irradiated with 2.6 MeV H+ ions using different multilayer approximations of the damaged region. The limitation of an infinite damage layer approximation that assumes uniform damage across the thickness of the sample and neglects the undamaged region is discussed. Presented analysis outlines the method for determination of the applicability range for the infinite damage model. Finally, an analysis of the impact of point defects on thermal transport in irradiated samples is presented as an example for implementation of the MTR approach for validation of thermal transport models.
Creep tests on notched specimens of copper J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-13 Fangfei Sui, Rolf Sandström, Rui Wu
In Sweden, spent nuclear fuel is planned to be disposed off by placing it in canisters which are made of oxygen free copper alloyed with 50 ppm phosphorus. The canisters are expected to stay intact for thousands of years. During the long term disposal, the canisters will be exposed to mechanical pressure from the surroundings at temperatures up to 100 °C and this will result in creep. To investigate the role of the complex stress conditions on the canisters, creep tests under multiaxial stress state are needed. In the present work, creep tests under multiaxial stress state with three different notch profiles (acuity 0.5, 2, and 5, respectively) at 75 °C with net section stresses ranging from 170 MPa to 245 MPa have been performed. To interpret the experimental results, finite element computations have been conducted. With the help of the reference stress, the rupture lifetime in the multiaxial tests was estimated. The prediction was more precise for the higher acuities than for the lower one. In order to predict the creep deformation of the canisters for the long service period, fundamental creep models are considered. Previously developed basic models are used to compute the creep deformation in the multiaxial tests. Although the scatter is large, the agreement with the experiments is considered as acceptable, indicating that the basic models which have been successfully developed for uniaxial creep tests can also be used to describe multiaxial creep tests. Notch strengthening was observed for copper.
Multiscale modeling of irradiation hardening: Application to important nuclear materials J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-13 Ghiath Monnet
The recent progress in investigation techniques and the accumulation of simulation results across the scales have allowed for the development of a multiscale modeling framework for the physical interpretation, analysis and assessment of the flow stress in industrial materials. Although this framework is not completely achieved yet, it has reached a maturity level enabling the investigation of some complex materials properties such as irradiation hardening. In this paper, simulation and experimental results are reviewed in order to derive the constitutive equations of the flow stress and the specific contributions of the relevant microstructure components, such as solid solution, dislocation network, precipitates, dislocation loops and voids. This approach is challenged in the case of irradiated reactor pressure vessel steels, Fe-Cr alloys and austenitic stainless steels. Predictions are found to reproduce the well-known experimental trends reported in the literature.
Hot crack susceptibility in multi-pass welds of reduced activation ferritic/martensitic steel F82H J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-12 H. Mori, H. Tanimura, R. Kiyoku, M. Fujiwara, T. Kato, T. Hirose, H. Tanigawa
The research and development of nuclear fusion reactors are going on conducting to acquire a next-generation energy source. It is supposed to be that the reduced activation ferritic/martensitic (RAFM) steel F82H is adopted as a structural material of the blanket module, which is the device set on the inner wall of fusion reactors. As welding for thicker plates of F82H, multi-pass welding will be adopted to the joints. In the case of multi-pass welding, hot cracking is one of serious defects in welds and is concerned in welds reheated by following weld passes. Therefore, the objectives of this study are to evaluate hot crack susceptibility in multi-pass welds of F82H and to clarify the cause of hot cracking in multi-pass welds of F82H. The hot crack susceptibility in multi-pass welds is evaluated by the longitudinal Varestraint test with double-bead and triple-bead types. From the observation of fractured surface occurred in welds of F82H after the Varestraitnt test, the crack is identified as ductility-dip crack. The cause of hot crack in multi-pass welds of F82H is clarified by Vickers hardness test, SEM microstructure observations and X-ray diffraction pattern analyses. Based on these tests, when multi-pass welding is conducted for the F82H steels with high Ta contents, it should be considered to control welding conditions for prevention of hot cracking in weld metal as well as high strength martensitic steels. These results suggest that the cause of the crack is intragranular hardening by the precipitation of TaN during welding thermal cycles.
Effects of iron content in Ni Cr Fe alloys on the oxide films formed in an oxygenated simulated PWR water environment J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-12 Xiangkun Ru, Jiarong Ma, Zhanpeng Lu, Junjie Chen, Guangdong Han, Jinlong Zhang, Pengfei Hu, Xue Liang, Weibao Tang
The iron content in Ni Cr Fe alloys strongly affected the properties of oxide films after 1012 h of immersion in an oxygenated simulated PWR primary water environment at 310 °C. Increasing the iron content in the alloy increased the porosity and thickness of the outer layer and the inner layer. Increasing the iron content in the alloy increased the amount of spinel needle-like oxides and decreased the amount of NiO particles in the outer layer. Cellular oxide in the inner layer was nickel-rich in the center and chromium-rich at the edge.
Molecular dynamics simulations of high-energy displacement cascades in hcp-Zr J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-07 Wei Zhou, Jiting Tian, Qijie Feng, Jian Zheng, Xiankun Liu, Jianming Xue, Dazhi Qian, Shuming Peng
High-energy (up to 80 keV) displacement cascades in hcp-Zr are studied by classical Molecular Dynamics (MD) simulations. The statistics of defect production are reported and the formation of subcascades and defect clusters are analyzed. We find that the probability of subcascade formation increases along with the incident energy, and already reaches 100% at 80 keV. Our simulations also reveal that high-energy cascades could significantly promote the formation of defect clusters, sometimes even directly create experimental-scale (around 3 nm) vacancy clusters. Our research provides basic knowledge of high-energy displacement cascades in hcp-Zr, and offers a possible explanation for the low-dose irradiation experiments of Zr-based alloys.
Microstructural characterization of as-fabricated and irradiated U-Mo fuel using SEM/EBSD J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-07 Daniel Jadernas, Jian Gan, Dennis Keiser, James Madden, Mukesh Bachhav, Jan-Fong Jue, Adam Robinson
Durability of hot uniaxially pressed Synroc derivative wasteform for EURO-GANEX wastes J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-06 Yun-Hao Hsieh, Samuel A. Humphry-Baker, Denis Horlait, Daniel J. Gregg, Eric R. Vance, William E. Lee
A new titanate wasteform, Synroc-Z, was developed to contain minimal host phases for actinides and thus better suit the EURO-GANEX process. The processing conditions, waste loading and surface roughness were varied, and their effects on wasteform durability and microstructure were examined. Hot uniaxial pressing temperature was the most important factor in controlling density (<0.5 vol% porosity) and phase composition. Synroc-Z was found to have similar aqueous durability to Synroc-C. Leached samples formed Ti-oxide films and crystals on their surfaces as found by other researchers.At low waste loadings (<20 wt%), Synroc-Z showed slightly poorer durability performance than Synroc-C owing to its greater perovskite content (30 vs. 20 wt% respectively). However, at 35 wt% waste loading, Synroc-Z maintained its durability performance. This result is explained by the higher volume fraction of buffer phase, rutile, which allows greater flexibility for waste loading.
