Journal of Nuclear Materials最新文献

{"title":"In-situ temperature-dependent evaluation of phase makeup and gas evolution of hydrogen-loaded Ni-plated Zircaloy-4","authors":"Jose Marcial,&nbsp;Michaella S. Harris,&nbsp;Dushyant Barpaga,&nbsp;Jian Liu,&nbsp;Jarrod V. Crum,&nbsp;Walter G. Luscher,&nbsp;Andrew Casella,&nbsp;David Senor,&nbsp;Joshua A. Silverstein","doi":"10.1016/j.jnucmat.2025.156051","DOIUrl":"10.1016/j.jnucmat.2025.156051","url":null,"abstract":"<div><div>The transformation kinetics of metastable gamma-zirconium hydride has been investigated by several authors. In this work, we evaluate the transformation kinetics of gamma-ZrH within hydrogen loaded, nickel-plated Zircaloy-4 samples containing a hydrogen to zirconium atom ratio near 1. Samples were analyzed <em>in-situ</em> via high temperature X-ray diffraction (XRD) and the transformation kinetics were correlated to the degree of hydrogen out-gassing measured using a thermogravimetric analysis instrument outfitted with a mass spectrometer (TGA-DSC-MS). The correlation of <em>in-situ</em> high temperature hot-stage XRD (HS-XRD) and TGA-MS was verified by outfitting the HS-XRD with a mass spectrometer (HS-XRD-MS). Lastly, scanning electron microscopy equipped with electron backscatter diffraction (EBSD) was utilized to understand the inherent microstructures prior to thermal cycling as well as to verify the crystalline phase assemblage observed via HS-XRD.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156051"},"PeriodicalIF":3.2,"publicationDate":"2025-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144842100","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":"Enabling BWR fuel rod analysis in the BISON fuel performance code","authors":"Kyle A. Gamble ,&nbsp;Aysenur Toptan ,&nbsp;Pierre-Clément A. Simon ,&nbsp;Daniel J. van Wasshenova ,&nbsp;Jason D. Hales","doi":"10.1016/j.jnucmat.2025.156038","DOIUrl":"10.1016/j.jnucmat.2025.156038","url":null,"abstract":"<div><div>Nuclear fuel vendors around the world are pursuing approaches to sustain the existing nuclear reactor fleet consisting primarily of pressurized-water reactors (PWRs) and boiling-water reactors (BWRs). To support the industry's efforts, advanced modeling and simulation tools need to be capable of analyzing both legacy reactor concepts. BWR fuel rods are significantly different than those used in PWRs, which can affect fuel performance analysis. BWR fuel rods include: (1) an extensive use of gadolinia dopant as a burnable absorber, (2) an axial variation in fuel enrichment and gadolinia content, (3) the inclusion of a liner on the inner cladding surface to mitigate the impact of pellet-clad mechanical interaction (which impacts hydrogen and hydride distribution), (4) a lower initial fill gas pressure, (5) bottom-entry control rods, and (6) a lower coolant pressure that results in the two-phase flow boiling phenomenon. Although the primary focus of BISON has been in the area of PWR and advanced reactor fuel analyses, this paper presents the developments in BISON to support BWR fuel performance analysis. An overview of the models that account for the effects of gadolinia is highlighted. Internal mesh generation capabilities to include a liner is presented. Normal operating and transient (reactivity insertion accident) demonstration problems are presented to illustrate the impact of gadolinia, the hydrogen and hydride evolution due to the presence of the liner, and BISON's ability to simulate axially varying enrichments and dopant concentration. Bottom-entry control effects are captured by the axial power peaking factors present in the demonstration cases. Comparisons to integral experiments from the Halden IFA-681 experiments are discussed as initial validation. Reasonable comparisons are obtained for fuel centerline temperature and rod internal pressure as a function of time. Simulations of additional experiments containing Gd₂O₃-bearing fuel are necessary to completely validate the code for BWR applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156038"},"PeriodicalIF":2.8,"publicationDate":"2025-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144703342","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":"Mitigation of hydride embrittlement in Zr-2.5 %Nb pressure tube material using palladium coating","authors":"Saurav Sunil ,&nbsp;Yatindra Kumar ,&nbsp;N. Keskar ,&nbsp;Apu Sarkar ,&nbsp;Shefali Shukla ,&nbsp;R.N. Singh","doi":"10.1016/j.jnucmat.2025.156052","DOIUrl":"10.1016/j.jnucmat.2025.156052","url":null,"abstract":"<div><div>Pressure tubes made of Zr-2.5 %Nb alloy are critical components in Pressurized Heavy Water Reactors (PHWRs), where maintaining their structural integrity throughout service life is essential for safe reactor operation. One of the key in-service degradation mechanisms affecting the performance of Zr alloy components is hydride embrittlement, which significantly deteriorates their mechanical properties. In this study, we investigate a novel strategy to mitigate hydride embrittlement by promoting removal of hydrogen from Zr-2.5 %Nb alloy using a palladium (Pd) coating. Optimized conditions for depositing a uniform and adherent Pd layer via an electroless deposition technique were established. The efficiency of hydrogen removal from hydrided Zr-2.5 %Nb samples coated with Pd was demonstrated through heat treatment. Results show a significant reduction in hydrogen content, highlighting the potential of Pd coating as an effective method for mitigating hydride embrittlement by removing the hydrogen from the Zr-alloy pressure tubes and enhancing the service life of PHWR pressure tubes.