U3Si2 behavior in H2O: Part I, flowing steam and the effect of hydrogen J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-17 E. Sooby Wood, J.T. White, C.J. Grote, A.T. Nelson
Recent interest in U3Si2 as an advanced light water reactor fuel has driven assessment of numerous properties, but characterization of its response to H2O environments is absent from the literature. The behavior of U3Si2 in H2O containing atmospheres is investigated and presented in a two-part series of articles aimed to understand the degradation mechanism of U3Si2 in H2O. Reported here are thermogravimetric data for U3Si2 exposed to flowing steam at 250–470 °C. Additionally the response of U3Si2 to flowing Ar-6% H2 from 350 to 400 °C is presented. Microstructural degradation is observed following hours of exposure at 350 °C in steam. U3Si2 undergoes pulverization on the timescale of minutes when temperatures are increased above 400 °C. This mechanism is accelerated in flowing Ar-H2 at the same temperatures.
Extended study on oxidation behaviors of UN0.68 and UN1.66 by XPS J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-17 Lizhu Luo, Yin Hu, Qifa Pan, Zhong Long, Lei Lu, Kezhao Liu, Xiaolin Wang
The surface oxidation behaviors of UN0.68 and UN1.66 thin films are investigated by X-ray photoelectron spectroscopy (XPS), and the traditional U4f/N1s, O1s, valence band spectra as well as the unconventional U4d and U5d spectra are collected for the understanding of their oxidation behavior in-depth. Similar asymmetrical peak shape of the U4f spectra to uranium is observed for both uranium nitrides, despite of the slight shift to higher energy side for UN1.66 clean surface. However, significant difference among the corresponding spectra of UN0.68 and UN1.66 during oxidation reveals the distinctive properties of each own. The coexistence of UO2-x, UO2 and UO2-x.Ny on UN0.68 surface results in the peculiar features of U4f spectra as well as the others within the XPS energy scale, where peaks of the oxidized species firstly shift to higher energy side compared to the clean surface, and then return towards those of stoichiometric UO2. For UN1.66, the generation of U-N-O ternary compounds on the surface is identified with the symmetrical U4f peaks at 379.9eV and 390.8 eV, which locate intermediate between UO2 and UN1.66, and gradually expanding to higher energy side during oxidation. Furthermore, the formation of N-O species on UN1.66 surface is also detected as an oxidation product. The metallic character of UN1.66 is identified by the intense signals at Fermi level, which is greatly suppressed by the increasing oxygen exposure and implies the weakening metallic properties of the U-N-O compounds. A multistage mechanism for UN1.66 following the exposure to oxygen is discussed.
A study on recovery of uranium in the anode basket residues delivered from the pyrochemical process of used nuclear fuel J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-17 H.C. Eun, T.J. Kim, J.H. Jang, G.Y. Kim, S.B. Park, D.S. Yoon, S.H. Kim, S.W. Paek, S.J. Lee
In this study, the chlorination of uranium oxide (UO2) using ammonium chloride and zirconium as chemical agents was conducted to recover the uranium in the anode basket residues from the pyrochemical process of used nuclear fuel. The chlorination of UO2 was predicted using thermodynamic equilibrium calculations. The experimental conditions for the chlorination were determined using a chlorination test with cerium oxide (CeO2). In the chlorination test, it was confirmed that UO2 was chlorinated into UCl3 at 320 °C, some UO2 remained without changes in the chemical form, and ZrO2, Zr2O, and ZrCl2 were generated as byproducts.
Manufacturing and characterization of Ni-free N-containing ODS austenitic alloys J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-17 A. Mori, H. Mamiya, M. Ohnuma, J. Ilavsky, K. Ohishi, J. Wozniak, A. Olszyna, N. Watanabe, J. Suzuki, H. Kitazawa, M. Lewandowska
Ni-free N-containing oxide dispersion strengthened (ODS) austenitic alloys were manufactured by mechanical alloying (MA) followed by spark plasma sintering (SPS). The phase evolutions during milling under a nitrogen atmosphere and after sintering were studied by X-ray diffraction (XRD). Transmission electron microcopy (TEM) and alloy contrast variation analysis (ACV), including small-angle neutron scattering (SANS) and ultra-small-angle X-ray scattering (USAXS), revealed the existence of nanoparticles with a diameter of 3–51 nm for the samples sintered at 950 °C. Sintering at 1000 °C for 5 and 15 min caused slight growth and a significant coarsening of the nanoparticles, up to 70 nm and 128 nm, respectively. The ACV analysis indicated the existence of two populations of Y2O3, ε-martensite and MnO. The dispersive X-ray spectrometry (EDS) confirmed two kinds of nanoparticles, Y2O3 and MnO. The material was characterized by superior micro-hardness, of above 500 HV0.1.
Effect of milling time and annealing temperature on nanoparticles evolution for 13.5% Cr ODS ferritic steels by joint application of XAFS and TEM J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-16 P. He, J. Hoffmann, A. Möslang
The characteristics of strengthening nanoparticles have a major influence on the mechanical property and irradiation resistance of oxide dispersion strengthened (ODS) steels. To determine how to control nanoparticles evolution, 0.3% Ti with 0.3% Y2O3 are added in 13.5%Cr pre-alloyed steel powders via different milling and consolidation conditions, then characterized by transmission electron microscopy (TEM) and X-ray absorption fine structure (XAFS) at synchrotron irradiation facility. The dissolution of Y2O3 is greatly dependent on the milling time at fixed milling speeds. After 24 h of milling, only minor amounts of the initially added Y2O3 dissolves into the steel matrix whereas TEM results reveals nearly complete dissolution of Y2O3 in 80-h-milled powder. The annealed powder FT-A800 (at 800 °C for 1 h) exhibit a structure near to the initially added Y2O3. The slightly deviation may be accounted for considerable lattice distortion related to the presence of atomic vacancies or formation of Y-Ti-O nucleus. The annealed powders FT-A1000 and FT-A1100 contain complex mixtures of Y-O/Y-Ti-O oxides, which cannot be fitted by any single thermally stable compounds. The coordination numbers of these first two shells in the annealed powders significantly raise as a function of the annealing temperature, indicating the formation of more ordered Y-O or Y-Ti-O particles. The EXAFS spectrum could not necessarily distinguish the dominant oxide species.
Accident tolerant fuel cladding development: Promise, status, and challenges ☆ J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-13 Kurt Terrani
The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber–reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.
The effects of crystallization and residual glass on the chemical durability of iron phosphate waste forms containing 40 wt% of a high MoO3 Collins-CLT waste J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-12 Jen-Hsien Hsu, Jincheng Bai, Cheol-Woon Kim, Richard K. Brow, Joe Szabo, Adam Zervos
The effects of cooling rate on the chemical durability of iron phosphate waste forms containing up to 40 wt% of a high MoO3 Collins-CLT waste simulant were determined at 90 °C using the product consistency test (PCT). The waste form, designated 40wt%-5, meets appropriate Department of Energy (DOE) standards when rapidly quenched from the melt (as-cast) and after slow cooling following the CCC (canister centerline cooling)-protocol, although the quenched glass is more durable. The analysis of samples from the vapor hydration test (VHT) and the aqueous corrosion test (differential recession test) reveals that rare earth orthophosphate (monazite) and Zr-pyrophosphate crystals that form on cooling are more durable than the residual glass in the 40wt%-5 waste form. The residual glass in the CCC-treated samples has a greater average phosphate chain length and a lower Fe/P ratio, and those contribute to its faster corrosion kinetics.