Effect of radial hydride fraction on fracture toughness of CWSR Zr-2.5%Nb pressure tube material between ambient and 300 °C temperatures J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-04 Rishi K. Sharma, A.K. Bind, G. Avinash, R.N. Singh, Asim Tewari, B.P. Kashyap
High-energy x-ray diffraction microscopy study of deformation microstructures in neutron-irradiated polycrystalline Fe-9%Cr J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-04 Xuan Zhang, Meimei Li, Jun-Sang Park, Peter Kenesei, Hemant Sharma, Jonathan Almer
Expanding the capability of reaction-diffusion codes using pseudo traps and temperature partitioning: Applied to hydrogen uptake and release from tungsten J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-04 M.J. Simmonds, J.H. Yu, Y.Q. Wang, M.J. Baldwin, R.P. Doerner, G.R. Tynan
Simulating the implantation and thermal desorption evolution in a reaction-diffusion model requires solving a set of coupled differential equations that describe the trapping and release of atomic species in Plasma Facing Materials (PFMs). These fundamental equations are well outlined by the Tritium Migration Analysis Program (TMAP) which can model systems with no more than three active traps per atomic species. To overcome this limitation, we have developed a Pseudo Trap and Temperature Partition (PTTP) scheme allowing us to lump multiple inactive traps into one pseudo trap, simplifying the system of equations to be solved. For all temperatures, we show the trapping of atoms from solute is exactly accounted for when using a pseudo trap. However, a single effective pseudo trap energy can not well replicate the release from multiple traps, each with its own detrapping energy. However, atoms held in a high energy trap will remain trapped at relatively low temperatures, and thus there is a temperature range in which release from high energy traps is effectively inactive. By partitioning the temperature range into segments, a pseudo trap can be defined for each segment to account for multiple high energy traps that are actively trapping but are effectively not releasing atoms. With increasing temperature, as in controlled thermal desorption, the lowest energy trap is nearly emptied and can be removed from the set of coupled equations, while the next higher energy trap becomes an actively releasing trap. Each segment is thus calculated sequentially, with the last time step of a given segment solution being used as an initial input for the next segment as only the pseudo and actively releasing traps are modeled. This PTTP scheme is then applied to experimental thermal desorption data for tungsten (W) samples damaged with heavy ions, which display six distinct release peaks during thermal desorption. Without modifying the TMAP7 source code the PTTP scheme is shown to successfully model the D retention in all six traps. We demonstrate the full reconstruction from the plasma implantation phase through the controlled thermal desorption phase with detrapping energies near 0.9, 1.1, 1.4, 1.7, 1.9 and 2.1 eV for a W sample damaged at room temperature.
Microstructure and phase evolution of Li4TiO4 ceramics pebbles prepared from a nanostructured precursor powder synthesized by hydrothermal method J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-02 Ruichong Chen, Qiwu Shi, Mao Yang, Yanli Shi, Hailiang Wang, Chen Dang, Jianqi Qi, Zhijun Liao, Tiecheng Lu
Li4TiO4 ceramics pebbles have been proposed to be a promising tritium breeder materials for the fusion reactor blanket of the International Thermonuclear Experimental Reactor (ITER). In the present study, the hydrothermal method was first used to synthesize the nanostructured precursor powders for fabricating the Li4TiO4 ceramic pebbles. The precursor powders were composed of LiOH, Li2CO3 and Li2TiO3 with a grain size of 30–80 nm. Moreover, the microstructure and phase evolution of Li4TiO4 ceramics pebbles were investigated. The results indicated that the formation temperature of Li4TiO4 by hydrothermal method was lower than that of other methods. The pure phase Li4TiO4 ceramic pebbles with small grain size (average value 0.50 μm), satisfactory crush load (average value 41 N) and reasonable pore structure was sintered at 750 °C for 4 h in vacuum. This study provides a new method for preparing a high quality Li4TiO4 ceramic pebble, which has potential application as a tritium breeding material.
Results on the use of tungsten heavy alloys in the divertor of ASDEX Upgrade J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-02 R. Neu, H. Maier, M. Balden, R.Dux, S. Elgeti, H. Gietl, H. Greuner, A. Herrmann, T. Höschen, M. Li, V. Rohde, D. Ruprecht, B. Sieglin, I. Zammuto
Tungsten heavy alloy (97 wt% W, 2 wt% Ni, 1 wt% Fe) was investigated as an alternative for tungsten (W) as plasma facing material. It is produced commercially by several companies and compared to bulk W it is readily machinable and considerably cheaper. In order to qualify the material for use in the divertor of the mid-size tokamak ASDEX Upgrade (AUG) dedicated laboratory investigations as well as high heat flux tests in the neutral beam facility GLADIS were performed. These investigations revealed that the thermal conductivity at high temperature is close to that of W, the magnetisation is small and saturates already at low magnetic field and the hydrogen retention is similarly low as that of W. In high heat flux tests at power densities up to 20 MWm−2 no failure was observed up to the melting temperature (≈1500∘ ≈ 1500 ∘ C) of the binder phase. Even at surface temperatures of up to 2200 °C the mechanical integrity was sustained. Mechanical tests confirm the ductile behaviour of the W heavy alloy at room temperature and finite element analyses using the aforementioned data suggest a lower tendency for cracking. The increase of the long term dose-rate resulting from the activation of Ni under neutron irradiations appears to be moderate. During the 2017 campaign more than one fifth of the AUG divertor tiles consisted of W heavy alloy. Under nominal operation conditions the tiles showed no macroscopic failure and no increased Fe/Ni influx into the plasma was detected. Even though a few tiles showed strong melting at the edges due to accidental misalignment no failure due to cracking was observed.
Restructuring in high burnup UO2 studied using modern electron microscopy ☆ J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-02 Tyler J. Gerczak, Chad M. Parish, Philip D. Edmondson, Charles Baldwin, Kurt A. Terrani
Modern electron microscopy techniques were used to conduct a thorough study of an irradiated urania fuel pellet microstructure to attempt at an understanding of high burnup structure formation in this material. The fuel was irradiated at low power to high burnups in a light water reactor, proving ideal for this purpose. Examination of grain size and orientation with strict spatial selectivity across the fuel pellet radius allowed for capturing the progression of the restructuring process, from its onset to full completion. Based on this information, the polygonization mechanism was shown to be responsible for restructuring, involving formation of low-angle grain boundaries with their initiation occurring at the original high-angle grain boundaries of the as-fabricated pellet and at the gas bubble-matrix interfaces. The low-angle character of boundaries between the subdivided grains disappeared in the fully developed high burnup structure, likely due to creep deformation in the pellet.
Evaluation of Fuel-Clad Chemical Interaction in PFBR MOX test fuel pins J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-02 V.V. Jayaraj, S. Thirunavukkarasu, V. Anandaraj, B.K. Ojha, Ran Vijay Kumar, S. Vinodkumar, M. Padalakshmi, C. Padmaprabu, C.N. Venkiteswaran, V. Karthik, R. Divakar, B. Purna Chandra Rao, Jojo Joseph
Fuel Clad Chemical Interaction (FCCI) is one of the life limiting issues in the MOX fuel pins of fast breeder reactors. Clad wastage due to FCCI coupled with stress arising from fission gas pressure and Fuel Clad Mechanical Interaction (FCMI) due to fuel swelling can lead to fuel pin failure. Non-destructive evaluation (NDE) techniques such as Gamma Scanning, Eddy current testing and Neutron Radiography have been successfully used on MOX fuel pins of a test sub-assembly irradiated in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GW d/t to detect signatures of FCCI. The results obtained by the various NDE techniques have been correlated and verified through metallographic examination on fuel pin cross-sections. The measured clad wastage of 85 μm agrees well with a model developed for MOX fuel with high Pu content. The results of examinations have enabled validation of the model and given confidence to the designer that PFBR MOX fuel can safely attain the target burn-up of 100 GW d/t.
