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Effects of impurities on stability of TiC, TaC and ZrC particles in tungsten Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-03-11 Yu-Wei You, Jinwei Xuan, Kui Hou, Li Wang, Rong Yan, Dongdong Li, X.S. Kong, Xuebang Wu, C.S. Liu
Titanium carbide (TiC), tantalum carbide (TaC) and zirconium carbide (ZrC) in form of particles are widely added in W to improve its mechanical and anti-irradiation properties. However, TiC, TaC and ZrC particles are decomposed to Ti-C-O, Ta-C-O and Zr-C-O in W, respectively. To understand the micro-mechanisms, we carry out systematical simulations and find that the presence of impurities such as O
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3D benchmark experiments of tritium in tungsten for tritium measurements by detecting β-ray induced X-rays Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-03-11 Yang Yang, Zhilin Chen, Po Huang, Yong Yang, Caifeng Lai, Wenxiang Jiang
Tungsten is one of the most promising materials for plasma facing components (PFCs), and -ray induced X-ray spectrometry (BIXS) is an important non-destructive method to obtain tritium depth profile and retention information for the development of PFCs, recovery of fuel and control of tritium safety in a fusion reactor. In this article, benchmark experiments to obtain tritium 3D profile in tungsten
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Study of hydrogen sorption and desorption processes of zirconium beryllide ZrBe2 Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-03-11 Yergazy Kenzhin, Inesh Kenzhina, Timur Kulsartov, Zhanna Zaurbekova, Saulet Askerbekov, Yuriy Ponkratov, Yuriy Gordienko, Alexandr Yelishenkov, Sergey Udartsev
One of the promising intermetallic compounds for use in nuclear and fusion reactors, as well as in hydrogen energy technology, are intermetallic compounds of beryllium with metals such as Ti, V, Zr and Nb. Beryllium-based intermetallics are not only a promising material for blankets of future fusion reactors, but can also be utilized in other areas of the nuclear industry, including fission reactor
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Corrigendum to “Comprehensive modeling of hydrogen transport and accumulation in titanium and zirconium”[Nucl. Mater. Energy 23 (2022) 100751] Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-03-08 Chihiro Kobayashi, Yoshiki Hamamoto, Katsuaki Tanabe
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Exploring the impacts of Li and He impurities in a tungsten matrix: A First-Principles study Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-03-07 Jia-Cheng Liang, Chuan-Lu Yang, Xue-Lin Wang
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DONES EVO: Risk mitigation for the IFMIF-DONES facility Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-03-05 M. Weber, A. Marchena, J. Aguilar, N. Ballesteros, I. Cobo, A. Echeverría, D. Esperante, E. Fernández, E. Fernández, A. García, T. García, B. Garcinuño, D. Gavela, P. Gil, A. Gómez, L. Gutiérrez, A. Ibarra, D. Iriarte, A. Lázaro, R. Lorenzo, E. López-Melero, J. Maestre, R. Maldonado, A.J. Martínez, L. Mendoza, C. de la Morena, F. Mota, C. Oliver, M.I. Ortíz, J. Patiño, I. Podadera, I. Porras, J. Praena
The International Fusion Materials Irradiation Facility- DEMO Oriented Neutron Source (IFMIF-DONES) is a scientific infrastructure aimed to provide an intense neutron source for the qualification of materials to be used in future fusion power reactors. Its implementation is critical for the construction of the fusion DEMOnstration Power Plant (DEMO).
