Journal of Nuclear Materials最新文献

{"title":"Evolution of dislocation loops and voids in post-irradiation annealed ThO2: A combined in-situ TEM and cluster dynamics investigation","authors":"Sanjoy Kumar Mazumder ,&nbsp;Kaustubh Bawane ,&nbsp;J. Matthew Mann ,&nbsp;Aaron French ,&nbsp;Lin Shao ,&nbsp;Lingfeng He ,&nbsp;Anter El-Azab","doi":"10.1016/j.jnucmat.2023.154686","DOIUrl":"10.1016/j.jnucmat.2023.154686","url":null,"abstract":"<div><p><span>The effect of isochronal annealing on the evolution of dislocation loop and void population in proton irradiated ThO</span>₂ has been investigated. Post-irradiation annealing in other actinide oxides like UO₂ shows significant loop coarsening. ThO₂ samples were irradiated with 2 MeV protons up to a dose of 0.1 dpa at 600 °C. Post-irradiation isochronal annealing was performed at 600, 800, 1000 and 1100 °C for 1 h at each temperature using <em>in-situ</em><span> TEM. Only faulted 1/3&lt;111&gt; type dislocation loops were observed, and their sizes and distribution were characterized. The population of self-interstitial atom (SIA) dislocation loops did not show any significant growth and coarsening. Additionally, nanometric voids were observed at annealing temperatures of 1000 and 1100 °C. Using cluster dynamics (CD), we have studied the nucleation and growth of point defects and defect clusters, i.e., SIA prismatic dislocation loops and nanometric and sub-nanometric voids in proton irradiated ThO</span>₂<span><span><span>. The CD model was further utilized to predict the growth and coarsening of loops and voids during isochronal annealing at the experimental and higher temperatures. The model did not predict significant SIA loop growth which closely corresponds to the TEM observations. CD predicted SIA loop coarsening is insignificant even at high annealing temperature of 1500 °C because the model only considers the growth of defect clusters by absorption of like point defects, i.e., SIA loops absorb interstitials and voids absorb vacancies, and cannot account for their migration and </span>coalescence due to </span>elastic interaction. The CD model also predicts the evolution of nanometric voids having mean size within the error bounds of TEM observations.</span></p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"586 ","pages":"Article 154686"},"PeriodicalIF":3.1,"publicationDate":"2023-08-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44303032","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Investigating the interactions between hydrotalcite and U(IV) nanoparticulates","authors":"Chris Foster ,&nbsp;Samuel Shaw ,&nbsp;Thomas S. Neill ,&nbsp;Nick Bryan ,&nbsp;Nick Sherriff ,&nbsp;Scott Harrison ,&nbsp;Louise S. Natrajan ,&nbsp;Bruce Rigby ,&nbsp;Katherine Morris","doi":"10.1016/j.jnucmat.2023.154482","DOIUrl":"https://doi.org/10.1016/j.jnucmat.2023.154482","url":null,"abstract":"<div><p>In the UK, the decommissioning of legacy spent fuel storage facilities at the Sellafield nuclear facility requires the retrieval of radioactive sludge resulting from Magnox fuel corrosion. However, sludge retrievals may enhance uranium mobility including via sorption of radionuclide nanoparticles onto colloidal phases such as hydrotalcite (Mg₄Al₂(OH)₁₆(CO₃).4H₂O). Hydrotalcite is a Mg-Al layered double hydroxide (LDH) which is a corrosion product of Magnox fuel cladding. Currently, there are a paucity of studies examining interactions between actinide nanoparticles and LDH phases such as hydrotalcite. Here, a multi-technique approach was used to investigate the interactions between colloidal hydrotalcite and three different forms of nanoparticulate U(IV): nanoparticulate uraninite (UO₂); nanoparticulate UO₂ reacted with silica (UO₂-Si); and U(IV)-Si-coprecipitate under anoxic, neutral-to-alkaline conditions. Ultrafiltration and zeta potential analyses indicated that for UO₂ and UO₂-Si nanoparticulate phases, sorption to colloidal hydrotalcite was limited due to rapidly settling UO₂ and UO₂-Si aggregates (&gt;450 nm). By contrast, ultrafiltration and zeta potential analyses confirmed the U(IV)-Si-coprecipitate nanoparticle phase showed