Corrosion of the bonding at FeCrAl/Zr alloy interfaces in steam J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-02 Dongliang Jin, Na Ni, Yi Guo, Zhonghua Zou, Xin Wang, Fangwei Guo, Xiaofeng Zhao, Ping Xiao
The interfacial bonding at protective coatings and Zr alloy substrates could have a significant effect on the corrosion behaviors of the coatings. In this work, FeCrAl-Zr couples were fabricated at 900 °C by spark plasma sintering (SPS), and the interfacial bonding of the couples was modified through the addition of Si and/or thermal diffusion at 1100 °C–1300 °C. Uphill diffusion of silicon occurred during SPS. The corrosion resistance of the couples in 400 °C/10.3 MPa steam, and the effects of the interfacial bonding characteristics were studied. Due to the presence of ZrSi2 in the interfacial layer, the interfacial bonding of the FeCrAl/Si-Zr couple had a better corrosion resistance than that of the FeCrAl-Zr couple, withstanding 14 days of corrosion in steam compared to 7 days for the FeCrAl-Zr couple. The results also suggested that bimetallic effect could accelerate the oxidation of Zr alloy at the interface of the couples, and galvanic corrosion significantly contributed to the serious corrosion of the interface of the diffusion bonded couples with thermal diffusion. The influence of oxide growth stress around interface on the bonding lifetime of the couples was also discussed.
Ion beam induced phase transformation and krypton bubble formation in monoclinic zirconium oxide J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-01 P. Balasaritha, S. Amirthapandian, P. Magudapathy, R.M. Sarguna, S.K. Srivastava, B.K. Panigrahi
Low energy krypton ion beam induced phase transformation and bubble formation in monoclinic zirconia (ZrO2) has been studied using electron microscopy, Raman scattering, grazing incidence X-ray diffraction and photoluminescence spectroscopy. The zirconia samples were synthesized by the thermal decomposition method and 60 keV Kr+ ion irradiation was carried out at 300 K and 143 K. The as-sintered ZrO2 particles were found to be monoclinic in structure, however, upon 60 keV Kr+ ion irradiation, a fraction of the sample (∼4.9% (300 K) and ∼8.3% (143 K) for the ion fluence of 1 × 1017 ions/cm2) was transformed from monoclinic to tetragonal phase along with the formation of krypton bubbles. The size of the bubbles is found to increase with the ion fluence, irrespective of the sample temperature during ion irradiation. The phase transformation from monoclinic to tetragonal structure was mainly due to radiation damage process where the driving force is strain field associated with the O-vacancies. The monoclinic to tetragonal transformation rate is faster when the ion irradiation was carried out at 143 K and this is attributed to the immobility of defects and production of large number of oxygen vacancies. The krypton bubble formation might be hindering the monoclinic to tetragonal phase transformation rate.
Influence of mean stress and light water reactor environment on fatigue life and dislocation microstrucures of 316L austenitic steel J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-01 P. Spätig, M. Heczko, T. Kruml, H.-P. Seifert
Influence of mean stress on fatigue life of the austenitic stainless steel 316 L in air and light water environments (boiling water reactor/hydrogen water chemistry) at 288 °C was determined with a series of tests carried out in load-control mode. Fatigue life was found to increase with application of compressive and tensile mean stress in air and light water reactor environments. Secondary hardening was regarded as the main reason for this behavior. A modified Smith-Watson-Topper (SWT) model was considered to account for mean stress and was shown to predict fatigue life accurately in air and water environments. The reduction of fatigue life in water environment, determined with the SWT curves, was about 2.5. Observations of the end-of-life dislocation arrangements by transmission electron microscopy showed that the dislocation microstructure depends essentially on plastic strain amplitude, which in turn is strongly correlated to stress amplitude and mean stress. The microstructures were found consistent with those usually observed after strain-controlled experiments. At rather low plastic strain amplitudes, corduroy structure consisting of small dislocation loops was observed. Acting as significant obstacle to dislocation motion, corduroy structure affects overall dislocation mobility therefore contributing to notable secondary cyclic hardening.
Thermal stability of Li films on polycrystalline molybdenum substrates J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-01 O. Fasoranti, B.E. Koel
Lithium (Li) coatings on plasma facing components (PFCs) have been proposed as potential solutions to first wall and divertor challenges in tokamak fusion reactors. We report on the thermal behavior of ultrathin pure Li films deposited on polycrystalline substrates of molybdenum (Mo) and a molybdenum alloy (titanium zirconium molybdenum, TZM). These Li films were studied under controlled ultrahigh vacuum (UHV) conditions and thermal stabilities were primarily compared via temperature programmed desorption (TPD) measurements. In addition, on TZM, which is of particular interest, we obtained additional spectroscopic data using Auger electron spectroscopy (AES) and low energy ion scattering (LEIS) to further characterize the film structure and composition. The monolayer of Li in these films in contact with the substrate is bound much stronger than in bulk Li films, and thermally desorbs at much higher temperatures. Interfacial Li on Mo (poly) has a higher thermal stability than that on TZM(poly), where the limiting values for the desorption activation energies, Ed, are 3.56 and 2.84 eV, respectively, in the low coverage, high temperature desorption tail. LEIS indicates some clustering or interdiffusion of the Li films on the TZM substrate at 500 K. No appreciable irreversible absorption of Li occurs on Mo or TZM under the conditions of these experiments.
Electrochemistry of UBr3 and preparation of dendrite-free uranium in LiBr-KBr-CsBr eutectic melts J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-01 Hao Tang, Yunfeng Du, Yingru Li, Ming Wang, Huaisheng Wang, Zhenliang Yang, Bingqing Li, Rui Gao
The electrochemistry of UBr3 was studied on W working electrodes in LiBr-KBr-CsBr eutectic salt at 623 K. The U(III)/U (0) redox reaction was evaluated with respect to its electrochemical behavior, diffusion and electrocrystallization properties. According to cyclic voltammetry and square wave voltammetry, the reduction of U(III) ions to U (0) metal is a one-step three-electron reaction and the oxidation of U(III) ions is an one-electron transferred reaction. Cyclic voltammograms at varied scan rates show that U(III)/U (0) redox reactions are quasi-reversible. Chronopotentiometry and Sand's equation were used to determine the diffusion coefficients of U(III) ions. The nucleation mechanisms of uranium metal deposited on W substrates with different UBr3 concentrations were predicted by using Scharifker-Hill model, which shows that the nucleation and growth mode changes from progressive mechanism to instantaneous mechanism with an increasing concentration of UBr3 in the salt.At 623 K, potentiostatic electrodeposition was conducted to prepare mass uranium. The current-time evolution curve indicates that the actual surface area of electrodeposits keeps stable. Furthermore, the macroscopic appearance and microcosmic morphology show compact structure, rather than dendrites. Bulk uranium metal was further analyzed by inductive coupled plasma atomic emission spectrometer and X-ray diffraction, which indicate that the compact uranium is high-purity metal.
Effect of heavy ion pre-irradiation on blistering and deuterium retention in tungsten exposed to high-fluence deuterium plasma J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-01 Shiwei Wang, Xiuli Zhu, Long Cheng, Wangguo Guo, Mi Liu, Chuan Xu, Yue Yuan, Engang Fu, Xing-Zhong Cao, Guang-Hong Lu
Surface blistering and deuterium (D) retention of heavy ion pre-irradiated (1 dpa) tungsten (W) exposed to low-energy (40 eV) and high-flux (1–2 × 1022 D/m2s) D plasma has been investigated with low fluence of 0.1 × 1027 D/m2 and a high fluence of 2.2 × 1027 D/m2. Surface morphology observations show that a large number of blisters are formed on the undamaged W after low-fluence exposure while the area density of the blister will have significantly increased and blister bursting will be triggered in the case of high-fluence exposure. In contrast, the heavy ion pre-irradiation noticeably reduces the area density of blisters for both low- and high-fluence exposures. The thermal desorption spectroscopy (TDS) shows that the total D retention in the pre-damaged W is greater than that of the undamaged sample, and this trend is more significant with increasing fluence/duration of D plasma exposure. The results of positron annihilation Doppler broadening spectrometry (PA-DBS) and TDS indicate that a large number of vacancy-type defects, especially those with a higher trapping energy of D, are induced by the heavy ion pre-irradiation. The increasing defects/D-trap sites may result in two outcomes, the D concentration will disperse at each blister nucleation site and D inward diffusion is enhanced and, therefore, lead to the mitigation of D-induced blistering and increase the D retention. In addition, combined with the enhanced D inward diffusion caused by increasing exposure duration, the D retention is therefore much higher in the pre-damaged W in the case of high-fluence exposure.