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Effects of beam scanning modes on ion-irradiated iron microstructure Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-03-01 T. Dunatov, M. Roldan, T. Tadić
Neutron induced damage in future nuclear materials can be studied using heavy ion beams only if the differences in the microstructure evolution are well understood. Large variations in the damage dose rate, caused by scanning of the ion beam, can alter the microstructure compared to steady-state irradiation. We study the effect of scanning on the microstructure by irradiating pure iron with a 10 MeV
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First study of the location of deuterium in displacement-damaged tungsten by nuclear reaction analysis in channeling configuration Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-03-01 S. Markelj, E. Punzón-Quijorna, M. Kelemen, T. Schwarz-Selinger, R. Heller, X. Jin, F. Djurabekova, E. Lu, J. Predrag
Nuclear reaction analysis (NRA-C) together with Rutherford backscattering spectrometry (RBS-C), both in a channeling configuration were used to study the location of deuterium (D) in irradiation-induced defects in tungsten (W) using a He probe beam. The defects were created by W ion irradiation at two different damage doses of 0.02 and 0.2 dpa and two temperatures of 290 K and 800 K. Angular scans
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HICU PIE results of neutron-irradiated lithium metatitanate pebbles Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-03-01 Julia Leys, Rolf Rolli, Hans-Christian Schneider, Regina Knitter
Lithium metatitanate (LiTiO) pebbles were irradiated with neutrons within the HICU (High neutron fluence Irradiation of pebble staCks for fUsion) experiment to investigate their material properties and tritium release behaviour in a post-irradiation examination (PIE). The irradiation temperature is the most significant influence on the material. Besides a higher irradiation temperature, a higher initial
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Deuterium reclamation from C-Si codeposits using thermo-oxidation Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-03-01 Adam W. Cruse, James W. Davis
SiC exhibits a remarkable resistance to neutron irradiation damage and, being a low-Z material, is seen as a potential material candidate for plasma-facing components in magnetic confinement fusion reactors. The current work investigates the reclamation of deuterium from C-Si codeposits produced by sputter-deposition at temperatures from 300 K to 700 K using thermo-oxidation at 350 °C and 400 °C (623
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D retention in e-beam powder-bed fused (3-D printed) tungsten exposed to high-flux deuterium plasma in Pisces-RF Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-03-01 M.J. Baldwin, H. Zhang, A. Založnik, M.I. Patino, M.J. Simmonds, D. Nishijima, P.R. Carriere, G.R. Tynan, T. Horn
Tungsten targets produced by the additive manufacturing (AM) method of electron-beam powder-bed fusion, or 3-D metal printing, are exposed to high flux D plasma in the linear plasma device with the plasma-exposed surface normal to the AM build direction. D retention was measured by thermal desorption mass spectrometry following exposure to D plasma with an associated D ion flux. D fluence, and operational
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Study on irradiation hardening by He+ and subsequent V/self-ion irradiation in V-4Cr-4Ti under near-service conditions Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-29 Shoushuai Zhang, Shaoning Jiang, Jianghai Lin, Yuhai Xia, Liping Guo, Pengfei Zheng
V-4Cr-4Ti was irradiated by He and subsequent V/self-ions at 773 K to simulate the effect of the damages generated under near-service conditions on irradiation hardening. Microstructure observation and hardness test were carried out by transmission electron microscopy (TEM) and nanoindentation testing. Microstructure analysis showed that bubbles, dislocation loops and precipitates were produced whether
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Microstructure evolution of γ-Al2O3/FeAl tritium permeation barrier coatings under 6.4 MeV Fe3+ ion irradiation Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-29 Fengcheng Liu, Hucheng Yu, Xiaoou Yi, Shunjie Deng, Shulei Li, Kiyohiro Yabuuchi, Somei Ohnuki
γ-AlO/FeAl coatings for tritium permeation barrier (TPB) application in fusion reactors were fabricated via electro-chemical deposition and selective oxidation. The TPB coatings consisted of a γ-AlO scale (200 nm), an upper interlayer of FeAl (20 μm) and a lower interlayer of FeAl (10 μm). Their potential in-reactor performance was investigated based on 6.4 MeV Fe ion irradiations at 400 °C, up to
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Study on Al2O3/Fe-Al gradient coating assisted prepared by ionic liquid for Lead-Bismuth eutectic corrosion resistance Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-24 Sihao Huang, Lilong Pang, Pengfei Tai, Zhiguang Wang, Tielong Shen, Jianlong Chai, Zhiwei Ma, Hailong Chang
Alumina coating which has excellent corrosion resistance to lead–bismuth eutectic (LBE) is very promising to be used in future LBE cooling system. In this paper, the AlO/Fe-Al gradient coating is prepared on the surface of SIMP steel assisted by ionic liquid aluminum electroplating and subsequent heat treatments and selective oxidation. The protective AlO layer is composed of much γ-phase and a small
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Studies of irradiated two-phase lithium ceramics Li4SiO4/Li2TiO3 by thermal desorption spectroscopy Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-23 Yevgen Chikhray, Saulet Askerbekov, Regina Knitter, Timur Kulsartov, Asset Shaimerdenov, Magzhan Aitkulov, Assyl Akhanov, Darkhan Sairanbayev, Zhanar Bugybay, Aigerim Nessipbay, Kirill Kisselyov, Gunta Kizane, Arturs Zarins
Two-phase ceramics LiTiO-LiSiO are one of the potentially promising materials for creating a ceramic blanket for DEMO reactor. However, until now, a limited number of studies have been carried out on the release of tritium from this material under the influence of neutron irradiation, which poses fundamental problems for its application.