significantly higher sorption to colloidal hydrotalcite. This was due to the increased colloidal stability of intrinsic U(IV)-silicate nanoparticles which in turn promoted increased sorption to hydrotalcite. TEM imaging showed some evidence for smaller UO₂ and UO₂-Si aggregates (&lt;20 nm) sorbed to colloidal hydrotalcite. Similar behaviour was observed in TEM images of authentic pond effluent samples from Sellafield, providing confidence that the model laboratory experiments provided a bridge to the highly radioactive spent nuclear fuel pond interactions. This study highlights the potential for U(IV) nanoparticles to form a new type of colloid-colloid interaction with hydrotalcite, especially when silica is present. This further informs predictions of U(IV) (and An(IV)) behaviour in the legacy pond and silo environments, as well as in environmental scenarios where LDH mineral phases and silica are present (e.g. in geological disposal of radioactive waste).</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"582 ","pages":"Article 154482"},"PeriodicalIF":3.1,"publicationDate":"2023-08-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"2953248","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Behavior of FeCrAl-ODS cladding tube under loss-of-coolant accident conditions","authors":"Takafumi Narukawa ,&nbsp;Keietsu Kondo ,&nbsp;Yuki Fujimura ,&nbsp;Kazuo Kakiuchi ,&nbsp;Yutaka Udagawa ,&nbsp;Yoshiyuki Nemoto","doi":"10.1016/j.jnucmat.2023.154467","DOIUrl":"https://doi.org/10.1016/j.jnucmat.2023.154467","url":null,"abstract":"<div><p><span><span>To evaluate the behavior of an oxide-dispersion-strengthened FeCrAl (FeCrAl-ODS) cladding tube under loss-of-coolant accident (LOCA) conditions of light-water reactors (LWRs), the following two laboratory-scale LOCA-simulated tests were performed: the burst and integral thermal shock tests. Four burst and three integral thermal shock tests were performed on unirradiated, stress-relieved FeCrAl-ODS cladding tube specimens, simulating ballooning and rupture, </span>oxidation, and quenching, which were postulated during a LOCA. The burst temperature of the FeCrAl-ODS cladding tube was 200–</span><figure><img></figure><span><span> higher than that of the Zircaloy<span> cladding tube, and the FeCrAl-ODS cladding tube's maximum circumferential strain was smaller than or equal to that of the Zircaloy-4 cladding tube. These results indicate that the FeCrAl-ODS cladding tube has higher </span></span>strength at high temperatures than the conventional Zircaloy cladding tube. The FeCrAl-ODS cladding tube did not fracture after being subjected to an axial restraint load of ∼</span><figure><img></figure><span>, which is more than 10 times higher than the axial restraint load estimated for existing LWRs, during quenching, following isothermal oxidation at </span><figure><img></figure> for <figure><img></figure>. The FeCrAl-ODS cladding tube was hardly oxidized during this isothermal oxidation condition. However, it melted after a short oxidation at <figure><img></figure> and fractured after abnormal oxidation at <figure><img></figure> for <figure><img></figure>. Based on these results, the FeCrAl-ODS cladding tube should not fracture in the time range expected during LOCAs below <figure><img></figure>, where no melting or abnormal oxidation occurs.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"582 ","pages":"Article 154467"},"PeriodicalIF":3.1,"publicationDate":"2023-08-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"3206843","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental and numerical analysis of mechanical fields on cross-shaped specimens for stress corrosion cracking of cold-worked austenitic stainless steels exposed to primary environment","authors":"Qi Huang ,&nbsp;Yann Charles ,&nbsp;Marc Maisonneuve ,&nbsp;Cécilie Duhamel ,&nbsp;Catherine Guerre ,&nbsp;Monique Gasperini ,&nbsp;Jérôme Crepin","doi":"10.1016/j.jnucmat.2023.154478","DOIUrl":"https://doi.org/10.1016/j.jnucmat.2023.154478","url":null,"abstract":"<div><p>A stress corrosion cracking (SCC) test was performed on a cold-worked austenitic stainless steel in