Evaluation of anisotropic deformation behaviors in H-charged Zircaloy-4 tube J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-01 Yao Ding, Ju-Seong Kim, Hoa Kim, Chanhee Won, Seungho Choi, Sung Hyuk Park, Jonghun Yoon
In this paper, we focus mainly on evaluating the anisotropy evolution in Zircaloy-4 tubes with respect to hydrogen pickup which tends to affect the deformation behavior and fracture elongation due to embrittlement phenomenon induced by Zr hydrides. To capture the complex material behaviors of the H-charged (166 ppm) Zircaloy-4 tubes, we have applied the Hill anisotropic yield criterion by carrying out several material tests, such as the axial tensile, ring tensile, and axial crushing tests, for calibrating the anisotropic coefficients based on the directional strength and strain ratios. It makes possible to simulate the directional flow curves of the as-received and the H-charged Zircaloy-4 tubes based on the directional strength ratios along the axial and hoop directions, respectively, as well as a variation in deformation modes in the axial crushing test with substantial accuracy.
Fretting wear comparison of cladding materials for reactor fuel cladding application J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-01 Thomas C. Winter, Richard W. Neu, Preet M. Singh, Lynne E. Kolaya, Chaitanya S. Deo
Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, such grid-to-rod-fretting is a significant cause of fuel failures in U.S. pressurized water reactor power plants. Kanthal advanced powder metallurgy technology or APMT, an FeCrAl steel alloy, and a braided SiC fiber, Chemical Vapor Infiltration SiC matrix (SiC/SiC) cladding by General Atomics are possible alternatives to conventional fuel cladding in a nuclear reactor due to their favorable performance under accident conditions. Tests were performed to examine the reliability of the cladding candidates and a conventional cladding, Zircaloy-4, under dry fretting conditions at elevated temperature. The contact was simulated with a rectangular and a cylindrical specimen over a line contact area. Confocal scanning laser microscopy was used to obtain a 3D map of the surface, which was in turn used for wear and work rate calculations on the samples. The wear rate coefficient was used as a measure of the performance and wear under fretting. Additionally, Energy Dispersive Spectroscopy was performed to qualitatively describe the microchemical changes the material undergoes during fretting. While APMT steel and SiC/SiC can perform favorably in loss of coolant accident scenarios, they also need to perform well when compared to Zircaloy-4 with respect to fretting wear. Wear coefficient measurements showed that APMT steel performs favorably in comparison to Zircaloy-4 with respect to fretting wear.
Post-test examinations on Zr-1%Nb claddings after ballooning and burst, high-temperature oxidation and secondary hydriding J. Nucl. Mater. (IF 2.048) Pub Date : 2018-06-01 Eszter Kozsda-Barsy, Katalin Kulacsy, Zoltán Hózer, Márta Horváth, Zoltán Kis, Boglárka Maróti, Imre Nagy, Richárd Nagy, Tamás Novotny, Erzsébet Perez-Feró, Anna Pintér-Csordás, László Szentmiklósi
The objective of the present study was to provide further data on E110G cladding behaviour. The results presented here are from new post-test examinations (PTEs) carried out on samples of secondary hydriding experiments conducted earlier in MTA EK.The as-received Zr-1%Nb cladding samples were pressurised at high temperature to balloon and burst and then oxidised in steam atmosphere. The post-test investigation was focusing on geometric change in the cladding, ductility, oxidation and hydrogen absorption.Outer and inner oxide layers were formed on the samples, with increasing thickness near the thermal centre. The results include radial and axial distribution of oxygen in the cladding after oxidation. The hydrogen uptake of the alloy shows the expected characteristic axial distribution. Mechanical testing of the oxidised and non-oxidised samples confirmed the results of the previous mechanical tests that after ballooning the samples still had notable flexural strength, whereas after oxidation this decreased.The results were evaluated against those obtained through simulations, making it possible to estimate the level of oxidation, and to develop better models through further simulation.
A review of microstructural features in fast reactor mixed oxide fuels J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-31 Riley Parrish, Assel Aitkaliyeva
As an alternative to traditional uranium dioxide (UO2) fuels, mixed oxide (MOX) fuels were developed to dispose of industrial and military stores of plutonium (Pu) through the incorporation of plutonium dioxide (PuO2) powder into a UO2 base fuel. The high temperature and chemical stability characteristic of oxide fuels would be maintained, while the added Pu would ultimately be eliminated from long term storage. Plutonium could be extracted from spent light water reactor (LWR) fuels, acting as an additional step to close the fuel cycle and mitigate potential environmental or proliferation concerns. This review summarizes the primary features associated with fast reactor MOX fuels, including fuel restructuring, actinide redistribution, solid fission products, plutonium agglomerates, joint oxide gain, and fuel-cladding chemical interaction. A summary of research efforts within the last 10 years and directions for future research are discussed.
Anisotropy in the thermal expansion of uranium silicide measured by neutron diffraction J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-30 E.G. Obbard, K. Johnson, P. Burr, D.A. Lopes, D.Gregg, K.-D. Liss, G. Griffiths, N. Scales, S.C. Middleburgh
Hydrogen diffusion under stress in Zircaloy: High-resolution neutron radiography and finite element modeling J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-30 Weijia Gong, Pavel Trtik, Stéphane Valance, Johannes Bertsch
He irradiation effects in bulk Cu/V nanolayered composites fabricated by cross accumulative roll bonding J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-30 L.F. Zeng, P. Fan, L.F. Zhang, R. Gao, Z.M. Xie, Q.F. Fang, X.P. Wang, D.Q. Yuan, T. Zhang, C.S. Liu
Bulk nanolayered Cu/V composites simultaneously exhibit high strength and outstanding thermal stability due to its unusually high density of interfaces. However, investigation of the irradiation stability of this material is still in its infancy, limiting further application of these materials under radiation conditions. Herein we investigated the radiation response of bulk nanolayered Cu/V composites exposed to 200 keV He ions with two irradiation fluences of 2 × 1021 ions/m2 and 7 × 1022 ions/m2. It is demonstrated that the bulk Cu/V nanolayered composites remained stable with respect to mechanical property and microstructure after irradiation with fluence of 2 × 1021 ions/m2. In contrast, for materials exposed to high irradiation fluence of 7 × 1022 ions/m2, severe irradiation damage, such as obvious surface blistering, elongated He voids, and layer morphological instability, were developed. In addition, asymmetric He bubbles distribution and obvious He bubble-free zones were observed near the Cu/V interfaces and within layers. The mechanisms of radiation-induced instabilities and He bubbles formation are discussed in detail.
Numerical simulation of grain boundary carbides evolution in 316H stainless steel J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-30 Qingrong Xiong, Joseph D. Robson, Litao Chang, Jonathan W. Fellowes, Mike C. Smith
In the present work, a numerical model based on the coupling of Kampmann and Wagner Numerical (KWN) framework and thermodynamic software ThermoCalc has been developed to predict grain boundary precipitate evolution in 316H stainless steel during thermal aging. The model is calibrated and validated against precipitate size distributions obtained by accelerated isothermal heat treatment and analysed using scanning electron microscopy (SEM). Elemental distribution was also investigated using electron microprobe analysis (EPMA). The predicted average particle size, particle size distribution and precipitate number density predicted by the model were found to be in good agreement with the experimental results. The model was then applied to predict the particle size distribution after several years exposure at service temperature. It is demonstrated that these predictions are consistent with measurements from a service-exposed part. The sensitivity of the precipitate size distribution to temperature is emphasised, and it is demonstrated that the model has potential as a useful tool for predicting evolution of the precipitate size distribution during service, providing reliable thermal data are available for the whole service life.