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The effect of High-Temperature Pre-Damage on Vacancy-Type defects and deuterium retention in tungsten Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-21 Xiu-Li Zhu, Zhen-Hua Ke, Long Cheng, Peng Zhang, Yue Yuan, Xing-Zhong Cao, Guang-Hong Lu
This work investigated tungsten samples that were pre-damaged at room (300 K) and high temperatures (573 and 873 K) using 3.5 MeV iron ions and then exposed to the high-flux deuterium plasma. The vacancy-type defects and the behavior of deuterium retention in the investigated tungsten were characterized and compared to study the effect of high-temperature pre-damage. Results of slow positron annihilation
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Effect of metal impurities on the adsorption energy of cesium and work function of the cesiated Mo (0 0 1) surface Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-20 Heng Li, Xin Zhang, Yuhong Xu, Guangjiu Lei, Sanqiu Liu, Katsuyoshi Tsumori, Haruhisa Nakano, Masaki Osakabe, Mitsutaka Isobe, Shoichi Okamura, Akihiro Shimizu, Kunihiro Ogawa, Hiromi Takahashi, Zilin Cui, Jun Hu, Yiqin Zhu, Xiaolong Li, Huaqing Zheng, Xiaoqiao Liu, Shaofei Geng, Xiaochang Chen, Haifeng Liu, Xianqu Wang, Hai Liu, Changjian Tang
Based on the DFT method, the effects of copper and tungsten impurities present in the negative ion source of neutral beams on the cesiated surface were studied, including their effects on the adsorption energy of cesium on the surface and the surface work function. The results indicate that copper impurities significantly increase the average adsorption energy of cesium, whereas tungsten has limited
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Experimental determination of the surface rate constants of protium and deuterium in Eurofer97 Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-17 Igor Peñalva, Marta Malo, María Urrestizala, Jon Azkurreta, Natalia Alegría, Carlos Moreno, David Rapisarda
The EUROFER reduced activation ferritic-martensitic (RAFM) steel is meant to be a firm candidate as structural material to be part of future fusion reactors. The correct definition of the transport parameters of hydrogen isotopes in these materials turns out to be a key issue in this field, as long as it can affect the operation of the reactor itself when quantifying the fuel inventory as well as the
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On the factors enhancing hydrogen trapping in spherical cavities in metals Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-16 M. Zibrov, K. Schmid
Using the reaction–diffusion model for hydrogen (H) trapping in spherical cavities in metals (Zibrov and Schmid, 2022), we theoretically analyze the case of H trapping only in chemisorption sites at the cavity surface. We show that the model can be reduced to a form that is similar to the conventional model of H trapping in point defects. The reduced model includes analytical expressions for correction
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First thermal fatigue studies of tungsten armor for DEMO and ITER at the OLMAT High Heat Flux facility Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-15 D. Alegre, D. Tafalla, A. De Castro, M. González, J.G. Manchón, F.L. Tabarés, T. Hernández, M. Wirtz, J.W. Coenen, Y. Mao, E. Oyarzábal
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Study on photoluminescence properties of Er2O3 materials as irradiation damage and temperature sensors Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-15 Teruya Tanaka, Masahito Yoshino, Miyuki Yajima, Daiji Kato
Photoluminescence (PL) properties of ErO specimens were examined by using visible lasers (532 nm and 635 nm) and a UV LED light source (365 nm) to investigate the applicability for irradiation damage monitoring of materials in fusion reactors. Both in the laser induced and UV light induced PL spectra, green (510–590 nm) and red (630–725 nm) luminescence was observed. In the spectrum measurements on
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Composition design and preparation of lithium lead titanate (Li2PbxTi1-xO3, 0.1 < x < 0.9): A novel tritium breeding ceramic Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-15 Xinyu Gao, Lizhi Zhao, Jing Wang, Wei Lu, Delin Chu, Weihua Wang
In deuterium–tritium (D-T) fusion reactors, the design of tritium breeding materials plays very critical role in tritium self-sufficiency. Compared to materials like LiO, LiTiO has a weaker tritium breeding capability, because of the lower density of Li. For this reason, there is a need to further explore new lithium-containing ceramics. In this paper we have designed a new tritium-breeding LiPbTiO
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Microstructure and mechanical performance of W base candidate shielding materials sintered by HIP Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-15 Xiang Geng, Qiang Qi, Yubo Cai, Qingjun Zhu, Hai-Shan Zhou, Songlin Liu, Guang-Nan Luo