Helium bubble evolution and hardening in 316L by post-implantation annealing J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-11 I. Villacampa, J.C. Chen, P. Spätig, H.P. Seifert, F. Duval
The effect of annealing temperature on the helium (He) bubble evolution after He implantation and their hardening effect were systematically studied by transmission electron microscopy (TEM) and tensile tests. A plate made of 316L austenitic stainless steel was implanted with 1000 appm He at 300 °C and then post-implantation annealed from 650 to 1000 °C for 1h. The TEM investigations showed that increasing the annealing temperature, the average bubble size increased and the density decreased. The average He bubble size and distribution in the grain interior and on the grain boundary (GB) were similar in the range of temperatures studied. In both cases, bubbles grew by the Ostwald ripening mechanism. The results suggest that the coarsening mechanism might also depend on the initial bubble size rather than only on the annealing temperature. No preferential He build-up takes place in the GBs. The increase of yield stress at 0.2% strain produced by He bubbles was calculated and compared to tensile test results. The hardening coefficients obtained using distinct models are consistent with simulations in published data.
Effect of the oxidation front penetration on in-clad hydrogen migration J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-11 F. Feria, L.E. Herranz
In LWR fuel claddings the embrittlement due to hydrogen precipitates (i.e., hydrides) is a degrading mechanism that concerns in nuclear safety, particularly in dry storage. A relevant factor is the radial distribution of the hydrogen absorbed, especially the hydride rim formed. Thus, a reliable assessment of fuel performance should account for hydrogen migration. Based on the current state of modelling of hydrogen dynamics in the cladding, a 1D radial model has been derived and coupled with the FRAPCON code. The model includes the effect of the oxidation front progression on in-clad hydrogen migration, based on experimental observations found (i.e., dissolution/diffusion/re-precipitation of the hydrogen in the matrix ahead of the oxidation front). A remarkable quantitative impact of this new contribution has been shown by analyzing the hydrogen profile across the cladding of several high burnup fuel scenarios (>60 GW d/tU); other potential contributions like thermodiffusion and diffusion in the hydride phase hardly make any difference. Comparisons against PIE measurements allow concluding that the model accuracy notably increases when the effect of the oxidation front is accounted for in the hydride rim formation. In spite of the promising results, further validation would be needed.
Molten salt corrosion behaviour of structural materials in LiCl-KCl-UCl3 by thermogravimetric study J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-11 Ch Jagadeeswara Rao, S. Ningshen, C. Mallika, U. Kamachi Mudali
The corrosion resistance of structural materials has been recognized as a key issue in the various unit operations such as salt purification, electrorefining, cathode processing and injection casting in the pyrochemical reprocessing of spent metallic nuclear fuels. In the present work, the corrosion behaviour of the candidate materials SS410, 2.25Cr-1Mo and 9Cr-1Mo steels was investigated in molten LiCl-KCl-UCl3 salt by thermogravimetric analysis under inert and reactive atmospheres at 500 and 600 °C, for 6 h duration. Insignificant weight gain in the inert atmosphere and marginal weight gain in the reactive atmosphere were observed at both the temperatures. Chromium depletion rates and formation of Cr-rich corrosion products increased with increasing temperature of exposure in both inert and reactive atmospheres as evidenced by SEM and EDS analysis. The corrosion attack by LiCl-KCl-UCl3 molten salt, under reactive atmosphere for 6 h duration was more in the case of SS410 than 9Cr-1Mo steel followed by 2.25Cr-1Mo steel at 500 °C and the corrosion attack at 600 °C followed the order: 9Cr-1Mo steel >2.25Cr-1Mo steel > SS410. Outward diffusion of the minor alloying element, Mo was observed in 9Cr-1Mo and 2.25Cr-1Mo steels at both temperatures under reactive atmosphere. Laser Raman spectral analysis of the molten salt corrosion tested alloys under a reactive atmosphere at 500 and 600 °C for 6 h revealed the formation of unprotected Fe3O4 and α-as well as γ-Fe2O3. The results of the present study facilitate the selection of structural materials for applications in the corrosive molten salt environment at high temperatures.
Thermal diffusivity and conductivity of thorium- uranium mixed oxides J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-11 M. Saoudi, D. Staicu, J. Mouris, A. Bergeron, H. Hamilton, M. Naji, D. Freis, M. Cologna
Thorium-uranium oxide pellets with high densities were prepared at the Canadian Nuclear Laboratories (CNL) by co-milling, pressing, and sintering at 2023 K, with UO2 mass contents of 0, 1.5, 3, 8, 13, 30, 60 and 100%. At the Joint Research Centre, Karlsruhe (JRC-Karlsruhe), thorium-uranium oxide pellets were prepared using the spark plasma sintering (SPS) technique with 79 and 93 wt. % UO2. The thermal diffusivity of (Th1-xUx)O2 (0 ≤ x ≤ 1) was measured at CNL and at JRC-Karlsruhe using the laser flash technique. ThO2 and (Th,U)O2 with 1.5, 3, 8 and 13 wt. % UO2 were found to be semi-transparent to the infrared wavelength of the laser and were coated with graphite for the thermal diffusivity measurements. This semi-transparency decreased with the addition of UO2 and was lost at about 30 wt. % of UO2 in ThO2. The thermal conductivity was deduced using the measured density and literature data for the specific heat capacity. The thermal conductivity for ThO2 is significantly higher than for UO2. The thermal conductivity of (Th,U)O2 decreases rapidly with increasing UO2 content, and for UO2 contents of 60% and higher, the conductivity of the thorium-uranium oxide fuel is close to UO2. As the mass difference between the Th and U atoms is small, the thermal conductivity decrease is attributed to the phonon scattering enhanced by lattice strain due to the introduction of uranium in ThO2 lattice. The new results were compared to the data available in the literature and were evaluated using the classical phonon transport model for oxide systems.
Thermochemical effect of fission products on sodium – MOX fuel reaction: The case of niobium J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-10 Dan T. Costin, Lionel Desgranges, Victor Cabello-Ortiga, Marcus Hedberg, Jenny Halleröd, Teodora Retegan, Christian Ekberg
The influence of niobium on the sodium MOX fuel chemical interaction was studied by different heat treatments of airtight capsules containing fresh MOX, sodium and a niobium strip. The characterisation results evidenced a two-step process with first MOX oxidation and then MOX reduction. This result was interpreted by considering the formation of sodium niobiate that captures oxygen from the MOX. This interpretation is used to discuss the influence of niobium as fission product on the sodium –irradiated MOX fuel reaction.
Development of low-Cr ODS FeCrAl alloys for accident-tolerant fuel cladding ☆ J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-10 Sebastien Dryepondt, Kinga A. Unocic, David T. Hoelzer, Caleb P. Massey, Bruce A. Pint
Low-Cr oxide dispersion strengthened (ODS) FeCrAl alloys were developed as accident tolerant fuel cladding because of their excellent oxidation resistance at very high temperature, high strength and improved radiation tolerance. Fe-12 Cr-5Al wt.% gas atomized powder was ball milled with Y2O3+FeO, Y2O3+ZrO2 or Y2O3+TiO2, and the resulting powders were extruded at 950 °C. The resulting fine grain structure, particularly for the Ti and Zr containing alloys, led to very high strength but limited ductility. Comparison with variants of commercial PM2000 (Fe-20Cr-5Al) highlighted the significant impact of the powder consolidation step on the alloy grain size and, therefore, on the alloy mechanical properties at T < 500 °C. These low-Cr compositions exhibited good oxidation resistance at 1400 °C in air and steam for 4 h but could not form a protective alumina scale at 1450 °C, similar to observations for fine grained PM2000 alloys. The effect of alloy grain size, Zr and Ti additions, and impurities on the alloy mechanical and oxidation behaviors are discussed.