Effect of addition of Ti on hardness change during tempering in reduced activation ferritic/martensitic (RAFM) steels J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-30 Byung Hwan Kim, Jae Hoon Jang, Woo-Kyung Seol, Joonoh Moon, Tae-Ho Lee, Hyoung Chan Kim, Chang-Hoon Lee, Kyung-Mox Cho
The specificities of hardness changes during tempering in Ti-containing RAFM steels despite larger fraction of MX precipitate were found during the development of new RAFM steels with improved mechanical properties. The decline in hardness during tempering of RAFM steels where Ti replaced Ta (Ti-RAFM) and RAFM steel with both Ta and Ti (TaTi-RAFM) was higher than that in conventional RAFM steel with Ta. It is believed that Ti addition accelerated the growth of MX precipitates and thereby accelerated diffusion of carbon from the martensite matrix, leading to a large decrease in hardness of martensite in Ti-containing RAFM steels during tempering.
Tensile properties and deformation microstructure of highly neutron-irradiated 316 stainless steels at low and fast strain rate J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-31 A. Renault-Laborne, J. Hure, J. Malaplate, P. Gavoille, F. Sefta, B. Tanguy
Post-irradiation deformation behavior of solution-annealed (SA) and cold-worked (CW) 316 austenitic stainless steel irradiated to doses from 9 to 39 dpa is examined as a function of strain rate and irradiation conditions (neutron spectrum, temperature). Tensile properties are found to be significantly higher for lower irradiation temperature and for CW material, for similar irradiation levels. The effect of strain rate on tensile properties is shown to be weak in the range [10−8s−1; 10−4s−1]. TEM investigations after deformation for levels of plastic strain of about 1% show on SA 316 the presence of deformation bands corresponding to one or even a mixture of twins, extended stacking faults, α′-martensite islands and ε-martensite nanobands. Bundles of crisscrossing bands, found to be a composite of overlapping stacking faults, nanotwins and ε-martensite nanolayers, are observed at TEM foils edges near the grain boundaries with α′-martensite islands decorating these edges. Except observation of a slight decrease of the number of deformation bands in the specimen deformed at slower strain rate, no qualitative microstructural differences appear between specimens tested at slow and fast strain rates.
Uranium exchange kinetics in a molten LiCl-KCl/Cd system at 500 °C J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-31 Tae-Sic Yoo, Guy L. Fredrickson, Gregory M. Teske
A dissolution experiment is performed in order to examine the kinetics of uranium isotope equilibration in a molten salt and cadmium system. The salt contains a solute concentration of UCl3 in LiCl-KCl eutectic and the cadmium contained dissolved uranium at 500 °C. A small piece of high enriched uranium is introduced to the cadmium, which perturbs the equilibrium with respect to the uranium enrichment of the salt and cadmium phases. The series of salt and cadmium samples trace how the enrichment of the two phases reach equilibrium (in the absence of a chemical driving force) over the course of approximately 24 h. A model linking dissolution and isotope exchange kinetics is developed to track the evolution of uranium isotope composition. With the assumption of unaccounted salt mass incurred from the dated sample reporting practice, the model explains the trend of the measured uranium isotope composition well.
Mechanical performance of neutron-irradiated dissimilar transition joints of aluminum alloy 6061-T6 and 304L stainless steel ☆ J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-29 Richard H. Howard, Ryan C. Gallagher, Kevin G. Field
Bimetallic transition joints using aluminum alloy 6061-T6 and stainless steel 304 L are useful in providing reliable stainless steel welds on nuclear components at temperatures below 200 °C, while maintaining the attractive radiation tolerance of the aluminum alloy. The mechanical performance of inertia welded Al6061-T6:304LSS transition joints was evaluated after neutron irradiation up to 3.45 dpa at 100 °C (maximum) in the High Flux Isotope Reactor to determine the viability of using these transition joints for nuclear and reactor applications. Neutron radiation produced moderate hardening (Δσy≈≈90 MPa) with limited change in ductility. Tensile specimens were produced from multiple transition joints and no batch-to-batch variation was found. Tensile responses were found to align with typical responses of wrought Al6061-T6, indicating that the behavior of the joints was dictated by the Al6061-T6 section of the joint.
The trapping and dissociation process of hydrogen in tungsten vacancy: A molecular dynamics study J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-29 Baoqin Fu, Mingjie Qiu, Jiechao Cui, Min Li, Qing Hou
Tungsten (W) is a primary candidate for plasma facing materials (PFM) for future fusion devices. The interaction between hydrogen (H) and vacancy (V) is the key for understanding many material behaviors under irradiation. Therefore, it is necessary to study carefully the kinetic process between H and W vacancy. In this work, the dynamical parameters, including effective capture radii (ECRs) and dissociation coefficients, for various trapping and dissociation processes (VHx + H⇌VHx+1), have been investigated using an ingenious method based on molecular dynamics (MD) simulations. It was found that the parameters are dependent not only on the reaction types but also on the temperatures. The ECRs decrease gradually as the increase of the trapped H atoms in the W vacancy, and decrease roughly with increasing temperature for T < 1200 K. The dissociation energies decrease gradually as the increase of the trapped H atoms in the W vacancy. The evolution of concentration of the trapped H atoms in W vacancy was investigated by coupling the trapping process and dissociation process and using the dynamical parameters calculated by the MD simulations. The H retention in W obviously depends on the state of trapping sites and the temperatures. These results should be potentially applicable for the long-term simulation methods such as kinetic Monte Carlo (KMC) and rate theory (RT) models.
Chemical thermodynamics of RuO2(s) J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-29 C. Chatillon, I. Nuta, F.Z. Roki, E. Fischer
Thermodynamic data for the ruthenium oxide RuO2(s) are basic data for the calculation of gaseous release in the case of severe nuclear accident. The present study is a critical analysis of thermochemical data for RuO2(s) based on available published experimental data from calorimetry, vapor pressures in equilibrium with the diphasic Ru<img border="0" alt="single bond" src="https://cdn.els-cdn.com/sd/entities/sbnd">RuO2 and Electromotive Force Measurements EMF. A full critical review and reinterpretation of data are presented in this work with a proposition of new and accurate data for RuO2(s): C°p (RuO2, s, 298.15 K) = 56.42 ± 0.08 J·K-1 ·mol-1, S°298.15 (RuO2,s) = 46.15 ± 0.05 J·K-1 ·mol-1, and Δ; fH298 (RuO2,s) = −312.3 ± 1.6 kJ.mol-1,
Impact of a dense helium-bubble superlattice on the deformation of copper by twinning J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-28 Ian S. Winter, Zhang-Jie Wang, Peter Hosemann, D.C. Chrzan
Recent experimental work on helium-irradiated single-crystal (SC) and nano-twinned (NT) copper has shown that the formation of a helium bubble superlattice has a dramatic effect on the critical resolved shear stress of twin formation in SC copper pillars and twin migration in NT copper pillars. The mechanisms governing this dramatic change in the mechanical response of the material after helium irradiation are explained theoretically in this work. Atomistic simulations show that the presence of a helium bubble superlattice has a profound effect on twin nucleation and propagation, which are related to the mechanical responses of SC and NT copper respectively. Based on the simulations, we estimate that the bubble lattice can decrease the ideal twin nucleation stress compared to pure copper by approximately 50% 50 % . The modeled response of a step in a twin boundary to an applied shear stress leads to an increase in the critical resolved shear stress due to the bubble lattice. The computed increase is in qualitative agreement with experimental findings for NT copper.