simulated primary water, using a cross-shaped specimen permitting sequential loading. Crack density and location were investigated by scanning electron microscopy after the SCC test. To analyse the mechanical fields in the cracking areas, finite element simulations of the whole mechanical loading were conducted, involving both strain-path and temperature changes. Combined isotropic-kinematic hardening was used as constitutive equation and identified with tensile tests performed at room temperature and at 340 °C. Partial validation of the model was obtained by comparison of numerical strain fields with experimental measurements obtained by digital image correlation performed on a representative “ex-situ” test in air. Variations of the strain and stress fields during this test were discussed in relation with the cracking network observed at the end of the SCC test.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"582 ","pages":"Article 154478"},"PeriodicalIF":3.1,"publicationDate":"2023-08-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"3206841","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Long-term stability of upgraded SiO2-Al2O3-P2O5 waste form in domestic groundwater for application in radioactive salt waste disposal","authors":"Hyeon Jin Eom ,&nbsp;Ki Rak Lee ,&nbsp;Byeonggwan Lee ,&nbsp;Ga yeong Kim ,&nbsp;Kyuchul Shin ,&nbsp;Hwan-Seo Park","doi":"10.1016/j.jnucmat.2023.154483","DOIUrl":"https://doi.org/10.1016/j.jnucmat.2023.154483","url":null,"abstract":"","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"582 ","pages":"Article 154483"},"PeriodicalIF":3.1,"publicationDate":"2023-08-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"3400091","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Nitriding model for zirconium based fuel cladding in severe accident codes","authors":"B.S. Jäckel ,&nbsp;J.C. Birchley ,&nbsp;T. Lind ,&nbsp;M. Steinbrück ,&nbsp;S. Park","doi":"10.1016/j.jnucmat.2023.154466","DOIUrl":"https://doi.org/10.1016/j.jnucmat.2023.154466","url":null,"abstract":"<div><p>A model has been developed to describe the nitriding of partially oxidized zirconium based cladding during an air ingress sequence when the reaction has become starved of oxidant (oxygen and/or steam), and the subsequent re-oxidation of nitride following of restoration of coolant. Key aspects of the model are the estimation of oxygen-stabilised alpha zirconium, α-Zr(O), formed during pre-oxidation and its reaction with the nitrogen. Nitriding of metallic Zr is much slower than α-Zr(O), and plays a comparatively minor role. The model is based on data from separate-effects tests comprised pre-oxidation, nitriding in the absence of oxidant, and re-oxidation in the absence of nitrogen, which were used to derive the kinetic parameters for the main reaction processes. Developmental assessment was performed using the test results, demonstrating favourable agreement for the main reaction signatures. Independent assessment against Integral Test data is underway.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"582 ","pages":"Article 154466"},"PeriodicalIF":3.1,"publicationDate":"2023-08-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"2953245","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Effect of potassium doping and exposure temperature on the deuterium behavior in large-scale potassium-doped tungsten","authors":"Xiaolei Ma ,&nbsp;Chenlu Ye ,&nbsp;Ting Wang ,&nbsp;Fan Feng ,&nbsp;Wei Lv ,&nbsp;Shaoting Lang ,&nbsp;Changchun Ge ,&nbsp;Qingzhi Yan ,&nbsp;Huimin Shao ,&nbsp;Xiaoxin Zhang","doi":"10.1016/j.jnucmat.2023.154481","DOIUrl":"https://doi.org/10.1016/j.jnucmat.2023.154481","url":null,"abstract":"<div><p><span><span>In order to meet the needs of future fusion power plants in engineering application, a newly large-volume potassium-doped </span>tungsten<span> (WK) plate with a thickness of 15 mm was prepared by powder metallurgy and hot rolling technology. Pure tungsten (PW) as a comparison material was prepared using the same preparation process. To figure out the effect of K doping and exposure temperature on D retention and blistering morphology in