With the development of fusion devices towards compact designs, the space available for shielding is more limited. So high performance shielding materials are urgently needed to meet the shielding requirements. W-B-Fe-Cr-C Reactive Sintered Borides (RSB) has attracted attention as a potential advanced shielding material for fusion reactors. In this study, the hot isostatic pressing (HIP) sintering
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Positron lifetime study of ion-irradiated tungsten: Ion type and dose effects Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-14 B. Wieluńska-Kuś, M. Dickmann, W. Egger, M. Zibrov, Ł. Ciupiński
Polycrystalline recrystallized tungsten samples were irradiated with 7.5 MeV Si ions and 9 MeV Cu ions to three different damage levels (0.01, 0.1, 0.5dpa at 200 nm depth) at 295 K. The resulting vacancy-type defects in the samples were studied using positron annihilation lifetime spectroscopy. The dependence of the average positron lifetime on the damage level is found to be non-linear: a steep increase
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Effects of minor rhenium additions on the thermal properties and recrystallization temperature of tungsten alloy Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-14 Jinbo Shi, Jiupeng Song, Mengxia Liang, Youyun Lian, Jianbao Wang, Fan Feng, Xiang Liu
Tungsten (W) stands out as a highly promising plasma-facing materials (PFMs) for future nuclear fusion reactors due to its advantages of excellent high temperature strength, low tritium retention and high recrystallization temperature (RCT). Notably, the potassium-doped tungsten-rhenium (W-Re-K) alloy exhibits exceptional performance when compared to other potential W-based alloys. Nevertheless, even
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Atomistic modelling of tritium thermodynamics and kinetics in tungsten and its oxides Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-14 M. Christensen, E. Wimmer, M.R. Gilbert, C. Geller, B. Dron, D. Nguyen-Manh
Atomistic simulations using density functional theory and machine-learned potentials have been employed to map the structural, thermodynamic, and kinetic properties of the T-WO system (x = 0 to 3). The simulations reveal that the T permeability is low in WO, intermediate in W, and relatively high in WO. Diffusion of T is slowest in WO. Vacancies and self-interstitials are strong traps for T. Oxygen
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High-temperature interaction of water vapor with lithium ceramics Li2TiO3 Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-13 Timur Kulsartov, Kuanysh Samarkhanov, Inesh Kenzhina, Zhanna Zaurbekova, Vadim Bochkov, Alexandr Yelishenkov, Yevgen Chikhray
Selection of materials for fusion reactors and, in particular, for blanket systems is one of the most important tasks that determine the development of fusion energy. Lithium ceramics LiTiO is considered as prospective candidate for solid breeders of future fusion reactors' blankets. The paper presents the results of high-temperature tests of LiTiO + 5 mol% TiO ceramics, performed under conditions
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Chemical compatibility of high entropy alloys with liquid PbLi Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-09 T. Hernández, M.A Monge, F.J. Sánchez, A. Rodriguez-Lopez, Y. Ortega, L. Serrador, B. Savoini
The chemical compatibility and corrosion resistance of three different high entropy alloys of CrFeV-X, where X = CuW, CuMo, TaW, to stagnant eutectic PbLi was studied. Three nominal HEAs compositions, CrFeVWTa (H-TaW), CuCrFeVW (H-CuW) and CuCrFeVMo (H-CuMo) were prepared by vacuum arc melting under a low pressure He atmosphere using a non-consumable W electrode and starting from pure elemental pieces
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The angle of incidence dependence of the sputtering energy threshold of tungsten in future nuclear fusion device Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-08 Al-Montaser Bellah A. Al-Ajlony, Ghadeer H. Al-Malkawi
The angle of incidence dependence of the sputtering energy thresholds (E) for tungsten targets irradiated by Helium, Tritium, and Deuterium ions was studied using two different simulation programs (RDS-BASIC and SDTrimSP). For each of the studied irradiation systems, the sputtering energy threshold was found to slightly decrease (3 to 6 eV) by increasing the angle of incidence of the irradiating ion
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Microstructural changes in He-irradiated V-Cr-Ti alloys with low Ti addition at 700 °C Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-06 Yichen Zou, Ken-ichi Fukumoto, Ryoya Ishigami, Takuya Nagasaka
V-Cr-Ti alloys with reduced Ti contents are expected to be good candidates as recyclable structural materials for fusion reactors. Herein, He irradiation was applied to V-4Cr-xTi (x = 0 to 4) alloys to investigate the additional effect of Ti and gas interstitial impurities on their microstructural evolution and irradiation hardening. Following He irradiation of the specimens at 700 °C with a maximum