Hydrogen diffusion behavior in tungsten under anisotropic strain J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-29 Xuesong Zhang, Ke Xu, Liang-Liang Niu, Ying Zhang, Guang-Hong Lu
In future fusion devices, the incident hydrogen plasma with high mobility can diffuse deep into tungsten bulk, which is directly relevant with hydrogen isotopes permeation and retention in tungsten. In this work, density functional theory (DFT) and object kinetic Monte Carlo (OKMC) simulations are adopted to investigate the hydrogen diffusion behavior in tungsten under anisotropic uniaxial strain from −2.5% to 2.5%. As presented by our DFT calculations, there are two types of hydrogen diffusion paths when applying strain, including one path perpendicular to the strain direction and another path largely along the strain direction. The migration energy barriers of these two paths have opposite variation tendencies in tensile or compressive condition. Our OKMC calculations based on DFT input show that, in tensile condition, the hydrogen diffusion is restrained despite the lower energy barrier of the corresponding diffusion path. In compressive condition, the hydrogen diffusion along the strain direction is enhanced, while that perpendicular to the strain direction is suppressed. The hydrogen diffusivity under anisotropic strain at the temperature range from 400 K to 1800 K is determined. It is demonstrated that tensile strain can suppress the diffusivity, while compressive strain can either suppress or facilitate the diffusivity depending on the temperature and the strain value. The anisotropic strain exhibits distinct effect on hydrogen diffusivity at lower temperature but its effect is minimal as the temperature increases.
Chemical interactions between pre-oxidized Zircaloy-4 and 304 stainless steel-B4C melt at 1300 °C J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-29 Lichun Zheng, Kazuya Hosoi, Shigeru Ueda, Xu Gao, Shin-ya Kitamura, Yoshinao Kobayashi, Ayako Sudo
During severe nuclear accidents, control rods rapidly liquefy at temperatures above 1250 °C due to eutectic reaction, forming a 304 stainless steel (304SS)-B4C melt. The melt will relocate and attack surrounding fuel rod claddings made of Zircaloy-4 (Zry-4). To understand to what extent ZrO2 oxide scale formed on Zry-4 will protect Zry-4 claddings against 304SS-B4C melt attack, we studied the chemical interactions between pre-oxidized Zry-4 and 304SS-B4C melt at 1300 °C. Bare Zry-4 was completely dissolved in 304SS-B4C melt within 60 min. The presence of ZrO2 oxide scale on Zry-4 significantly delayed the interactions, especially when ZrO2 oxide scale was dense. Typically, the reaction zone consists of ZrB2, Zr6(Fe,Ni,Cr)23 and Fe-Ni-Cr metallic phase at room temperature. Due to the presence of α-Zr and β-Zr in Zry-4 metal matrix, ZrO2 oxide scale becomes thermodynamically unstable. Dissolution of dense ZrO2 oxide scale can be described in three stages with different dissolution rates. Dissolution of ZrO2 oxide scale provides Zr source for the growth of reaction zone. Generally, the thickness of reaction zone linearly increases with time. Compared with the reaction couples of pre-oxidized Zry-4 and solid 316SS, both ZrO2 dissolution rate and reaction zone growth rate are much slower in the reaction couples of pre-oxidized Zry-4 and 304SS-B4C melt. The corresponding reasons were discussed.
Microstructural characterization of as-cast U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-29 Michael T. Benson, Lingfeng He, James A. King, Robert D. Mariani, Alexander J. Winston, James W. Madden
Palladium is being investigated as a potential additive to metallic fuel to bind fission product lanthanides, with the goal of reducing or preventing fuel-cladding chemical interactions (FCCI). A primary cause of FCCI is the lanthanide fission products moving to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. The current study is a scanning electron microscopy (SEM) and transmission electron microscopy (TEM) investigation of as-cast U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln in wt. %, where Ln = 53Nd-25Ce-16Pr-6La. In U-20Pu-10Zr-3.86Pd, PdZr2 is forming, along with a possible ternary phase between Pu, Zr, and Pd. Pu is also present in the Pd-Ln precipitates formed in U-20Pu-10Zr-3.86Pd-4.3Ln. In the LnPd phase, Pu appears to be substitutional, forming (Ln,Pu)Pd. The other prominent phase, which appears to be Ln7Pd3, has a fine, lamellar structure. The lanthanides remain essentially constant across this fine structure, but Pu and Pd alternate as to which has the higher concentration.
Molecular dynamics simulation of hydrogen and helium trapping in tungsten J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-25 Petr Grigorev, Aleksandr Zinovev, Dmitry Terentyev, Giovanni Bonny, Evgeny E. Zhurkin, Guido Van Oost, Jean-Marie Noterdaeme
Tungsten has been chosen as the divertor armour material in ITER and is the main candidate material for plasma-facing components for future fusion reactors. Interaction of plasma components with the material leads to degradation of the performance and thus the lifetime of the in-vessel components. On top of that special attention is drawn to tritium retention in the reactors vessel from a safety point of view, since tritium is radioactive material. In order to gain better understanding of the mechanisms driving accumulation of plasma components in the material and subsequent degradation of the material, atomistic simulations are employed. The focus of this work is on so-called self trapping of H and He atoms or, in other words, Frenkel pair formation in bulk tungsten in the presence of H and He atoms. Two versions of a model embedded atom interatomic potential and a bond order potential were tested by comparing it with ab initio data regarding the binding properties of pure He and He-H-Vacancy clusters and energetics of Frenkel pair formation. As a result of Molecular Dynamics simulations at finite temperature, the values of critical H concentration needed for the generation of a Frenkel pair in the presence of He clusters were obtained. The results show that the critical H concentration decreases with the size of He cluster present in the simulation cell and thus, Frenkel pair formation by H is facilitated in the presence of He clusters in the material.
Early progress on additive manufacturing of nuclear fuel materials J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-26 A. Bergeron, J.B. Crigger
Additive manufacturing of thorium dioxide has been investigated using a commercially available stereolithography-based 3D printer and photopolymer resin. Three-dimensional thorium dioxide objects have been printed with good dimensional accuracy. High-density thorium dioxide parts (>90% theoretical density) were achieved by sintering the 3D-printed parts. Despite significant shrinkage, the overall shape of the objects was maintained during sintering with slight distortion. Additive manufacturing is seen to have potential application for advanced nuclear fuel concepts with complex geometries.
Hydrogen trapping in helium-implanted W and W-Ta alloy: First-principles approach J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-26 ChuBin Wan, SuYe Yu, Xin Ju, WenWen Wang
We reveal the interaction of H with He in pure W and W-Ta alloy based on first-principles calculations. We show a strong attraction between H and He in both systems that drives H segregation towards He. The substitutional He and tetrahedral interstitial H defects in W-Ta alloy are more energetically favorable than pure W due to the decreased electronic density of the replaced Ta atom. Moreover, 1 He-Vac complex in both systems can trap as many as 12 H atoms. Based on the calculated formation energy of Hn-He-Vac complexes, the H3-He-Vac has the lowest formation energy in both systems. We believe that such understanding is generally applicable for H bubble formation in metals and metal alloys.