Characterization of ion irradiation effects on the microstructure, hardness, deformation and crack initiation behavior of austenitic stainless steel:Heavy ions vs protons J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-10 J. Gupta, J. Hure, B. Tanguy, L. Laffont, M.-C. Lafont, E. Andrieu
Irradiation Assisted Stress Corrosion Cracking (IASCC) is a complex phenomenon of degradation which can have a significant influence on maintenance time and cost of core internals of a Pressurized Water Reactor (PWR). Hence, it is an issue of concern, especially in the context of lifetime extension of PWRs. Proton irradiation is generally used as a representative alternative of neutron irradiation to improve the current understanding of the mechanisms involved in IASCC. This study assesses the possibility of using heavy ions irradiation to evaluate IASCC mechanisms by comparing the irradiation induced modifications (in microstructure and mechanical properties) and cracking susceptibility of SA 304 L after both type of irradiations: Fe irradiation at 450 °C and proton irradiation at 350 °C. Irradiation-induced defects are characterized and quantified along with nano-hardness measurements, showing a correlation between irradiation hardening and density of Frank loops that is well captured by Orowan's formula. Both irradiations (iron and proton) increase the susceptibility of SA 304 L to intergranular cracking on subjection to Constant Extension Rate Tensile tests (CERT) in simulated nominal PWR primary water environment at 340 °C. For these conditions, cracking susceptibility is found to be quantitatively similar for both irradiations, despite significant differences in hardening and degree of localization.
Evolution of ion damage at 773K in Ni- containing concentrated solid-solution alloys J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-10 Shi Shi, Mo-Rigen He, Ke Jin, Hongbin Bei, Ian M. Robertson
Quantitative analysis of the impact of the compositional complexity in a series of Ni-containing concentrated solid-solution alloys, Ni, NiCo, NiFe, NiCoCr, NiCoFeCr, NiCoFeCrMn and NiCoFeCrPd, on the evolution of defects produced by 1 MeV Kr ion irradiation at 773 K is reported. The dynamics of the evolution of the damage structure during irradiation to a dose of 2 displacements per atom were observed directly by performing the ion irradiations in electron transparent foils in a transmission electron microscope coupled to an ion accelerator. The defect evolution was assessed through measurement of the defect density, defect size and fraction of perfect and Frank loops. These three parameters were dependent on the alloying element as well as the number of elements. The population of loops was sensitive to the ion dose and alloy composition as faulted Frank loops were observed to unfault to perfect loops with increasing ion dose. These dependences are explained in terms of the influence of each element on the lifetime of the displacement cascade as well as on defect formation and migration energies.
An improved model of fission gas atom transport in irradiated uranium dioxide J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-09 J.H. Shea
The hitherto standard approach to predicting fission gas release has been a pure diffusion gas atom transport model based upon Fick's law. An additional mechanism has subsequently been identified from experimental data at high burnup and has been summarised in an empirical model that is considered to embody a so-called fuel matrix ’saturation’ phenomenon whereby the fuel matrix has become saturated with fission gas so that the continued addition of extra fission gas atoms results in their expulsion from the fuel matrix into the fuel rod plenum.The present paper proposes a different approach by constructing an enhanced fission gas transport law consisting of two components: 1) Fick's law and 2) a so-called drift term. The new transport law can be shown to be effectively identical in its predictions to the ’saturation’ approach and is more readily physically justifiable. The method introduces a generalisation of the standard diffusion equation which is dubbed the Drift Diffusion Equation. According to the magnitude of a dimensionless Péclet number, P, the new equation can vary from pure diffusion to pure drift, which latter represents a collective motion of the fission gas atoms through the fuel matrix at a translational velocity.Comparison is made between the saturation and enhanced transport approaches. Because of its dependence on P, the Drift Diffusion Equation is shown to be more effective at managing the transition from one type of limiting transport phenomenon to the other. Thus it can adapt appropriately according to the reactor operation.
Effects of Ti and Ta addition on microstructure stability and tensile properties of reduced activation ferritic/martensitic steel for nuclear fusion reactors J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-08 Han Kyu Kim, Ji Won Lee, Joonoh Moon, Chang Hoon Lee, Hyun Uk Hong
The effects of Ti and Ta addition on microstructure stability and tensile properties of a reduced activation ferritic/martensitic (RAFM) steel have been investigated. Ti addition of 0.06 wt% to conventional RAFM reference base steel (Fe-9.3Cr-0.93W-0.22V-0.094Ta-0.1C) was intended to promote the precipitation of nano-sized (Ti,W) carbides with a high resistance to coarsening. In addition, the Ti addition was substituted for 0.094 wt% Ta. The Ti-added RAFM steel (Ti-RAFM) exhibited a higher yield strength (ΔYS = 32 MPa) at 600 °C than the reference base steel due to additional precipitation hardening by (Ti,W)-rich MX with an average size of 6.1 nm and the area fraction of 2.39%. However, after thermal exposure at 600 °C for 1000 h, this Ti-RAFM was more susceptible to degradation than the reference base steel; the block width increased by 77.6% in Ti-RAFM after thermal exposure while the reference base steel showed only 9.1% increase. In order to suppress diffusion rate during thermal exposure, the large-sized Ta element with low activation was added to Ti-RAFM. The Ta-added Ti-RAFM steel exhibited good properties with outstanding microstructure stability. Quantitative comparison in microstructures was discussed with a consideration of Ti and Ta addition.
Strength of SiCf-SiCm composite tube under uniaxial and multiaxial loading J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-06 Kirill Shapovalov, George M. Jacobsen, Luis Alva, Nathaniel Truesdale, Christian P. Deck, Xinyu Huang
The authors report mechanical strength of nuclear grade silicon carbide fiber reinforced silicon carbide matrix composite (SiCf-SiCm) tubing under several different stress states. The composite tubing was fabricated via a Chemical Vapor Infiltration (CVI) process, and is being evaluated for accident tolerant nuclear fuel cladding. Several experimental techniques were applied including uniaxial tension, elastomer insert burst test, open and closed end hydraulic bladder burst test, and torsion test. These tests provided critical stress and strain values at proportional limit and at ultimate failure points. Full field strain measurements using digital image correlation (DIC) were obtained in order to acquire quantitative information on localized deformation during application of stress. Based on the test results, a failure map was constructed for the SiCf-SiCm composites.
A promising tritium breeding material: Nanostructured 2Li2TiO3–Li4SiO4 biphasic ceramic pebbles J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-06 Chen Dang, Mao Yang, Yichao Gong, Lan Feng, Hailiang Wang, Yanli Shi, Qiwu Shi, Jianqi Qi, Tiecheng Lu
As an advanced tritium breeder material for the fusion reactor blanket of the International Thermonuclear Experimental Reactor (ITER), Li2TiO3-Li4SiO4 biphasic ceramic has attracted widely attention due to its merits. In this paper, the uniform precursor powders were prepared by hydrothermal method, and nanostructured 2Li2TiO3-Li4SiO4 biphasic ceramic pebbles were fabricated by an indirect wet method at the first time. In addition, the composition dependence (x/y) of their microstructure characteristics and mechanical properties were investigated. The results indicated that the crush load of biphasic ceramic pebbles was better than that of single phase ceramic pebbles under identical conditions. The 2Li2TiO3-Li4SiO4 ceramic pebbles have good morphology, small grain size (90 nm), satisfactory crush load (37.8 N) and relative density (81.8 %T.D.), which could be a promising breeding material in the future fusion reactor.
