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Journal of Nuclear Materials
North-Holland
Journal of Nuclear Materials

North-Holland

0022-3115

Journal of Nuclear Materials/Journal Journal of Nuclear MaterialsSCIISTPEIAHCI
正式出版
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    Microstructure and tensile behavior of powder metallurgy FeCrAl accident tolerant fuel cladding

    Dolley E.Crawford C.Othon M.A.Spinelli I....
    5页
    查看更多>>摘要：© 2022 The Author(s)Defect-free seamless FeCrAl cladding tubes with 0.3 mm wall thickness have been successfully developed via full-scale powder metallurgy (PM) manufacturing routes, providing a cost neutral replacement of Zircaloy-2 tubes with enhanced accident tolerant fuel. Microstructure and tensile properties at room temperature and 315 °C were evaluated in the tubing of two yttrium-free FeCrAl alloy compositions PM-C26M and Ferritic Alloy – Sandvik Material Technology (FA-SMT) that differ in Cr, Al, Mo and minor addition of refractory elements. The powder metallurgy FeCrAl tubes reveal finer grain size than the smallest achievable grain size by cast/wrought tube fabrication process, low retained strain, and tensile properties superior to Zircaloy-2 cladding tubes. <101> fiber texture along the tube axial direction was observed. In-situ neutron diffraction during tensile loading shows qualitatively similar trend of intergranular load transfer during elastoplastic deformation in PM-C26M and FA-SMT, while FA-SMT indicates higher dislocation density and PM-C26M reveals more intensive <101> texture evolution along loading direction. Precipitates in FA-SMT are inferred to share load from the matrix, while such load sharing is not evident in PM-C26M. Compared to texture free ferritic steel data in the literature, the <101> fiber texture in the FeCrAl tubes seems to have little effect on the grain-level tensile deformation behavior including elastic anisotropy and plastic anisotropy.
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    Study on stable solidification of silica-based ammonium molybdophosphate adsorbing cesium: Micromechanics and density functional theory modeling

    Xu L.Li H.Sang H.Mao C....
    5页
    查看更多>>摘要：© 2021 Elsevier B.V.Silica-based ammonium molybdophosphate (AMP/SiO2) adsorbent can effectively remove 137Cs from simulated wastewater discharged during the Fukushima NPP-1 accident. A pressing/sintering method for the final disposal of the spent Cs adsorbent was described herein. Immobilization of Cs in the stable ceramic solid form of Cs4Al4Si20O48 was accomplished using natural mordenite. Different sintering temperatures and duration times exerted different effects on the final microstructures and the mechanical and loading abilities of the solidified products. At the sintering condition of 1 h duration time at 1,200 °C, the Cs immobilization ratio was almost 100%, and a relatively high hardness and elastic modulus of 3.50 and 35.38 GPa, respectively, were achieved. The formation of crystals as temperature increased corresponded to the micromechanical properties of the sintered product. Density functional theory calculations revealed that Cs4Al4Si20O48 is orthogonal with cubic unit cells, and Cs–O is more likely to form an ionic bond in the crystal. The dissolution of ceramic matrix was determined to be a temperature-dependent solidification mechanism that involves a corrosion medium. These findings provide a basic technological support for the stabilization of radioactive wastes.
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    Burst characteristics of advanced accident-tolerant FeCrAl cladding under temperature transient testing

    Garrison B.Cinbiz M.N.Gussev M.Linton K....
    5页
    查看更多>>摘要：© 2021Assessment of burst characteristics of accident-tolerant fuel (ATF) claddings is essential to evaluate the safety margins of nuclear reactors, and better understanding can enable accelerated licensing for reactor concepts that include the new materials. Therefore, this study investigated the burst behavior of an ATF candidate of C26M, which is an iron–chromium–aluminum alloy (FeCrAl) under transient testing like the simulated loss-of-coolant accident (LOCA) conditions, except the water-quenching phase, in light-water reactors (LWRs). The effect of LOCA specimen length was assessed in terms of post-burst tube parameters. No critical length effect was determined on the burst pressure, burst location, and burst size measurements. Burst temperature showed larger variation likely due to its measurement approach used in this study. The post-test diametral strain was identified as the critical parameter to reduce the specimen length as compared to tube burst length or width. Postmortem optical metrology and digital image correlation were employed to determine the local strain-state during the simulated LOCA, which showed the loading path was close to equibiaxial conditions in regions away from the burst location, rather than internally pressurized conditions (or “plane-strain” tension). Pst-test microstructural characterizations of the FeCrAl revealed that ductile damage was present at the edge of the outer radial surface while the rest of the material was ruptured via cleavage at LOCA burst temperatures. Furthermore, internal grain boundary cracks were observed at locations away from the tube burst region. This behavior was considered to occur during the cooling down period of the simulated LOCA test. Overall, this study aimed to provide essential data for fuel performance code development by considering micromechanics of the deformation and failure.
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    Strengthening/softening effects of vacancies on twinning deformation in zirconium