Ionizing radiation effects on the thermal stability of deuterium trapping in SiC J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-24 P. Muñoz, A. Moroño, F.J. Sánchez, A. Maira, I. García-Cortés
SiC materials are prime candidates for flow channel inserts in the dual coolant lithium lead blanket concept. Flow channel inserts made of SiC will be exposed to tritium from the Li transmutation, as well as to neutron and gamma radiation. Hence a critical issue for future fusion devices is to clarify hydrogen isotope behaviour in SiC under such conditions. The objective of the work presented here is to study the effect of ionizing radiation on the deuterium trapping in SiC in similar conditions as reactor materials. This effect is evaluated by studying the influence of ionizing radiation on deuterium trapping (for both implanted and loaded SiC samples). Moreover, it is investigated how deuterium trapping may be modified by displacement damage. The ionizing radiation effect on absorption has also been evaluated for samples pre-damaged by self-ion irradiation. The irradiation and implantation experiment have been carried out at the CMAM-UAM accelerator, and the Danfysik implanter and 2 MeV Van de Graaff electron accelerator at CIEMAT. Samples are analysed by thermal desorption spectroscopy and secondary ion mass spectrometry (SIMS) to clarify the mechanisms involved in the trapping processes, depending on the different experimental conditions. The results for the deuterium loaded samples indicate that absorption is increased by ionizing radiation. When samples are pre-damaged by C+4 ions, deuterium absorption is increased in the form of Si-D according to SIMS results. Furthermore, the effect of ionizing radiation after deuterium implantation is an enhancement of the deuterium released from SiC. The deuterium release observed in this case is forming hydrocarbons during irradiations.
First principle study of tritium trapping at oxygen vacancies in Li4SiO4 J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-24 Yanli Shi, Tiecheng Lu, Tao Gao, Xiaogang Xiang, Yichao Gong, Mao Yang, Lan Feng, Hailiang Wang, Chen Dang
Tritium trapping at point defects is a major issue concerning tritium extraction efficiency in solid state breeder materials in future fusion reactors. Here, tritium trapping behaviors at oxygen vacancies in Li4SiO4 have been investigated by density functional theory simulations. Formation energies have been calculated to determine the stability of defects in various charge states. Density of states, charge distribution and atomic charge has been calculated to investigate the mechanism of defect formation. The results showed that the neutral and 2 + charged oxygen vacancies are the most stable species and are both diamagnetic. The 1 + charged paramagnetic oxygen vacancy (E′ center) is less stable. In addition, the tritium-trapped O-vacancy is most stable when the defect complex is 1 + charged. The trapped tritium is bonded to Si forming a T-SiO3 unit. Other charge states of the complex may lead to unstable defect structures and high formation energies, forcing tritium to escape the vacancy.
Microscopic origin of black spot defect swelling in single crystal 3C-SiC J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-24 Ka Yu Fung, Yan Ru Lin, Pei Jun Yu, Ji Jung Kai, Alice Hu
In this study, we perform a series of simulation of a high-energy particle irradiation on a 3C-SiC at low temperature through molecular dynamic analysis. In order to determine the formation mechanism of black spot defects (BSD), the evolution of defect clusters during the cascade process is examined. Simulation results show that there are more isolated interstitials scattering across the structure while the less mobile vacancies are concentrated in defect clusters, which is consistent with the depleted zone theory proposed by Brinkman . These results also match the TEM observation and simulation results done by Lin et al.  and support the argument that black spot defects are in fact vacancy-rich regions, with individual interstitials spreading into bulk, stretching the lattice structure.
Wetting of liquid lithium on fusion-relevant materials microtextured by femtosecond laser exposure J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-24 S. Hammouti, B. Holybee, M. Christenson, M. Szott, K. Kalathiparambil, S. Stemmley, B. Jurczyk, D.N. Ruzic
As the use of liquid metals in plasma facing components becomes more widespread, it is important to investigate how these liquid metals interact with the surfaces onto which they are deposited. An important example of these interactions is the ability to control liquid metal wettability on fusion relevant substrates. In this work, we explore the influence of femtosecond laser induced nanostructured surfaces on the wetting degree of liquid lithium versus temperature. Three material candidates as a lithium wall in magnetic fusion devices have been investigated: molybdenum, tungsten and 304 L stainless steel. Laser parameters were tuned to induce periodical self-organized nanostructures (ripples or LIPSS) formation on each material. Wettability of laser treated materials was changed from lithium-philic to lithium-phobic for temperatures beyond 320 °C - 360 °C compared to untreated material. The effect of both laser induced topography and chemistry are quantified to explain the observed liquid lithium contact angles on each material. Finally, it was shown that topography in the form of self-organized periodical nanostructures as well as the surface chemistry in the form of oxides enrichment, both induced by a single step laser process, strongly influence the wetting degree of liquid lithium and enhance lithium-phobicity at high temperatures.
Vaporization behaviour of the Molten Salt Fast Reactor fuel: The LiF-ThF4-UF4 system J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-23 A. Tosolin, O. Beneš, J.-Y. Colle, P. Souček, L. Luzzi, R.J.M. Konings
A selected composition of the initial fuel of the Molten Salt Fast Reactor (MSFR) is assessed by differential scanning calorimetry (DSC) for melting point determination and by Knudsen effusion mass spectrometry (KEMS) for vaporization behaviour. Partial vapour pressures and thermodynamic activities of the MSFR fuel mixture are discussed indicating departures from ideal behaviour, and further interpreted by phase equilibria calculations. The boiling point of the mixture is obtained extrapolating vapour pressure experimental results. New results on the vaporization behaviour of pure uranium tetrafluoride are presented, together with the ionization potentials of UF4 by electron impact.
A multiscale microstructural approach to ductile-phase toughened tungsten for plasma-facing materials J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-23 Ba Nghiep Nguyen, Charles H. Henager Jr., Nicole R. Overman, Richard J. Kurtz
Increasing fracture toughness and modifying the ductile-brittle transition temperature of a tungsten-alloy relative to pure tungsten has been shown to be feasible by ductile-phase toughening (DPT) of tungsten for future plasma-facing materials for fusion energy. In DPT, a ductile phase is included in a brittle tungsten matrix to increase the overall work of fracture for the material. This research models the deformation behavior of DPT tungsten materials, such as tungsten-copper composites, using a multiscale modeling approach that involves a microstructural dual-phase (copper-tungsten) region of interest where the constituent phases are finely discretized and are described by a continuum damage mechanics model. Large deformation, damage, and fracture are allowed to occur and are modeled in a region that is connected to adjacent homogenized elastic regions to form a macroscopic structure, such as a test specimen. The present paper illustrates this multiscale modeling approach to analyze unnotched and single-edge notched (SENB) tungsten-copper composite specimens subjected to three-point bending. The predicted load-displacement responses and crack propagation patterns are compared to the corresponding experimental results to validate the model. Such models may help design future DPT composite configurations for fusion materials, including volume fractions of ductile phase and microstructural optimization.
Impact of corrosion on the emissivity of advanced reactor structural alloys J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-23 J.L. King, H. Jo, A. Shahsafi, K. Blomstrand, K. Sridharan, M.A. Kats
Under standard operating conditions, the emissivity of structural alloys used for various components of nuclear reactors may evolve, affecting the heat transfer of the systems. In this study, mid-infrared emissivities of several reactor structural alloys were measured before and after exposure to environments relevant to next-generation reactors. We evaluated nickel-based alloys Haynes 230 and Inconel 617 exposed to helium gas at 1000 °C, nickel-based Hastelloy N and iron-based 316 stainless steel exposed to molten salts at 750–850 °C, 316 stainless steel exposed to liquid sodium at 650 °C, and 316 stainless steel and Haynes 230 exposed to supercritical CO2 at 650 °C. Emissivity was measured via emissive and reflective techniques using a Fourier transform infrared (FTIR) spectrometer. Large increases in emissivity are observed for alloys exposed to oxidizing environments, while only minor differences were observed in other exposure conditions.