Suppression of deuterium-induced blistering in pre-damaged tungsten exposed to short-duration deuterium plasma J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-06 Xiu-Li Zhu, Ying Zhang, Long Cheng, Li-Qun Shi, Gregory De Temmerman, Yue Yuan, Hui-Ping Liu, Guang-Hong Lu
Effects of pre-damage by 500 keV argon ion implantation on deuterium-induced blistering in tungsten has been investigated. After low-energy (40 eV) and high-flux (∼1024 D/m2s) deuterium plasma exposure with short exposure duration (100 s), a large increase of deuterium retention is found in the pre-damaged tungsten, while surface blistering is significantly suppressed as compared to the un-damaged one. According to elastic recoil detection analysis, a local deuterium concentration peak is observed at a depth of ∼100 nm for the un-damaged tungsten, which is supposed to be related to the surface blistering with nanometer size. Comparison of deuterium retention in the near surface (within 300 nm) and in the bulk suggests that deuterium inward diffusion is more significant in the pre-damaged tungsten. It is speculated that the creation of deuterium trap-sites and enhancement of deuterium inward diffusion give rise to an increase of critical deuterium concentration for blistering and contribute to the suppressed deuterium-induced blistering on pre-damaged tungsten under the present exposure conditions.
Hydrogen migration modeling in a symmetric tilt boundary of the Iron-Chromium system J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-06 V.P. Ramunni
Previous experimental studies of H permeation in 9%Cr-Fe alloys have found a permeation coefficient 10 times lower and a diffusion coefficient 200 times lower than in pure annealed Fe. In an effort to shed some light on the microscopic origin of these findings, we perform an extensive study of Fe, Cr, and H migration in a high-angle symmetric tilt grain boundary in bcc Fe, both via vacancy and interstitial mechanism. This is undertaken in the framework of transition state theory with the relevant energies obtained from classical interatomic potentials, and partially from Density Functional Theory calculations, in order to check the consistency of structures. Trapping sites for H and possible migration paths are explored. We find that the presence of Cr and its migration via vacancy and interstitials creates the conditions in produce stable preferential trapping sites for H in the grain boundary, that delay the H migration, thereby explaining the experimental results.
Blistering behavior and deuterium retention in tungsten vanadium alloys exposed to deuterium plasma in the linear plasma device STEP J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-05 Jun Wang, Long Cheng, Yue Yuan, Shao-Yang Qin, Kameel Arshad, Wang-Guo Guo, Zheng Wang, Zhang-Jian Zhou, Guang-Hong Lu
The behavior of tungsten-vanadium (W-V) alloys fabricated by powder metallurgy as a plasma facing material has been studied. W-V alloys with different vanadium concentrations (5 and 10 wt %) manufactured by hot pressing (HP) were exposed to deuterium plasma (flux ∼4.6 × 1021 m−2s−1, fluence ∼5.6 × 1025 m−2, ion energy ∼60 eV, target temperature ∼450 K) in the linear plasma device STEP at Beihang University. Three typical grains are observed on HP sintered W-V alloys and exhibit a significant effect on its performance under deuterium plasma irradiation. Surface blistering only occurs at W-enriched grains and is significantly mitigated in W-V alloys, especially in W-10 V, blistering is completely suppressed. On the other hand, deuterium retention dramatically increases in the W-V alloys due to vanadium addition. The deuterium retention in W-5 wt. % V is about 6.2 times more than that in rolled pure W, and this factor further increases to 6.9 when the V concentration rises to 10 wt %. We ascribe these phenomena to the changes of microstructures and components caused by vanadium addition.
Properties of the LiCl-KCl-Li2O system as operating medium for pyro-chemical reprocessing of spent nuclear fuel J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-04 Albert Mullabaev, Olga Tkacheva, Vladimir Shishkin, Vadim Kovrov, Yuriy Zaikov, Leonid Sukhanov, Yuriy Mochalov
Crystallization temperatures (liquidus and solidus) in the LiCl-Li2O and (LiCl-KCl)-Li2O systems with the KCl content of 10 and 20 mol.% were obtained with independent methods of thermal analysis using cooling curves, isothermal saturation, and differential scanning calorimetry. The linear sweep voltammetry was applied to control the time of the equilibrium establishment in the molten system after the Li2O addition, which depended on the composition of the base melt and the concentration of Li2O. The fragments of the binary LiCl-Li2O and quazi-binary [LiCl-KCl(10 mol.%)]-Li2O and [LiCl-KCl(20 mol.%)]-Li2O phase diagrams in the Li2O concentration range from 0 to 12 mol.% were obtained. The KCl presence in the LiCl-KCl-Li2O molten mixture in the amount of 10 and 20 mol.% reduces the liquidus temperature by 30 and 80°, respectively, but the region of the homogeneous molten state of the system is considerably narrowed, which complicates its practical application. The Li2O solubility in the molten LiCl, LiCl-KCl(10 mol.%) and LiCl-KCl(20 mol.%) decreases with increasing the KCl content and is equal to 11.5, 7.7 and 3.9 mol.% at 650°С, respectively. The LiCl-KCl melt with 10 mol.% KCl can be recommended for practical use as a medium for the SNF pyro-chemical reprocessing at temperature below 700 °C.
The effect of thermomechanical processing on second phase particle redistribution in U-10 wt%Mo J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-04 Xiaohua Hu, Xiaowo Wang, Vineet V. Joshi, Curt A. Lavender
Microstructural evolution of nanochannel CrN films under ion irradiation at elevated temperature and post-irradiation annealing J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-02 Jun Tang, Mengqing Hong, Yongqiang Wang, Wenjing Qin, Feng Ren, Lan Dong, Hui Wang, Lulu Hu, Guangxu Cai, Changzhong Jiang
High-performance radiation tolerance materials are crucial for the success of future advanced nuclear reactors. In this paper, we present a further investigation that the “vein-like” nanochannel films can enhance radiation tolerance under ion irradiation at high temperature and post-irradiation annealing. The chromium nitride (CrN) nanochannel films with different nanochannel densities and the compact CrN film are chosen as a model system for these studies. Microstructural evolution of these films were investigated using Transmission Electron Microscopy (TEM), Scanning Electron Microscopy (SEM), Elastic Recoil Detection (ERD) and Grazing Incidence X-ray Diffraction (GIXRD). Under the high fluence He+ ion irradiation at 500 °C, small He bubbles with low bubble densities are observed in the irradiated nanochannel CrN films, while the aligned large He bubbles, blistering and texture reconstruction are found in the irradiated compact CrN film. For the heavy Ar2+ ion irradiation at 500 °C, the microstructure of the nanochannel CrN RT film is more stable than that of the compact CrN film due to the effective releasing of defects via the nanochannel structure. Under the He+ ion irradiation and subsequent annealing, compared with the compact film, the nanochannel films have excellent performance for the suppression of He bubble growth and possess the strong microstructural stability. Basing on the analysis on the sizes and number densities of bubbles as well as the concentrations of He retained in the nanochannel CrN films and the compact CrN film under different experimental conditions, potential mechanism for the enhanced radiation tolerance are discussed. Nanochannels play a crucial role on the release of He/defects under ion irradiation. We conclude that the tailored “vein-like” nanochannel structure may be used as advanced radiation tolerance materials for future nuclear reactors.
Comparative study of He bubble formation in nanostructured reduced activation steel and its coarsen-grained counterpart J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-02 W.B. Liu, J.H. Zhang, Y.Z. Ji, L.D. Xia, H.P. Liu, D. Yun, C.H. He, C. Zhang, Z.G. Yang
On the possibility of universal interstitial emission induced annihilation in metallic nanostructures J. Nucl. Mater. (IF 2.048) Pub Date : 2018-01-02 Xiangyan Li, Yichun Xu, Guohua Duan, Jingjing Sun, Congyu Hao, Yange Zhang, Wei Liu, C.S. Liu, Q.F. Fang
Sodium aluminum-iron phosphate glass-ceramics for immobilization of lanthanide oxide wastes from pyrochemical reprocessing of spent nuclear fuel J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-30 S.V. Stefanovsky, O.I. Stefanovsky, M.I. Kadyko, B.S. Nikonov
Sodium aluminum (iron) phosphate glass ceramics containing of up to 20 wt.% rare earth (RE) oxides simulating pyroprocessing waste were produced by melting at 1250 °C followed by either quenching or slow cooling to room temperature. The iron-free glass-ceramics were composed of major glass and minor phosphotridymite and monazite. The iron-bearing glass-ceramics were composed of major glass and minor monazite and Na-Al-Fe orthophosphate at low waste loadings (5–10 wt.%) and major orthophosphate and minor monazite as well as interstitial glass at high waste loadings (15–20 wt.%). Slowly cooled samples contained higher amount of crystalline phases than quenched ones. Monazite is major phase for REs. Leach rates from the materials of major elements (Na, Al, Fe, P) are 10−5-10−7 g cm−2 d−1, RE elements – lower than 10−5 g cm−2 d−1.