    Xie R.Xu C.Tian X.Wang Q....
    5页
    查看更多>>摘要：© 2022Duo to the low thermal neutron capture cross-section and high corrosion resistance to water and steam at high temperatures, zirconium alloys were used as fuel cladding in light water reactors. In this work, the effects of vacancies on {101¯2} twin boundary (TB) migration were studied by molecular dynamics (MD) simulations in zirconium at different temperatures. The simulation results indicate that vacancies have a softening effect on TB migration at low temperatures < 300 K. At high temperatures >300 K, vacancies have a strengthening effect on TB migration. By mechanism analysis, we found that those vacancies in coherent positions on TB have a softening effect, while those in incoherent positions have a strengthening effect on TB migration. As the temperature increases, the vacancies in coherent positions migrate to the incoherent positions, and as a result, the softening effect at low temperatures turns into a strengthening effect at high temperatures. This work provides new insights into understanding the irradiation effects on the mechanical property of zirconium and its alloys.
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    Compatibility of UN with refractory metals (V, Nb, Ta, Cr, Mo and W): An ab initio approach to interface reactions and diffusion behavior

    Liu H.Costa D.R.Lopes D.A.Olsson P....
    5页
    查看更多>>摘要：© 2021 The AuthorsUranium mononitride (UN)-uranium dioxide (UO2) composites are being considered as an innovative advanced technology fuel option for light water reactors, where an optimal balance between the chemical advantages of UO2 and the thermal and neutronic properties of UN is struck. However, the effect and extent of chemical interactions between UN and UO2 during sintering and operation are still open issues of importance. A possibility to avoid these interactions is to protect the UN phase before sintering the UN-UO2 composites by encapsulating the UN. This protective material must have a high melting point, high thermal conductivity, and reasonably low neutron cross-section. Among many candidates, the use of refractory metals is a promising option. In this study, density functional theory calculations (DFT) were performed to study the interactions and kinetics at the UN-X interfaces respectively (X = V, Nb, Ta, Cr, Mo, and W). The diffusion behaviors in UN and in the metal were studied using the self-consistent mean field (SCMF) theory. Generally, the diffusion of metal atoms in UN is slow compared to the diffusion of N atoms in the metals. Furthermore, the DFT calculations predict that Ta and V may react with UN to form UTaN2 and V8N at the UN-X interfaces, respectively. In some cases, the formation of these phases also promotes the formation of point defects in the UN and metal phases. The interaction between W and Mo with the UN phase is largely prohibited. According to this work, Mo and W can be regarded as highly promising candidate materials for the fabrication of stable UN-UO2 composite fuel.
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    Steam oxidation of uranium mononitride in pure and reducing steam atmospheres to 1200 °C

    Sooby E.S.Brigham B.A.Robles G.White J.T....
    5页
    查看更多>>摘要：© 2021 Elsevier B.V.The present study addresses the lack of steam oxidation kinetic data for high purity uranium mononitride (UN), assesses the sensitivity of the dynamic response of UN to the density of the monolithic sample exposed to flowing steam, as well as observes the variation in the oxidation kinetic response of UN to the presence of a common fabrication contaminant, carbon. In addition, the mass spectrometry performed on the evolved gas during the reaction with steam informs the reactions thermochemistry for UN with flowing steam during a thermal ramp and during an isothermal exposure at 500 °C. It is concluded here that high purity, high density samples, (≥94% theoretical density), have a significantly delayed onset of reaction with steam compared to values previously reported, the kinetic reaction is retarded in a reducing steam atmosphere (steam + 100 ppm H2), and the reaction products are predominantly ammonia (NH3 gas) and UO2 at temperatures less than 700 °C, above which UO2+x (0.2 ≥ x ≥ 0.0) begins to form and NH3(g) dissociates.
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    Evaluating the reliability of PWSCC resistance of TT Alloy 690 and associated welds to the end of anticipated PWR plant life

    Yonezawa T.Hashimoto A.
    5页
    查看更多>>摘要：© 2021 The AuthorsIn order to evaluate the primary water stress corrosion cracking (PWSCC) resistance and possible SCC growth rate (SCCGR) for thermally treated (TT) Alloy 690 to the end of anticipated PWR plant life, including various levels of cold work and associated welds, long-term PWSCC initiation tests using reverse U-bend specimens and SCCGR measurement tests were conducted at 360 °C in simulated PWR primary water. In addition, the possibility of long-range ordering (LRO) and/or short-range ordering (SRO) was evaluated by micro-Vickers hardness measurements, electron diffraction analysis and nano-scale EDS by high resolution TEM. The studied alloys were TT Alloy 690 and a Ni-33at%Cr model alloy after long-term ageing in air at 360, 400, 420 or 475 °C. No SCC was detected on any of the reverse U-bend specimens after testing for up to 45,000 h at 360 °C in simulated PWR primary water. Moreover, SCCGRs for the as-received TT Alloy 690 and with various levels of cold work or its associated welds did not increase when tested at 360 °C in simulated PWR primary water after thermal ageing at 360, 400, 420 or 475 °C in air for up to ∼20,000 h. In the case of the Ni-33at%Cr model alloy, LRO/SRO was clearly identified by electron diffraction analysis after thermal ageing at 475 °C in air. However, no evidence of LRO/SRO was observed in the case of the as-received TT Alloy 690 or after various levels of cold work or in its associated welds by any of the test methods after long-term thermal ageing in air at any temperature. From these test results, it is concluded that TT Alloy 690 and its associated welds in real components of PWRs have extremely high SCC resistance and extremely low SCCGRs should PWSCC be initiated, up to the end of any currently anticipated PWR plant life time.
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    Void swelling of conventional and composition engineered HT9 alloys after high-dose self-ion irradiation