Measurement of displacement cross sections of aluminum and copper at 5 K by using 200 MeV protons J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-21 Yosuke Iwamoto, Makoto Yoshida, Toshimasa Yoshiie, Daiki Satoh, Hiroshi Yashima, Hiroki Matsuda, Shin-ichiro Meigo, Tatsushi Shima
To validate the Monte Carlo codes for prediction of radiation damage in metals irradiated by > 100 MeV protons, we developed a proton irradiation device with a Gifford–McMahon (GM) cryocooler to cryogenically cool two 0.25-mm-diameter wire samples of aluminum and copper. By using this device, the defect-induced electrical resistivity changes related to the displacement cross section of copper and aluminum were measured under irradiation with 200-MeV protons at 5 K at the beamline of the cyclotron facility at RCNP, Osaka University. After irradiation to a 3.89 × 1018 proton/m2 flux, the damage rate of the aluminum sample was 1.31 × 10−31 Ωm3/proton at 185 MeV and that of copper was 3.60 × 10−31 Ωm3/proton at 196 MeV. Based on measurements of recovery of the accumulated defects in aluminum and copper through isochronal annealing, which is related to the defect concentration in the sample, about 50% of the damage remained at 40 K, with the same tendency observed in other experimental results for reactor neutron, fusion neutron, and 125-MeV proton irradiations. A comparison of the measured displacement cross sections with the calculated results of the NRT-dpa and the athermal-recombination-corrected displacement damage (arc-dpa) cross sections indicates that arc-dpa with the defect production efficiencies provided by Almazouzi for aluminum and Nordlund for copper provide better quantitative descriptions of the displacement cross section than NRT-dpa.
Pressure resistance welding of MA-957 to HT-9 for advanced reactor applications J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-21 N.D. Jerred, I. Charit, L.R. Zirker, J.I. Cole
This study examines the feasibility of using pressure resistance welding to create solid-state bonding between oxide dispersion strengthened MA-957 alloy cladding tubes and ferritic-martensitic HT-9 alloy end-plugs. The joint configuration is prototypic of the closure welding process used in the fabrication of reactor fuel pins. Optimized joining parameters were found to create sound metallurgical bonding at the joint interface. Using electron backscatter diffraction, a characteristic microstructural gradient is noted across the weld joint with dynamic recrystallization occurring at the thermomechanically affected zone (TMAZ) of the MA-957 whereas martensite formation occurs in both the TMAZ and heat-affected zone (HAZ) of the HT-9. Dynamic recrystallization coupled with slight agglomeration of the nano-sized oxide particles in the TMAZ of the MA-957 likely decreased the hardness whereas martensite strengthening in the HT-9 coupled with some degree of grain refinement pushed the microhardness to more than 500 HV. In order to level the hardness gradient, a post-weld heat treatment was applied on the joint creating a tempering effect in the harder HT-9 side and decreasing the hardness gradient. Further studies will be needed to understand all the performance characteristics of these type of joints.
Calculation of the displacement energy of α α and γ γ uranium J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-21 Benjamin Beeler, Yongfeng Zhang, Maria Okuniewski, Chaitanya Deo
Uranium (U) is alloyed with molybdenum (Mo) or zirconium (Zr) in order to stabilize the high-temperature body-centered cubic γ γ phase of uranium for use in nuclear reactors. Although these two alloy systems possess different mechanical, chemical and thermodynamic properties, they exhibit a similarity in that there exists a variable degree of phase decomposition from the cubic γ γ phase of uranium to the orthorhombic α α phase of uranium, depending on both the Mo/Zr content and fabrication conditions. These two phases of uranium are believed to exhibit distinct swelling and radiation damage behavior. Understanding the differences in behavior under irradiation between the α α and γ γ phases can provide valuable information to guide the manufacturing process of U alloys and can inform multi-physics, continuum-level fuel performance codes. The threshold displacement energy (TDE) is the minimum amount of kinetic energy required to displace an atom from its lattice site. It is critically important to determine an accurate value of the TDE in order to calculate the total number of displacements due to a given irradiation condition, and thus to understand the materials response to irradiation. In this study, molecular dynamics simulations have been performed to calculate the threshold displacement energy for both the α α and γ γ phases of uranium as a function of temperature. This study utilizes three different interatomic potentials that have been previously developed: U MEAM, U-Zr MEAM and U-Mo ADP. The threshold displacement energy in γU γ U at 800 K is 73.2 eV, 47.1 eV and 35.6 eV for the U MEAM, U-Zr MEAM and U-Mo ADP potentials. respectively. The threshold displacement energy for α α U at 600 K is 66.3 eV for the U-Mo ADP.
Japanese activities of the R&D on silicon carbide composites in the broader approach period and beyond J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-19 Takashi Nozawa, Kazumi Ozawa, ChangHo Park, Joon-Soo Park, Akira Kohyama, Akira Hasegawa, Shuhei Nogami, Tatsuya Hinoki, Sosuke Kondo, Toyohiko Yano, Tamaki Shibayama, Bun Tsuchiya, Tatsuo Shikama, Shinji Nagata, Teruya Tanaka, Hirotomo Iwakiri, Yasushi Yamamoto, Satoshi Konishi, Ryuta Kasada, Masatoshi Kondo, Tomoaki Kunugi, Takehiko Yokomine, Yoshitaka Ueki, Nariaki Okubo, Tomitsugu Taguchi, Hiroyasu Tanigawa
The R&D on SiC/SiC composites under the broader approach (BA) activities between Japan and the EU for fusion DEMO developed a fundamental database of mechanical (Task-1) and physical/chemical (Task-2) properties, with a primary target of the application of SiC/SiC composites as functional structure to be used in the dual coolant breeding blanket concept. This paper aims to summarize previous 10-years activities of the R&D of Japan and to provide the key deliverables toward the DEMO design. In Task-1, good creep and fatigue durability were first demonstrated. Besides, in-plane and inter-laminar strength anisotropy maps at elevated temperatures were comprehensively identified. In parallel, the irradiation effects of SiC materials were specifically determined as input parameters of the analytical model to provide for the irradiation-induced residual stresses. In Task-2, the apparent dose-dependence of the radiation-induced electrical conductivity and the indicative radiation-induced electrical degradation was identified by various irradiation sources. In addition, good gas confinement was identified. Furthermore, no accelerated corrosion for duration of 3000 h at below 1173 K was first demonstrated. With these achievements, it is suggested that the in-vessel component technology, e.g., material corrosion database development, activated corrosion product evaluation code development, compact module tests for validation of the key functions of the components, technology integration assessment for fusion nuclear tests, etc., should be further developed toward DEMO in near-term.
Modeling porosity migration in LWR and fast reactor MOX fuel using the finite element method J. Nucl. Mater. (IF 2.048) Pub Date : 2018-05-18 Stephen Novascone, Pavel Medvedev, John W. Peterson, Yongfeng Zhang, Jason Hales
An engineering-scale finite element simulation of pore migration in oxide fuel is presented. The porosity field is governed by an advection-diffusion equation which is coupled to the fuel temperature and stress fields through the thermal conductivity and volumetric heat source term. The engineering-scale porosity equation models the microscopic process of vapor transport of fuel across pores, taking into account thermal and vapor pressure gradients within the fuel. In the simulations, the porosity is initialized to a constant value at every point in the domain, and as the temperature gradient is increased by application of a heat source, the pores move up the thermal gradient and accumulate at the center of the fuel in a time frame that is consistent with experimental observations. Results from representative simulations are provided to demonstrate the new capability, and we show that a sufficiently high power ramp rate limits restructuring and leads to a corresponding increase in fuel temperature. We also discuss the finite element mesh density required to compute pore migration and present multidimensional results.
Some contents have been Reproduced by permission of The Royal Society of Chemistry.
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