Structural defect accumulation in tungsten and tungsten-5wt.% tantalum under incremental proton damage J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-29 I. Ipatova, R.W. Harrison, P.T. Wady, S.M. Shubeita, D. Terentyev, S.E. Donnelly, E. Jimenez-Melero
Thermodynamic assessment of the rhodium-ruthenium-oxygen (Rh-Ru-O) system J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-28 S. Gossé, S. Bordier, C. Guéneau, E. Brackx, R. Domenger, J. Rogez
Ruthenium (Ru) and rhodium (Rh) are abundant platinum-group metals formed during burn-up of nuclear fuels. Under normal operating conditions, Rh and Ru accumulate and predominantly form metallic precipitates with other fission products like Mo, Pd and Tc. In the framework of vitrification of high-level nuclear waste, these fission products are poorly soluble in molten glasses. They precipitate as metallic particles and oxide phases. Moreover, these Ru and Rh rich phases strongly depend on temperature and the oxygen fugacity of the glass melt. In case of severe accidental conditions with air ingress, oxidation of the Ru and Rh is possible. At low temperatures (T < 1422 K for rhodium sesquioxide and T < 1815 K for ruthenium dioxide), the formed oxides are relatively stable. On the other hand, at high temperatures (T > 1422 K for rhodium sesquioxide and T > 1815 K for ruthenium dioxide), they may decompose into (Rh)-FCC or (Ru)-HCP metallic phases and radiotoxic volatile gaseous species. A thermodynamic assessment of the Rh-Ru-O system will enable the prediction of: (1) the metallic and oxide phases that form during the vitrification of high-level nuclear wastes and (2) the release of volatile gaseous species during a severe accident. The Calphad method developed herein employs a thermodynamic approach in the investigation of the thermochemistry of rhodium and ruthenium at high temperatures. Current literature on the thermodynamic properties and phase diagram data enables preliminary thermodynamic assessments of the Rh-O and Ru-O systems. Additionally, select point in the ternary Rh-Ru-O system underwent experimental tests to complement data found in literature and to establish the phase equilibria in the ternary system.
Grain-boundary type and distribution in silicon carbide coatings and wafers J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-28 Felix Cancino-Trejo, Eddie López-Honorato, Ross C. Walker, Romelia Salomon Ferrer
Silicon carbide is the main diffusion barrier against metallic fission products in TRISO (tristructural isotropic) coated fuel particles. The explanation of the accelerated diffusion of silver through SiC has remained a challenge for more than four decades. Although, it is now well accepted that silver diffuse through SiC by grain boundary diffusion, little is known about the characteristics of the grain boundaries in SiC and how these change depending on the type of sample. In this work five different types (coatings and wafers) of SiC produced by chemical vapor deposition were characterized by electron backscatter diffraction (EBSD). The SiC in TRISO particles had a higher concentration of high angle grain boundaries (aprox. 70%) compared to SiC wafers, which ranged between 30 and 60%. Similarly, SiC wafers had a higher concentration of low angle grain boundaries ranging between 15 and 30%, whereas TRISO particles only reached values of around 7%. The same trend remained when comparing the content of coincidence site lattice (CSL) boundaries, since SiC wafers showed a concentration of more than 30%, whilst TRISO particles had contents of around 20%. In all samples the largest fractions of CSL boundaries (3 ≤ Σ ≤ 17) were the Σ3 boundaries. We show that there are important differences between the SiC in TRISO particles and SiC wafers which could explain some of the differences observed in diffusion experiments in the literature.
Reply to “On Vaporization of liquid Pb-Li eutectic alloy from 1000 K to 1200 K- A high temperature mass spectrometric study” J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-28 Uttam Jain, Abhishek Mukherjee
This communication is in response to a letter to editor commenting on the authors' earlier paper “Vaporization of liquid Pb-Li eutectic alloy from 1000 K to 1200 K - A high temperature mass spectrometric study”.
Oxidation of 316L(N) stainless steel in liquid sodium at 650 °C J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-28 Matthieu Rivollier, Jean-Louis Courouau, Michel Tabarant, Cécile Blanc, Marie-Laurence Giorgi
On α′ precipitate composition in thermally annealed and neutron-irradiated Fe- 9-18Cr alloys J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-28 Elaina R. Reese, Mukesh Bachhav, Peter Wells, Takuya Yamamoto, G. Robert Odette, Emmanuelle A. Marquis
Temperature dependent elastic properties of γ-phase U – 8 wt% Mo J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-28 M.A. Steiner, E. Garlea, J. Creasy, A. DeMint, S.R. Agnew
Polycrystalline elastic moduli and stiffness tensor components of γ-phase U – 8 wt% Mo have been determined by resonant ultrasound spectroscopy in the temperature range of 25–650 °C. The ambient temperature elastic properties are compared to results measured via other experimental methods and show reasonable agreement, though there is considerable variation of these properties within the literature at both the U – 8 wt% Mo composition and as a function of Mo concentration. The Young's modulus of U – 8 wt% Mo measured in this study decreases steadily with temperature at a rate that is slower than trends previously observed at similar Mo concentrations, though the difference is not statistically significant. This first measurement of the temperature dependent elastic stiffness tensor of a polycrystalline U-Mo alloy clarifies that the behavior of the Young's modulus is due to a strongly weakening C11 polycrystalline stiffness tensor component, along with milder decreases in C12 and C44. The unique partially auxetic properties recently predicted for single-crystalline U-Mo are discussed in regard to their possible impact on the polycrystalline behavior of the alloy.
Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-26 Rachel L. Seibert, Kurt A. Terrani, Daniel Velázquez, John D. Hunn, Charles A. Baldwin, Fred C. Montgomery, Jeff Terry
Effect of post-irradiation annealing on the irradiated microstructure of neutron-irradiated 304L stainless steel J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-24 Z. Jiao, J. Hesterberg, G.S. Was
Post-irradiation annealing was performed on a 304L SS that was irradiated to 5.9 dpa in the Barsebäck 1 BWR reactor. Evolution of dislocation loops, radiation-induced solute clusters and radiation-induced segregation at the grain boundary was investigated following thermal annealing at 500 °C and 550 °C up to 20 h. Dislocation loops, Ni-Si and Al-Cu clusters, and enrichment of Ni, Si and depletion of Cr at the grain boundary were observed in the as-irradiated condition. Dislocation loop size did not change significantly after annealing at 550 °C for 5 h but the loop number density decreased considerably and loops mostly disappeared after annealing at 550 °C for 20 h. The average size of Ni-Si and Al-Cu clusters increased while the number density decreased with annealing. The increase in cluster size was due to diffusion of solutes rather than cluster coarsening. Significant volume fractions of Ni-Si and Al-Cu clusters still remained after annealing at 550 °C for 20 h. Substantial recovery of Cr and Ni at the grain boundary was observed after annealing at 550 °C for 5 h but neither Cr nor Ni was fully recovered after 20 h. Annihilation of dislocation loops, driven by the thermal vacancy concentration gradient caused by the strain field and stacking fault associated with the loops appeared to be faster than annihilation of solute clusters and recovery of Ni and Si at the grain boundary, both of which are driven by the solute concentration gradients.