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Alloying effects of Zr, Nb, Ta, and W on thermodynamic and mechanical properties of TiC based on first-principles calculation Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-04 Rongqi Chen, Liang Chen, Qian Wang, Lei Wang, Chaoping Liang
First-principles calculation has been used to study the temperature-dependent thermodynamic and mechanical properties of TiC with additions of transition metal elements through the combination of quasi-harmonic Debye model and thermal electronic excitation. It is found that the substitution behaviors of Zr, Nb, Ta, and W doped into TiC are not only structurally stable, but also would increase its melting
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Neutron irradiation effect on superconductivity of Nb3Sn wire - 50 Hz data acquisition system - Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-03 Arata Nishimura, Yoshimitsu Hishinuma
A research facility for neutron irradiation effect on superconducting materials has been installed at Oarai center in Tohoku University in 2012 as part of a post irradiation experiment. It consists of a 15.5 T superconducting magnet, a variable temperature insert and a data acquisition and control system. Since the sample holder is cooled by thermal conduction with G-M refrigeration, there is a small
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Effect of W-Cu joining on D transport behavior in plasma-facing components for fusion reactors Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-02 Xue-Chun Li, Hai-Shan Zhou, Xin-Dong Pan, Cai-Bin Liu, Zi-Han Tao, Hao-Dong Liu, Guang-Nan Luo
In future fusion reactors, W-Cu joining technology will be utilized to fabricate plasma facing components. In this study, hydrogen isotope gas-driven permeation (GDP) and thermal desorption spectrometry (TDS) are performed for W-Cu, W and Cu samples to understand the impact of W-Cu joining on fuel transport. GDP results reveal that the diffusion activation energy of the W-Cu sample is 0.289 eV, significantly
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Effect of the heating rate and Y2O3 coating on the microstructure of Wf/Y2O3/W composites via field assisted sintering technology Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-02 Rui Shu, Yiran Mao, Alexander Lau, Jan W Coenen, Alexis Terra, Chao Liu, Johann Riesch, Christian Linsmeier, Christoph Broeckmann
Field assisted sintering technology (FAST) is one of the potential methods to fabricate tungsten fiber reinforced tungsten (W/W) composites. The microstructure and mechanical properties of W/W composites are closely related to the sintering parameters. In the present work, the W/W composites with a YO coating on the fiber (W/YO/W) were fabricated via a FAST process and the microstructure was characterized
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Ab-initio simulations of atomic hydrogen interaction with Nb and V at clean and oxygen covered surfaces Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-02-02 Alejandro Vazquez Cortes, Christian Day, Christopher Stihl, Pavel V. Vladimirov
Superpermeation allows for hydrogen fluxes through metal foil membranes at rates orders of magnitude higher than pressure driven permeation. This process occurs only for hydrogen isotopes, meaning it is hydrogen-selective, and it can work against a pressure gradient, implying pumping capabilities. These characteristics allow for using superpermeation as the base process for a very efficient, selective
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Advancement of strength and toughness in ultra-low carbon martensitic stainless steel by reversed austenite Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-28 Yu Wang, Xiaoxin Zhang, Latao Jiang, Chaolian Yuan, Jiahao Zhang, Qingzhi Yan
It has been proven that martensite matrix accompanied by fine reversed austenite is beneficial to improve the strength and toughness in martensitic steels to a certain extent. In order to achieve this unique microstructure, the existing methods, like quenching at high temperatures, multiple heat treatment processes, increasing Ni content, raise the economic cost and operation difficulty. In this work
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Optimization and evaluation of structural and shielding concrete for IFMIF-DONES Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-26 Tomasz Piotrowski, María José Martínez-Echevarría Romero, Piotr Prochoń, Mónica López Alonso, Rafał Michalczyk, Armando Arvizu-Montes, Łukasz Ciupiński, Santiago Becerril Jarque, Kazimierz Józefiak, Yuefeng Qiu, Martin Ansorge, Hari Chohan, Magdalena Wojtkowska
The aim of this study was to optimize and evaluate structural and shielding concrete for the IFMIF-DONES building. An ordinary concrete of lime-dolomite aggregate from local sources has been chosen for structural concrete and magnetite aggregate was chosen for heavy-weight radiation shielding. The reference for concrete materials design was the one used in the ITER project. After investigations of