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156052"},"PeriodicalIF":2.8,"publicationDate":"2025-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144714439","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":"Diffusion and dissolution mechanism of CrSi coated Zr alloy by experiments and first principles","authors":"Pengfei Ma ,&nbsp;Shihao Wu ,&nbsp;Bo Liu ,&nbsp;Haoli Wang ,&nbsp;Yaping Zhang ,&nbsp;Yicong Lan ,&nbsp;Dayan Ma ,&nbsp;Junkai Deng ,&nbsp;Yapei Zhang ,&nbsp;Kui Ge ,&nbsp;Wenxi Tian ,&nbsp;Suizheng Qiu ,&nbsp;G.H. Su","doi":"10.1016/j.jnucmat.2025.156050","DOIUrl":"10.1016/j.jnucmat.2025.156050","url":null,"abstract":"<div><div>Accident-tolerant fuel (ATF) represents a critical research direction for enhancing nuclear reactor safety, with Cr-coated zirconium alloy cladding being one of the leading ATF cladding candidates. However, under beyond-design-basis accidents (BDBA) and reactivity-initiated accident (RIA) conditions, the eutectic melting of Cr-Zr poses a significant challenge to cladding integrity and safety. Current research suggests that suppressing elemental diffusion at the Cr/Zr interface is key to designing Cr coatings with enhanced resistance to eutectic melting, with the most effective approach being the introduction of a diffusion barrier layer between the coating and the substrate. In this study, a combined experimental and first-principles simulation approach was employed to investigate atomic diffusion behavior and dissolution mechanisms in CrSi coatings. CrSi coatings were conducted using a multi-arc ion beam physical vapor deposition system, followed by vacuum annealing in the 1800s to examine interfacial diffusion phenomena between the CrSi coating and the zirconium substrate. Meanwhile, the scratch tests and steam oxidation experiments were performed to investigated the CrSi coating performance. Furthermore, first-principles calculations were further conducted to establish interfacial dissolution models for the CrSi coating and the zirconium substrate, elucidating the dissolution mechanisms of Cr and Si atoms in both the HCP-Zr and transformed BCC-Zr phases. This study provides novel insights and theoretical foundations for the development of advanced Cr-Si-containing coatings, facilitating the engineering application of coated zirconium cladding under BDBA and RIA conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156050"},"PeriodicalIF":2.8,"publicationDate":"2025-07-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144703341","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":"Deuterium erosion and retention properties of WAlB as integrated plasma-facing and shielding materials for compact tokamaks","authors":"Cong Li ,&nbsp;Linping He ,&nbsp;Jiaxing Chen ,&nbsp;Qiuyan Chen ,&nbsp;Qi Yin ,&nbsp;Zizhao Wang ,&nbsp;Meng Wang ,&nbsp;Wei Zhang ,&nbsp;Liqun Shi ,&nbsp;Ranran Su ,&nbsp;Hongliang Zhang","doi":"10.1016/j.jnucmat.2025.156035","DOIUrl":"10.1016/j.jnucmat.2025.156035","url":null,"abstract":"<div><div>Tungsten aluminum boride (WAlB) has been proposed as a promising candidate for integrated plasma-facing and shielding materials in compact tokamaks due to its excellent neutron and gamma-ray shielding properties, as well as its resistance to radiation-induced damage. However, the hydrogen isotope erosion and retention properties remain unclear. This study assesses and compares the erosion and retention properties of WAlB, molybdenum aluminum boride (MoAlB), and tungsten (W) using a combination of experiments and first-principle calculations. The WAlB sample, which contained impurity phases of W-Al and W-B, was annealed at 600 °C for 2 h prior to irradiation. This treatment increased the WAlB phase to over 80 %, making it the primary focus of this investigation. Both WAlB and W were subjected to deuterium (D) ion irradiation with fluences of 7.20 × 10²³ – 1.71 × 10²⁴ D/m² at temperatures of 452–598 K. Results show that WAlB undergoes preferential sputtering under D ions irradiation with a sputtering yield slightly higher than that of W but lower than MoAlB. D retention in the near-surface region of WAlB is only half of that in pure W but greater than that in MoAlB. Notably, no significant