Uranium migration in spark plasma sintered W/UO2 cermets J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-24 Dennis S. Tucker, Yaqiao Wu, Jatuporn Burns
W/UO2 CERMET samples were sintered in a Spark Plasma Sintering (SPS) furnace at various temperature under vacuum and pressure. High Resolution Transmission Electron Microscopy (HRTEM) with Energy Dispersive Spectroscopy (EDS) was performed on the samples to determine interface structures and uranium diffusion from the UO2 particles into the tungsten matrix. Local Electrode Atom Probe (LEAP) was also performed to determine stoichiometry of the UO2 particles. It was seen that uranium diffused approximately 10–15 nm into the tungsten matrix. This is explained in terms of production of oxygen vacancies and Fick's law of diffusion.
Effect of hydrogenation conditions on the microstructure and mechanical properties of zirconium hydride J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-20 Hiroaki Muta, Ryoji Nishikane, Yusuke Ando, Junji Matsunaga, Kan Sakamoto, Stefanus Harjo, Takuro Kawasaki, Yuji Ohishi, Ken Kurosaki, Shinsuke Yamanaka
Single-crystal and polycrystalline diamond erosion studies in Pilot-PSI J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-19 D. Kogut, D. Aussems, N. Ning, K. Bystrov, A. Gicquel, J. Achard, O. Brinza, Y. Addab, C. Martin, C. Pardanaud, S. Khrapak, G. Cartry
Diamond is a promising candidate for enhancing the negative-ion surface production in the ion sources for neutral injection in fusion reactors; hence evaluation of its reactivity towards hydrogen plasma is of high importance. Single crystal and polycrystalline diamond samples were exposed in Pilot-PSI with the D+ flux of (4‒7)·1024 m−2s−1 and the impact energy of 7–9 eV per deuteron at different surface temperatures; under such conditions physical sputtering is negligible, however chemical sputtering is important. Net chemical sputtering yield Y = 9.7·10−3 at/ion at 800 °C was precisely measured ex-situ using a protective platinum mask (5 × 10 × 2 μm) deposited beforehand on a single crystal followed by the post-mortem analysis using Transmission Electron Microscopy (TEM). The structural properties of the exposed diamond surface were analyzed by Raman spectroscopy and X-ray Photoelectron Spectroscopy (XPS). Gross chemical sputtering yields were determined in-situ by means of optical emission spectroscopy of the molecular CH A-X band for several surface temperatures. A bell-shaped dependence of the erosion yield versus temperature between 400 °C and 1200 °C was observed, with a maximum yield of ∼1.5·10−2 at/ion attained at 900 °C. The yields obtained for diamond are relatively high (0.5–1.5)·10−2 at/ion, comparable with those of graphite. XPS analysis shows amorphization of diamond surface within 1 nm depth, in a good agreement with molecular dynamics (MD) simulation. MD was also applied to study the hydrogen impact energy threshold for erosion of  diamond surface at different temperatures.
U3Si2 behavior in H2O environments: Part II, pressurized water with controlled redox chemistry J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-16 A.T. Nelson, A. Migdissov, E. Sooby Wood, C.J. Grote
Recent interest in U3Si2 as an advanced light water reactor fuel has driven assessment of numerous properties, but characterization of its response to H2O environments is sparse in available literature. The behavior of U3Si2 in H2O containing atmospheres is investigated and presented in a two-part series of articles. This work examines the behavior of U3Si2 following exposure to pressurized H2O at temperatures from 300 to 350 °C. Testing was performed using two autoclave configurations and multiple redox conditions. Use of solid state buffers to attain a controlled water chemistry is also presented as a means to test actinide-bearing systems. Buffers were used to vary the hydrogen concentration between 1 and 30 parts per million H2. Testing included UN, U3Si5, and UO2. Both UN and U3Si5 were found to rapidly pulverize in less than 50 h at 300 °C. Uranium dioxide was included as a control for the autoclave system, and was found to be minimally impacted by exposure to pressurized water at the conditions tested for extended time periods. Testing of U3Si2 at 300 °C found reasonable stability through 30 days in 1–5 ppm H2. However, pulverization was observed following 35 days. The redox condition of testing strongly affected pulverization. Characterization of the resulting microstructures suggests that the mechanism responsible for pulverization under more strongly reducing conditions differs from that previously identified. Hydride formation is hypothesized to drive this transition. Testing performed at 350 °C resulted in rapid pulverization of U3Si2 in under 50 h.
Sputtering effects on mirrors made of different tungsten grades J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-16 V.S. Voitsenya, O.V. Ogorodnikova, A.F. Bardamid, V.N. Bondarenko, V.G. Konovalov, P.M. Lytvyn, L. Marot, I.V. Ryzhkov, A.F. Shtan', O.O. Skoryk, S.I. Solodovchenko
Because tungsten (W) is used in present fusion devices and is a reference material for ITER divertor and possible plasma-facing material for DEMO, we strive to understand the response of different W grades to ion bombardment. In this study, we investigated the behavior of mirrors made of four polycrystalline W grades under long-term ion sputtering. Argon (Ar) and deuterium (D) ions extracted from a plasma were used to investigate the effect of projectile mass on surface modification. Depending on the ion fluence, the reflectance measured at normal incidence was very different for different W grades. The lowest degradation rate of the reflectance was measured for the mirror made of recrystallized W. The highest degradation rate was found for one of the ITER-grade W samples. Pre-irradiation of a mirror with 20-MeV W6+ ions, as simulation of neutron irradiation in ITER, had no noticeable influence on reflectance degradation under sputtering with either Ar or D ions.
Radiation-induced amorphization of Langasite La3Ga5SiO14 J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-15 Tiankai Yao, Fengyuan Lu, Haifeng Zhang, Bowen Gong, Wei Ji, Lei Zuo, Jie Lian
Single crystals of Langasite La3Ga5SiO14 (LGS) were irradiated by 1 MeV Kr2+ ions at temperature range from 298 to 898 K in order to simulate the damage effect of neutron radiation on Langasite, a candidate sensor material proposed as high temperature and pressure sensors in nuclear reactors. The microstructure evolution of LGS as functions of irradiation dose and temperature was determined by in-situ TEM observation through electron diffraction pattern. LGS is found to be sensitive to ion beam irradiation-induced amorphization from displacive heavy ions with a low critical dose of ∼0.46 ± 0.18 dpa (neutron fluence of (1.62 ± 0.62) × 1019 neutrons/cm2) at room temperature. The critical amorphization temperature, Tc, is determined to be 909 ± 11.6 K. Under simultaneous ionizing electron (300 keV, 45 nA) and displacive heavy ion irradiations (1-MeV Kr2+ and flux of 6.25 × 1011 ions/cm2·s), LGS displayed greater stability of crystal structure against amorphization, possibly due to the electron radiation-induced recovery of displacive damage by heavy ions.
Numerical modeling of oxygen mass transfer in a wire wrapped fuel assembly under flowing lead bismuth eutectic J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-14 A. Marino, J. Lim, S. Keijers, J. Deconinck, A. Aerts
Corrosion of steels in lead bismuth eutectic (LBE) cooled reactors can be mitigated by forming a protective oxide layer on the steel surfaces. The amount of oxygen necessary to ensure continuous oxide layer formation on fuel cladding depends on the characteristics of the steel and on the local temperature, local oxygen concentration and velocity of the LBE in contact with the steel. The most critical areas from a corrosion point of view are high temperature and low oxygen concentration regions. Wire-wrapped fuel assemblies (FAs) which are foreseen to be used in LBE cooled reactors, are characterized by hot spots and quasi-stagnant areas where oxygen could be depleted. Experimental measurements to verify whether the oxygen concentration in those critical areas is sufficiently elevated for oxide layer formation, are practically impossible. This information can be however obtained by numerical modeling. This paper focuses on the development of a numerical model of oxygen mass transfer in a 19-pin scaled fuel assembly (FA) representative of the MYRRHA reactor core. Oxidation of steels and oxygen transport from the bulk of the LBE to the surface of steels were simulated simultaneously. The simulations provide a local oxygen concentration mapping at steel/LBE interface enabling to identify the regions of the core which could be prone to corrosion due to oxygen depleted LBE. Operation recommendations for the MYRRHA reactor were given based on the simulation results.