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Effect of process control agent on the synthesis of Cu-Y2O3 by mechanical alloying Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-26 Bing Ma, Hao Ding, Feng Jiang, Yoshimitsu Hishinuma, Laima Luo, Yifan Zhang, Jing Wang, Xueyang Sheng, Hiroyuki Noto, Jiaqin Liu, Jingyi Shi, Takeo Muroga, Yucheng Wu
Two types of Cu composites containing 3 wt% Y2O3, with and without a process control agent (PCA), were fabricated using mechanical alloying (MA) and hot isostatic pressing (HIP). The results showed that the addition of PCA effectively inhibited the growth of MA powders. This was attributed to the PCA forming a lubricating film on the MA powders, preventing adhesion between the powders. However, residual
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Annealing of hydrogen trap sites in displacement-damaged EUROFER Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-20 A. Theodorou, K. Schmid, T. Schwarz-Selinger
For future fusion power plants, the reduced activation ferritic/martensitic (RAFM) steel EUROFER is considered to be the primary choice as a structural material for plasma facing components (PFC) of the first wall, such as the blanket modules. Its exposure to high energy fusion neutrons will lead to creation of radiation-induced defects, which could act as trapping sites for hydrogen isotopes (HI)
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Influence of helium bubbles location on hydrogen isotope retention and exchange behavior in plasma-facing materials: A numerical simulation investigation Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-23 Y.J. Huang, C. Hao, Q.H. Liu, J.P. Zhu, F. Sun, Y. Oya, Y.C. Wu
Tritium (T) is a costly radioactive element that, when retained in plasma-facing materials (PFMs), not only results in fuel loss but also raises issues of radioactive contamination. Hydrogen isotope exchange is a potential method for T removal in future fusion devices. However, in the nuclear fusion environment, PFMs will be subjected to low-energy and high-flux helium (He) plasma irradiation, forming
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Progress in the development of industrial scale tungsten fibre-reinforced composite materials Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-20 J. Riesch, A. von Müller, Y. Mao, J.W. Coenen, B. Böswirth, S. Elgeti, M. Fuhr, H. Greuner, T. Höschen, K. Hunger, P. Junghanns, A. Lau, S. Roccella, L. Vanlitsenburgh, J.-H. You, Ch. Linsmeier, R. Neu
Currently, tungsten fibre-reinforced (Wf) composites are regarded as promising materials for plasma-facing components of future magnetic confinement fusion devices. In this context, tungsten fibre-reinforced tungsten (Wf/W) is being investigated as a pseudo-ductile composite material overcoming the intrinsic brittleness of bulk tungsten while tungsten fibre-reinforced copper (Wf/Cu) is being developed
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The use of time-of-flight neutron Bragg edge imaging to measure the residual strains in W/Cu dissimilar joints for fusion reactors Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-20 Omar Mohamed, Bin Zhu, Nathanael Leung, Winfried Kockelmann, Thomas R. Barrett, Mark J. Whiting, Yiqiang Wang, Tan Sui
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Study on corrosion behavior of China low activation ferritic/martensitic steel in static liquid lithium Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-20 D.H. Zhang, G.Z. Zuo, X.C. Meng, G.P. Yang, B. Cao, H.B. Liao, L. Zhang, J.S. Hu
Liquid lithium (Li) is a candidate material for the first wall and blanket coolants/breeders in fusion devices. Reduced activation ferritic/martensitic (RAFM) steels are also considered primary candidate structural materials for fusion reactors. Thus, the compatibility of RAFM steel with liquid Li is one of the key issues for liquid Li first walls and blankets. In this research, the corrosion behaviors
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Beam-facing material selection for mitigation of residual doses in the HEBT of IFMIF-DONES Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-17 Francisco Ogando, Llorenç Macia, Victor Lopez, Ivan Podadera, Daniel Sanchez-Herranz
IFMIF-DONES will be an irradiation facility based on a 40 MeV deuteron accelerator. Unavoidable beam losses along the accelerator result in deuterium interactions with the beam facing materials of the vacuum beam pipe, some of them leading to material activation. The initial design of the beam pipe was based on stainless steel, but an evaluation of the residual doses from the pipe showed high values
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Garnet-type lithium sensors for the monitoring and control of Pb-Li alloys in tritium breeding modules Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-11 Marc Nel-lo, Enric Lujan, Antonio Hinojo, Sergi Colominas, Jordi Abella