blistering or plastic deformation appeared on the WAlB surface, whereas a substantial amount of D-induced blisters formed on the W surface. These findings imply that while WAlB may not match W in D retention resistance, it demonstrates superior resistance to surface blister formation. Optimistically, this work suggests that future advancements in composition and structural design could enhance WAlB's resistance to D retention and sputtering, boosting its potential for application in compact tokamak reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"615 ","pages":"Article 156035"},"PeriodicalIF":3.2,"publicationDate":"2025-07-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144749630","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":"How brittle is 'too brittle'?: Comparative mechanical assessment of silicon carbide and post-quench steam-oxidized zirconium alloy cladding","authors":"Hailemariam M. Gebrelibanos ,&nbsp;SungHoon Joung ,&nbsp;Farhad Mohammadi-Koumleh ,&nbsp;Youho Lee","doi":"10.1016/j.jnucmat.2025.156047","DOIUrl":"10.1016/j.jnucmat.2025.156047","url":null,"abstract":"<div><div>This study compares the ductility of Triplex SiC composite cladding to established post-quench ductility (PQD) limits of zirconium alloys under Design Basis Accident conditions. As-received SiC cladding exhibits brittleness exceeding the limits permitted for significantly oxidized zirconium alloys under current design basis accident (DBA) safety standards. Although Triplex SiC cladding experiences only minimal oxidation-induced ductility loss with great capability for coolable geometry retention, this advantage is overshadowed by its inherent brittleness. It implies that deployment of SiC cladding requires substantial revisions to current regulatory frameworks to accommodate the unexperienced brittleness of SiC composite cladding, even under steady-state reactor operations.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156047"},"PeriodicalIF":3.2,"publicationDate":"2025-07-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144757104","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":"Microstructure evolution of Zircaloy-2 irradiated at low temperature","authors":"Xiaodong (Carol) Song ,&nbsp;Wenjing Li ,&nbsp;Hongbing Yu ,&nbsp;Patrick Lysz ,&nbsp;David Cho","doi":"10.1016/j.jnucmat.2025.156046","DOIUrl":"10.1016/j.jnucmat.2025.156046","url":null,"abstract":"<div><div>The CANDU®<span><span>¹</span></span> reactor, which is based on natural uranium fuel and heavy water moderator, has 380 or 480 horizontal fuel channels in the core. A fuel channel consists of two concentric tubes, each approximately 6 meters long. The inner tube, known as the pressure tube, contains the uranium fuel bundles and the pressurized primary coolant. The outer tube, known as the calandria tube, separates the hot pressure tube (∼300 °C) from the cool heavy water moderator (∼70 °C) and provides support to the pressure tube through the garter spring spacers. The calandria tubes used in CANDU reactors are made from Zircaloy-2. The fast neutron flux (<em>E</em> &gt; 1 MeV) in the reactor produces irradiation damage in zirconium alloys, through nucleation and growth of dislocation loops. X-Ray diffraction (XRD) and transmission electron microscopy (TEM) analyses have been performed on calandria tube materials that were irradiated in CANDU reactors to characterize microstructural changes produced by neutron irradiation. Under these fluences and temperatures applicable to the calandria tube service life, the predominant form of irradiation damage observed is <em>a</em>-type dislocation loops. The <em>a</em>-type dislocation density evolves rapidly at low fluences and approaches saturation after a fast neutron fluence of about 7 × 10²⁵ n/m². This saturation fluence is significantly higher than that observed from Zr-2.5Nb pressure tubes operating at a temperature of about 300 °C. The trend in the <em>a</em>-type dislocations, as a function of fast neutron fluence, can be correlated with the observed calandria tube deformation behaviour in the reactor. Furthermore, it aligns with the changes in mechanical properties of irradiated Zircaloy-2 materials documented in the open literature.