Annealing effect on the microstructure of tungsten irradiated in SINQ target J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-14 Barbara Horvath, Yong Dai, Yongjoong Lee
In this work, the microstructure of pure tungsten irradiated in a target of the Swiss spallation neutron source is studied. The tested tungsten specimens were irradiated to two doses of 1.4 and 3.5 ddpa with 37 and 140 appm He at 80 and 110 °C, respectively. The specimen of 1.4 dpa was consecutively annealed at temperatures of 500 °C, 600 °C, 800 °C and 900 °C, for 1 h, and the post-irradiation annealing effect on the microstructure was investigated. Microstructural features such as dislocations, defect clusters, dislocation loops and bubbles were observed by means of transmission electron microscopy (TEM). TEM images were obtained in different areas of the samples, to obtain quantitative information of the dislocations and defect clusters. There was a significant change in the microstructure of the tungsten after irradiation and post-irradiation annealing. The average dislocation density and defect cluster density were evaluated.
Formation of oxide layers on tungsten at low oxygen partial pressures J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-14 J. Habainy, S. Iyengar, K.B. Surreddi, Y. Lee, Y. Dai
This work focuses on the oxidation of tungsten in inert gas atmospheres containing oxygen and moisture. It is particularly relevant for the European Spallation Source where the tungsten target is cooled by purified helium gas and the 5 MW proton beam can raise the maximum target temperature beyond the threshold for oxidation. Tungsten discs were oxidized isothermally at 400° to 900 °C for 2 h in pure helium and helium mixed with oxygen and water vapor, with varying partial pressures up to 500 Pa. Tungsten was oxidized even with a small amount of oxygen (≤5 ppm) present in industrially pure helium. Non-isothermal oxidation of tungsten foils was carried out in water vapor (∼100 Pa), in situ in an environmental scanning electron microscope. On specimens oxidized in inert gas containing water vapor (2 h, pH2O p H 2 O ∼790 Pa), Auger electron spectroscopy studies confirmed the presence of a thin oxide layer (40 nm) at 400 °C. At 500 °C the oxide layer was 10 times thicker. A dark, thin and adherent oxide layer was observed below 600 °C. Above this temperature, the growth rate increased substantially and the oxide layer was greenish, thick and porous. Oxide layers with varying stoichiometry were observed, ranging from WO3 at the surface to WO2 at the metal-oxide interface. For comparison, oxidation of tungsten alloys in He 5%O2 was studied. The implications of this work on the design and operation of the helium loop for cooling the target are discussed.
Comparison of NBG-18, NBG-17, IG-110 and IG-11 oxidation kinetics in air ☆ J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-14 Jo Jo Lee, Tushar K. Ghosh, Sudarshan K. Loyalka
The oxidation rates of several nuclear-grade graphites, NBG-18, NBG-17, IG-110 and IG-11, were measured in air using thermogravimetry. Kinetic parameters and oxidation behavior for each grade were compared by coke type, filler grain size and microstructure. The thickness of the oxidized layer for each grade was determined by layer peeling and direct density measurements. The results for NBG-17 and IG-11 were compared with those available in the literature and our recently reported results for NBG-18 and IG-110 oxidation in air. The finer-grained graphites IG-110 and IG-11 were more oxidized than medium-grained NBG-18 and NBG-17 because of deeper oxidant penetration, higher porosity and higher probability of available active sites. Variation in experimental conditions also had a marked effect on the reported kinetic parameters by several studies. Kinetic parameters such as activation energy and transition temperature were sensitive to air flow rates as well as sample size and geometry.
TEM characterization on new 9% Cr advanced steels thermomechanical treated after tempering J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-14 P. Fernández, J. Hoffmann, M. Rieth, M. Roldán, A. Gómez-Herrero
Phase transformation on new six alloys reduced activation (RAFM) steels was investigated to provide the basis for the design and development of advanced steels to maintain adequate strength and creep resistance above 500 °C. The new alloys are designed to increase the amount of fine MX precipitates and reduce coarse M23C6 carbides through alloy composition refinement and the application of thermomechanical treatments. The microstructural investigations by TEM have shown M23C6, M2X, and MX precipitation after tempering at 700 °C/2 h with low dislocation recovery, while at 825 °C/2 h the martensite developed to subgrain formation and growth. At this stage, only M23C6 and MX were detected. Preliminary results demonstrate that it is feasible to produce fine MX strengthened particles dispersed in the matrix with further optimization of tempering treatments.
Temperature measurement for in-situ crack monitoring under high-frequency loading J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-14 Takashi Naoe, Xiong Zhihong, Masatoshi Futakawa
A liquid mercury target vessel, composed of type 316L stainless steel and used in a pulsed spallation neutron source, suffers not only from proton and neutron damage but also cyclic impact stresses caused by proton beam-induced pressure waves. In a previous study, we carried out an ultrasonic fatigue test to gigacycles and observed that the specimen surface temperature rose abruptly just before failure. To understand the mechanism of this temperature rise, the temperature distribution of the specimen surface was measured using a thermography instrument during an ultrasonic fatigue test. The result showed that the temperature rose locally, especially at the crack tip, and the peak position moved with crack propagation. Furthermore, nonlinear structural analysis by LS-DYNA was performed to clarify the mechanism of this temperature rise. The analytical results showed that the heat due to plastic deformation at the crack tip is the dominant factor underlying the temperature rise rather than friction between crack surfaces.
A DFT study of the stability of SIAs and small SIA clusters in the vicinity of solute atoms in Fe J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-13 C.S. Becquart, R. Ngayam Happy, P. Olsson, C. Domain
First principle calculation of helium in La2Zr2O7: Effects on structural, electronic properties and radiation tolerance J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-13 C.G. Liu, Y.H. Li, Y.D. Li, L.Y. Dong, J. Wen, D.Y. Yang, Q.L. Wei, P. Yang
First principle calculations based on density functional theory have been employed to study structural effects of trapping helium in La2Zr2O7 pyrochlore. Lattice swelling and the distortion of unit cell have been found in He-La2Zr2O7 systems. By analyzing the electronic structures and chemical bonding of He-La2Zr2O7 systems, weak repulsive and attractive chemical interactions of helium in La2Zr2O7 pyrochlore have been observed. The formation energies have been calculated to assess the relative stability of various helium interstitial configurations and the results show that the octahedral interstitial site is the most stable structure. The cation antisite defect formation energies and the x positional parameter for 48f-site oxygen are calculated to predict the radiation resistance of He-La2Zr2O7 systems. The results indicate that the presence of low concentration of He interstitials may increase the radiation resistance of La2Zr2O7 pyrochlore.
Reactive spark plasma synthesis of CaZrTi2O7 zirconolite ceramics for plutonium disposition J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-12 Shi-Kuan Sun, Martin C. Stennett, Claire L. Corkhill, Neil C. Hyatt
Isolation of high purity americium metal via distillation J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-09 Leah N. Squires, James A. King, Randall S. Fielding, Paul Lessing
Pure americium metal is a crucial component for the fabrication of transmutation fuels. Unfortunately, americium in pure metal form is not available; however, a number of mixed metals and mixed oxides that include americium are available. In this manuscript a method is described to obtain high purity americium metal from a mixture of americium and neptunium metals with lead impurity via distillation.