During the EU-DEMO conceptualization, eutectic Pb-Li was considered as the prime candidate liquid breeding material. Nonetheless, the conversion of lithium into tritium through the breeding reaction will result in a decrease in the lithium concentration within the alloy. Then, the monitoring of the lithium concentration in Pb-Li alloys is critical for the optimal Tritium Breeding Blanket (TBB) performance
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Evaluation of hydrogen embrittlement in ODS-Cu, Cu–Cr–Zr, and Cu–Cr alloys using slow strain rate technique test Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-11 M. Hatakeyama, Y. Asai, D. Nakato, M. Nishimura, Y. Hatano, S. Sunada, K. Sato
Precipitation-hardened Cu–Cr–Zr alloy is proposed as a heat sink material for various components of the ITER owing to its high strength, high conductivity, and superior resistance against neutron irradiation. Oxide-dispersion-strengthened copper (ODS-Cu) was selected as the candidate material. Hydrogen embrittlement of Cu–Cr–Zr, Cu–Cr, and ODS-Cu (GlidCop® CuAl60) alloys was evaluated using the slow
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Exposure of tungsten heavy alloys at high thermal loads in LHD Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-10 Chandra Prakash Dhard, Suguru Masuzaki, Dirk Naujoks, Rudolf Neu, Daisuke Nagata, Mikhail Khokhlov
Tungsten has been considered a plasma-facing material in a future fusion reactor because of its low sputtering yield and low fuel retention. It has been examined in several tokamaks. In stellarators, it has recently been used for some plasma-facing components. However, in addition to its high cost, W is difficult to machine due to its hardness and brittleness and therefore alternative materials in
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Deuterium retention and transport in ion-irradiated tungsten exposed to deuterium atoms: Role of grain boundaries Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-09 S. Markelj, J. Zavašnik, A. Šestan, T. Schwarz-Selinger, M. Kelemen, E. Punzón-Quijorna, G. Alberti, M. Passoni, D. Dellasega
The influence of grain boundaries on deuterium (D) retention and transport was investigated in nanocrystalline tungsten (W) by exposing the samples to sub eV D atoms. Thin tungsten films with nanometer-sized grains were produced by pulsed laser deposition on tungsten substrates. Their grain size was increased up to one micrometer by thermal annealing in vacuum up to 1223 K. Irradiation damage was created
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Dissolution corrosion of FeCrAl alloy exposed to oxygen-depleted lead–bismuth eutectic containing Ni impurities at 600 ℃ Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-09 Hao Ren, Xian Zeng, Xiaoxin Zhang, Jun Zhang, Xiaodong Huang, Xintong Zhang, Qingzhi Yan
FeCrAl alloy is one of the potential candidates of structural materials due to its high strength, low radiation swelling and good oxidation resistance for the application of lead-cooled fast reactors (LFRs). Particularly, FeCrAl alloy can form protective Al-containing oxide scales after exposure to liquid oxygen-rich Pb and lead–bismuth eutectic (LBE) at 400–800 ℃, showing better performance than traditional
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Changeover between helium and hydrogen fueled plasmas in JET and WEST Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-07 T. Wauters, R. Bisson, E. Delabie, D. Douai, A. Gallo, J. Gaspar, I. Jepu, Y. Kovtun, E. Pawelec, D. Matveev, A. Meigs, S. Brezinsek, I. Coffey, T. Dittmar, N. Fedorczak, J. Gunn, A. Hakola, P. Jacquet, K. Kirov, E. Lerche
The pre-fusion power operation (PFPO) phase of ITER, as described in the ITER research plan with Staged Approach2, includes both hydrogen (H) and helium (He) plasma operations. In preparation for PFPO, both WEST and JET ran He plasma campaigns to study plasma-wall interactions in a tungsten environment. The campaigns included a back-and-forth transition between H or deuterium (D) and He plasma operation
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Determination of Tritium-Helium-3 differential cross-section in the energy range between 0.6 MeV and 3.3 MeV for tritium depth profiling in solids Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-06 S. Markelj, A. Cvetinović, M. Lipoglavšek, M. Kelemen, M. Čekada, P. Pelicon, M. Payet, C. Grisolia
The differential cross-section for the 3He+3H nuclear reaction was measured in a thin tritiated PdTi film that was deposited on a Si wafer. The sample was loaded with 3H2 gas at a temperature of 300°C and at a pressure of 1.8 bar. The total activity of the sample, measured by the liquid scintillation technique, was found to be 578 MBq. Two peaks were observed in the spectrum of the thick Si detector
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Optimisation of ECX permeation barriers towards thicker alumina Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-06 Carsten Schroer, Julia Lorenz, Olaf Wedemeyer, Aleksandr Skrypnik, Kateryna Khanchych