WK and PW, the high-flux (∼10</span></span>²² m⁻² s⁻¹) and low energy (∼ 50 eV) D plasma exposure with different temperatures (423, 473, 523 and 573 K) was performed. The results show that there is serious surface blistering in both PW and WK at all exposure temperatures and PW forms D blisters larger in size but smaller in number compared with WK. The D retention in PW is significantly lower than that in WK at all exposure temperatures. Moreover, the evolution of D blisters and retention in WK is more sensitive to exposure temperature compared to PW. The effects of K doping and exposure temperature on blistering morphology and D retention were analyzed and discussed in detail.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"582 ","pages":"Article 154481"},"PeriodicalIF":3.1,"publicationDate":"2023-08-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"3273588","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Threshold parameters of vacuum arcs with W-fuzz cathodes","authors":"P.S. Mikhailov ,&nbsp;I.L. Muzukin ,&nbsp;Yu.I. Mamontov ,&nbsp;Yu.A. Zemskov ,&nbsp;I.V. Uimanov ,&nbsp;A.V. Kaziev ,&nbsp;M.M. Kharkov ,&nbsp;S.A. Barengolts","doi":"10.1016/j.jnucmat.2023.154479","DOIUrl":"https://doi.org/10.1016/j.jnucmat.2023.154479","url":null,"abstract":"<div><p>An experimental technique was developed and performed to investigate the threshold currents and operation times of vacuum arcs with tungsten cathodes having a nanostructured (so-called “fuzz”) surface. The fuzz layer, pre-formed due to exposure of the tungsten samples with helium plasma, reduced the threshold current for arc initiation by an order of magnitude. For currents of tenths of an ampere, the average arc operation time was in the microsecond range. The experimental data were interpreted using the explosive emission (ecton) model of a vacuum arc cathode spot.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"582 ","pages":"Article 154479"},"PeriodicalIF":3.1,"publicationDate":"2023-08-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"3400085","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Corrigendum to `Effect of purity on the vacancy defects induced in self–irradiated tungsten: A combination of PAS and TEM’ Journal of Nuclear Materials 556 (2021) 153175","authors":"Z. Hu ,&nbsp;P. Desgardin ,&nbsp;C. Genevois ,&nbsp;J. Joseph ,&nbsp;B. Décamps ,&nbsp;R. Schäublin ,&nbsp;M.-F. Barthe","doi":"10.1016/j.jnucmat.2023.154434","DOIUrl":"https://doi.org/10.1016/j.jnucmat.2023.154434","url":null,"abstract":"","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"582 ","pages":"Article 154434"},"PeriodicalIF":3.1,"publicationDate":"2023-08-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"2953247","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Microstructural understanding on the fouling behavior of crud on PWR fuel cladding surface","authors":"Jianqiao Yang ,&nbsp;Jiahuan Wang ,&nbsp;Junkai Liu ,&nbsp;Shuzhong Wang ,&nbsp;Di Yun ,&nbsp;Dayan Ma ,&nbsp;Yanhui Li ,&nbsp;Donghai Xu","doi":"10.1016/j.jnucmat.2023.154500","DOIUrl":"https://doi.org/10.1016/j.jnucmat.2023.154500","url":null,"abstract":"<div><p>In the present paper, an accelerate crud creation experiment was conducted by a laboratory simulation facility. A TEM sample was prepared at the interface between the crud layer and pressurized water reactor (PWR) fuel cladding tube to investigate the fouling behavior of crud on the surface of cladding tubes at a microstructural level. The results show that the crud layer consisted mainly of NiFe₂O₄, as Fe³⁺ and Ni²⁺ serving as the primary sources. Heterogeneous growth of Fe₃O₄ was observed by HRTEM, which formed dislocations NiFe₂O₄ grains. Based on the microstructural characterization results, the growth mechanism of crud particles on cladding tube surface during subcooled nucleate boiling was discussed with a focus on mass and heat transfer.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"582 ","pages":"Article 154500"},"PeriodicalIF":3.1,"publicationDate":"2023-08-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"3459779","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}