    Kim H.Rietema C.J.Atwani O.E.Chancey M.R....
    5页
    查看更多>>摘要：© 2021 Elsevier B.V.Ferritic/martensitic (F/M) steels are being considered as potential structural materials for next generation nuclear reactors, and variants of the alloy HT9 are some of the most promising candidates. In this study, two conventional and two composition engineered HT9 alloys were irradiated using 3.5 MeV Fe2+ up to 600 peak displacement-per-atom (dpa) at 450 ℃. Void swelling and microstructure evolution were characterized for each alloy and compared. The two conventional HT9 alloys (INL and ACO3) showed similar void swelling behavior due to their similar elemental composition and processing conditions. The INL HT9 exhibited a maximum of 2.4% swelling and the ACO3 HT9 showed a maximum of 2.8% swelling at 342 and 393 average local dpa, respectively. On the other hand, the two-composition engineered HT9 alloys with varying N contents (10 ppm for low N and 440 ppm for high N) showed disparate swelling behavior. The low N HT9 exhibited a maximum of 4.6% swelling, while the high N HT9 showed a maximum of 0.7% swelling at 342 average local dpa. Changes in the N content also affected Ni/Si rich G-phase formation. The low N HT9 showed a larger size and lower density of G-phase precipitates compared with the high N HT9 after 600 peak dpa irradiation. This study compares the void swelling behavior of the ion irradiated four current HT9 alloys to extremely high doses, with the void swelling data from neutron irradiated HT9 alloys. The comparison lends critical insights into how well these current alloys can withstand high neutron fluxes in future reactors, especially since the low N and high N HT9 alloys have never been exposed to such high doses before.
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    Coexistence of a self-interstitial atom with light impurities in a tungsten grain boundary

    Fernandez-Pello D.Cerdeira M.A.Suarez-Recio J.Iglesias R....
    5页
    查看更多>>摘要：© 2021 The Author(s)In this paper, we report on ab initio simulations results focused on completing a thorough energetic, structural, charge and mobility analysis of the synergistic behaviour of diverse defects, namely self-interstitial atoms (SIA) and light impurity atoms (LIA), i.e., He and H, that would appear in W when simultaneously irradiated with the latter. In particular, the influence of a W〈110〉/W〈112〉 grain boundary (GB) in the behaviour of coexisting defects is studied and compared with the results obtained in the bulk. Four possible scenarios are analysed concerning the occupation of the GBs with: (i) a single SIA (ii) the simultaneous presence of two different defects, that is, He-H or SIA-LIA pairs, and (iii) the three types of defects together. The most stable configuration in each of these scenarios is detailed. Results show that GBs act as trapping sites for SIAs and LIAs and that the interaction between He and H is weak in all the analysed arrangements. They also indicate that the introduction of a SIA in a GB preloaded with He and H affects each of the atoms differently, as the former tends to stay close to the extra W atom, while the latter finds more comfortable accommodations away from the other two defects. In bulk W, the qualitative behaviour of He and H is quite similar and the presence of a LIA strongly affects the preferred orientation of the SIA dumbbell. Additionally, defect mobilities along the GB have been assessed concluding that the SIAs tend to move along the interfacial grooves, so to recombine with the vacancies present there.
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    Interaction between <c> dislocation loop and {101¯2} twin boundary in magnesium

    Fan H.Tang J.Xu C.Li Z....
    5页
    查看更多>>摘要：© 2021 Elsevier B.V.In this work, molecular dynamics simulations were employed to investigate the interaction between a {101¯2} twin boundary (TB) and irradiation-induced <c> dislocation loops in magnesium at room temperature. According to the geometry of the Burgers vectors with respect to TB, three sets of partial Burgers vectors can be identified, i.e. 1/6[202¯3]+1/6[2¯023], 1/6[22¯03]+1/6[2¯203] and 1/2[0001]+1/2[0001], which cover all the possible interaction styles between <c> dislocation loop and TB. The simulation results show that, although the partial Burgers vectors are different, the TB is able to always transform all the dislocation loops into glissile <a> dislocation loops in the twin. Therefore, an elimination mechanism of the irradiation-induced defects can be proposed in magnesium, i.e. by interaction with TB. The current work provides new insights into understanding the irradiation effects in magnesium and other hexagonal close-packed metals.
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