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156046"},"PeriodicalIF":2.8,"publicationDate":"2025-07-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144714057","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":"Understanding the effect of minor alloying elements on helium bubble formation in ferritic-martensitic steels","authors":"Xingyu Liu ,&nbsp;Jonathan Poplawsky ,&nbsp;Yongqiang Wang ,&nbsp;Xinyuan Xu ,&nbsp;Xiang (Frank) Chen ,&nbsp;Xing Wang","doi":"10.1016/j.jnucmat.2025.156045","DOIUrl":"10.1016/j.jnucmat.2025.156045","url":null,"abstract":"<div><div>Ferritic-martensitic steels are promising structural materials for advanced nuclear reactors. To minimize long-term radioactivity, reduced-activation ferritic-martensitic steels have been developed by substituting high-activation elements like Ni and Mo with low-activation elements such as W. However, the impact of these alloying modifications on helium bubble formation, which plays a key role in material swelling, remains unclear. In this study, we compared helium bubble formation in ferritic-martensitic steel T91 and reduced-activation ferritic-martensitic steel F82H. Both materials were irradiated with sequential 100 keV, 150 keV, and 200 keV helium ions to a dose of 0.5 dpa and a helium concentration of 9000 appm at 500 °C. The helium bubbles in F82H exhibited a larger average size and a lower density than those in T91, suggesting differences in minor alloying elements may influence the bubble growth. To investigate the effects of these alloying elements, we characterized radiation-induced segregation near bubbles and grain boundaries. Prominent Ni-Mn-Si enriched clusters were found near bubbles in T91, while only Mn-Si enriched clusters were found near bubbles in F82H. In addition, the obvious Cr enrichment near grain boundaries was absent around bubbles in both steels. The different segregation trends among elements revealed the variations in element diffusion mechanisms and the different sink biases between bubbles and grain boundaries. Cr enrichment near grain boundaries is mostly driven by interstitial-mediated diffusion. However, since bubble growth relies on net vacancy flux, vacancy-mediated diffusion plays a dominant role in controlling element segregation near bubbles. Therefore, Cr enrichment was not found near bubbles. Because of preferential vacancy-drag diffusion for Ni, Si and Mn, these elements were enriched near bubbles. Due to the strong binding energies of vacancies with these solute atoms, the vacancy diffusivity can be reduced near these solutes. Therefore, the more prominent Ni-Si-Mn clustered near helium bubbles in T91 lead to stronger suppression of helium bubble growth compared to F82H.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156045"},"PeriodicalIF":3.2,"publicationDate":"2025-07-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144750032","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":"Some observations on the habit plane of hydride plates precipitated in Zr-2.5Nb pressure tubes","authors":"Sebastian C. Müller ,&nbsp;María Inés Luppo ,&nbsp;Gladys Domizzi","doi":"10.1016/j.jnucmat.2025.156048","DOIUrl":"10.1016/j.jnucmat.2025.156048","url":null,"abstract":"<div><div>The habit plane of hydrides precipitated in a Zr-2.5Nb pressure tube has been studied through transmission electron microscopy. Results showed that the trace of the microscopic hydride, as seen on different orientations of the α-Zr crystal, indicates a habit plane close to <span><math><mrow><mo>{</mo><mn>1</mn><mspace></mspace><mn>0</mn><mspace></mspace><mover><mrow><mn>1</mn></mrow><mo>‾</mo></mover><mspace></mspace><mn>4</mn><mo>}</mo><mi>α</mi></mrow></math></span>. This type of habit plane corresponds to the formation of “primary” hydride plates during the early stage of the precipitation process. In light of the present results, the discrepancies found in the literature concerning the habit plane of hydrides have been discussed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156048"},"PeriodicalIF":3.2,"publicationDate":"2025-07-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144767087","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":"Atom probe study of second-phase particles in Zircaloy-4","authors":"Siyu Huang ,&nbsp;Levi Tegg ,&nbsp;Sima Aminorroaya Yamini ,&nbsp;Vidur Tuli ,&nbsp;Patrick Burr ,&nbsp;Ingrid McCarroll ,&nbsp;Limei Yang ,&nbsp;Katie L. Moore ,&nbsp;Julie M. Cairney","doi":"10.1016/j.jnucmat.2025.156049","DOIUrl":"10.1016/j.jnucmat.2025.156049","url":null,"abstract":"<div><div>Zirconium (Zr) alloys, such as Zircaloy-4 (Zy-4), are widely used for cladding in nuclear applications. Zy-4 consists of an α-Zr matrix and various second phase particles (SPPs). These precipitates play a crucial role in determining the overall alloy performance, so understanding their composition is essential for the development of these nuclear materials. We have studied two SPPs in this alloy, Zr(Fe,Cr)₂ and Zr₂(Si,Fe), using site-specific focused ion-beam lift-out and atom probe tomography, and measured the composition and distribution of alloying elements at the precipitate/matrix interface. Residual Cu and B segregated to the interfaces of both precipitates and the matrix while Sn only segregated to the interface in the Zr₂(Fe,Si) precipitate. Hydrogen segregation was observed at the interface of Zr(Fe,Cr)₂ and the matrix.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156049"},"PeriodicalIF":2.8,"publicationDate":"2025-07-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144696814","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}