The evolution of interaction between grain boundary and irradiation-induced point defects: Symmetric tilt GB in tungsten J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-09 Hong Li, Yuan Qin, Yingying Yang, Man Yao, Xudong Wang, Haixuan Xu, Simon R. Phillpot
Molecular dynamics method is used and scheme of calculational tests is designed. The atomic evolution view of the interaction between grain boundary (GB) and irradiation-induced point defects is given in six symmetric tilt GB structures of bcc tungsten with the energy of the primary knock-on atom (PKA) EPKA of 3 and 5 keV and the simulated temperature of 300 K. During the collision cascade with GB structure there are synergistic mechanisms to reduce the number of point defects: one is vacancies recombine with interstitials, and another is interstitials diffuse towards the GB with vacancies almost not move. The larger the ratio of the peak defect zone of the cascades overlaps with the GB region, the statistically relative smaller the number of surviving point defects in the grain interior (GI); and when the two almost do not overlap, vacancy-intensive area generally exists nearby GBs, and has a tendency to move toward GB with the increase of EPKA. In contrast, the distribution of interstitials is relatively uniform nearby GBs and is affected by the EPKA far less than the vacancy. The GB has a bias-absorption effect on the interstitials compared with vacancies. It shows that the number of surviving vacancies statistically has increasing trend with the increase of the distance between PKA and GB. While the number of surviving interstitials does not change much, and is less than the number of interstitials in the single crystal at the same conditions. The number of surviving vacancies in the GI is always larger than that of interstitials. The GB local extension after irradiation is observed for which the interstitials absorbed by the GB may be responsible. The designed scheme of calculational tests in the paper is completely applicable to the investigation of the interaction between other types of GBs and irradiation-induced point defects.
Object kinetic Monte Carlo model for neutron and ion irradiation in tungsten: Impact of transmutation and carbon impurities J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-09 N. Castin, G. Bonny, A. Bakaev, C.J. Ortiz, A.E. Sand, D. Terentyev
We upgrade our object kinetic Monte Carlo (OKMC) model, aimed at describing the microstructural evolution in tungsten (W) under neutron and ion irradiation. Two main improvements are proposed based on recently published atomistic data: (a) interstitial carbon impurities, and their interaction with radiation-induced defects (point defect clusters and loops), are more accurately parameterized thanks to ab initio findings; (b) W transmutation to rhenium (Re) upon neutron irradiation, impacting the diffusivity of radiation defects, is included, also relying on recent atomistic data. These essential amendments highly improve the portability of our OKMC model, providing a description for the formation of SIA-type loops under different irradiation conditions. The model is applied to simulate neutron and ion irradiation in pure W samples, in a wide range of fluxes and temperatures. We demonstrate that it performs a realistic prediction of the population of TEM-visible voids and loops, as compared to experimental evidence. The impact of the transmutation of W to Re, and of carbon trapping, is assessed.
TEM analysis of irradiation-induced interaction layers in coated UMo/X/Al trilayer systems (X= Ti, Nb, Zr, Mo) J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-09 H.-Y. Chiang, T. Wiss, S.-H. Park, O. Dieste-Blanco, W. Petry
Uranium-molybdenum (UMo) alloy powder embedded in an Al matrix is considered as a promising candidate for fuel conversion of research reactors. A modified system with a diffusion barrier X as coating, UMo/X/Al trilayer (X = Ti, Zr, Nb, and Mo), has been investigated to suppress interdiffusion between UMo and the Al matrix. The trilayer systems were tested by swift heavy ion irradiation, the thereby created interaction zone has been analyzed by scanning transmission electron microscopy (STEM) and energy-dispersive X-ray spectroscopy (EDX). Detailed structural characterization are presented and compared to earlier μ-xrd analysis.
Evolution of thermo-physical properties and annealing of fast neutron irradiated boron carbide J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-09 Dominique Gosset, Bernard Kryger, Jean-Pierre Bonal, Caroline Verdeau, Karine Froment
Boron carbide is widely used as a neutron absorber in most nuclear reactors, in particular in fast neutron ones. The irradiation leads to a large helium production (up to 1022/cm3) together with a strong decrease of the thermal conductivity. In this paper, we have performed thermal diffusivity measurements and X-ray diffraction analyses on boron carbide samples coming from control rods of the French Phenix LMFBR reactor. The burnups range from 1021 to 8.1021/cm3. We first confirm the strong decrease of the thermal conductivity at the low burnup, together with high microstructural modifications: swelling, large micro-strains, high defects density, and disordered-like material conductivity. We observe the microstructural parameters are highly anisotropic, with high micro-strains and flattened coherent diffracting domains along the (00l) direction of the hexagonal structure. Performing heat treatments up to high temperature (2200 °C) allows us to observe the material thermal conductivity and microstructure restoration. It then appears the thermal conductivity healing is correlated to the micro-strain relaxation. We then assume the defects responsible for most of the damage are the helium bubbles and the associated stress fields.
Investigation of mechanical and microstructural properties of Zircaloy-4 under different experimental conditions ☆ J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-09 Chinthaka M. Silva, Keith J. Leonard, Eric Van Abel, J. Wilna Geringer, Chris D. Bryan
Dimensional stability and anisotropy of SiC and SiC-based composites in transition swelling regime ☆ J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-08 Yutai Katoh, Takaaki Koyanagi, Joel L. McDuffee, Lance L. Snead, Ken Yueh
Swelling, or volumetric expansion, is an inevitable consequence of the atomic displacement damage in crystalline silicon carbide (SiC) caused by energetic neutron irradiation. Because of its steep temperature and dose dependence, understanding swelling is essential for designing SiC-based components for nuclear applications. In this study, swelling behaviors of monolithic CVD SiC and nuclear grade SiC fiber – SiC matrix (SiC/SiC) composites were accurately determined, supported by the irradiation temperature determination for individual samples, following neutron irradiation within the lower transition swelling temperature regime. Slightly anisotropic swelling behaviors were found for the SiC/SiC samples and attributed primarily to the combined effects of the pre-existing microcracking, fiber architecture, and specimen dimension. A semi-empirical model of SiC swelling was calibrated and presented. Finally, implications of the refined model to selected swelling-related issues for SiC-based nuclar reactor components are discussed.
The evolution of helium from aged Zr tritides: A thermal helium desorption spectrometry study J. Nucl. Mater. (IF 2.048) Pub Date : 2017-12-08 G.J. Cheng, G. Huang, M. Chen, X.S. Zhou, J.H. Liu, S.M. Peng, W. Ding, H.F. Wang, L.Q. Shi
The evolution of He from Zr-tritides was investigated for aging times up to about 6.5 years using analytical thermal helium desorption spectrometry (THDS). Zr films were deposited onto Mo substrates and then converted into Zr-tritides (ZrT1.70∼1.95) inside a tritiding apparatus loaded with pure tritium gas. During aging, there are at least five forms of He in Zr-tritides, and more than 99% of He atoms are in the form of He bubbles. The isolated He bubbles in lattices begin to link with each other when the He/Zr atom ratio reaches about 0.21, and are connected to grain boundaries or dislocation networks at He concentration of He/Zr ≈ 0.26. An interconnected system of channels decorated by bubbles evolves from the network dislocations, dislocation loops and internal boundaries. These He filled networks are formed completely when the He/Zr atom ratio is about 0.38. Once the He/Zr reached about 0.45, the networks of He bubble penetrate to the film surface and He begins an “accelerated release”. This critical ratio of He to Zr for He accelerated release is much greater than that found previously for Ti-tritides (0.23–0.30). The difference of He retention in Zr-tritides and Ti-tritides was also discussed in this paper.
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