The current development with respect to modifying the ECX (electrochemical X-deposition) process towards thicker alumina formed by heat treatment of aluminium electroplated on steel is described. As anticipated, changing the heat treatment atmosphere from argon to air or prolonging holding at high temperature (980 °C) increases the amount of alumina, but exposure to liquid lithium–lead eutectic reveals
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EXPERIMENTAL STUDY OF THE ISOTOPE EFFECT OF THE PERMEABILITY IN STRUCTURAL STEELS FOR FUSION REACTORS: EUROFER AND SS316 Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-05 María Urrestizala, Jon Azkurreta, Natalia Alegría, Igor Peñalva, Marta Malo, Carlos Moreno, David Rapisarda
The transport of hydrogen isotopes through the elements that make up a fusion reactor, and their corresponding interaction with these materials, have a direct impact on its operation. Consequently, it is essential to master in advance their transport parameters in the different materials proposed to constitute the elements that will compound the fusion reactors, and, consequently, this is one of the
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Effect of low-temperature neutron irradiation on the properties of titanium beryllide Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-06 A. Shaimerdenov, A. Akhanov, Sh. Gizatulin, A. Nessipbay, B. Shakirov, S. Askerbekov, T. Kulsartov, I. Kenzhina, A. Larionov, S. Akayev, S. Udartsev
Beryllium-based intermetallic compounds, such as Be12Ti, are increasingly being considered as a material capable of replacing pure beryllium in structural elements of fusion reactors. Be12Ti is considered as a neutron breeder material, a structural part of the Helium Cooled Pebble Bed of the DEMO reactor. It is expected that the replacement of beryllium by Be12Ti will make it possible to reduce the
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Comparative analysis of gas release from biphasic lithium ceramics pebble beds of various pebbles sizes and content under neutron irradiation conditions Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-06 Timur Kulsartov, Zhanna Zaurbekova, Regina Knitter, Inesh Kenzhina, Yevgen Chikhray, Asset Shaimerdenov, Saulet Askerbekov, Gunta Kizane, Alexandr Yelishenkov, Timur Zholdybayev
This paper presents the results of 4 reactor campaigns on the irradiation of biphasic lithium ceramics containing different ratios of lithium orthosilicate (LOS) and lithium metatitanate (LMT) components (25 and 35 mol% LMT in LOS). The size distribution of pebbles in pebble beds was 250–1250 μm and 500–710 μm, respectively. The studies were carried out sequentially with each type of ceramics. In experiments
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Helium swelling behavior for neutron multipliers after irradiation with He ions at high temperatures Nucl. Mater. Energy (IF 2.6) Pub Date : 2024-01-06 Yutaka Sugimoto, Mitsutaka Miyamoto, Jae-Hwan Kim, Taehyun Hwang, Masaru Nakamichi
Beryllides as beryllium intermetallic compounds are the advanced multiplier materials as candidate materials for fusion demonstration power plant (DEMO) due to having good properties such as low swelling rate and stable chemical properties. Under DEMO operation environment, the generated He amounts in beryllium are estimated around 20,000 appm He. This study was evaluated the effect of He fluence and
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Simulation and study of the milling parameters on CuFeTaTiW multicomponent alloy Nucl. Mater. Energy (IF 2.6) Pub Date : 2023-12-30 R. Martins, , J.B. Correia, , E. Alves, M. Dias
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Hyperspectral imaging and TRI3DYN simulation study of physical sputtering from a fuzzy surface Nucl. Mater. Energy (IF 2.6) Pub Date : 2023-12-30 F. Chang, D. Nishijima, M.J. Baldwin, W. Möller, G.R. Tynan
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Initial experiments to regenerate the surface of plasma-facing components by wire-based laser metal deposition Nucl. Mater. Energy (IF 2.6) Pub Date : 2023-12-28 Jannik Tweer, Robin Day, Thomas Derra, Daniel Dorow-Gerspach, Thorsten Loewenhoff, Marius Wirtz, Christian Linsmeier, Thomas Bergs, Ghaleb Natour
Plasma-facing components (PFC) in nuclear fusion reactors are exposed to demanding conditions during operation. The combination of thermal loads, plasma exposure as well as neutron induced damage and activation limits the number of materials suitable for this application. Due to its properties, tungsten (W) is foreseen as plasma-facing material (PFM) for the future DEMOnstration power plant. It is
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Production of 99Mo at IFMIF-DONES reusing the flux of neutrons Nucl. Mater. Energy (IF 2.6) Pub Date : 2023-12-26 E. López-Melero, F. Mota, M.I. Ortiz, F. Arias de Saavedra, I. Porras, L. Fernández-Maza, J. Praena
The accelerator-based neutron facility IFMIF-DONES is in the final design phase. At present, there is an important effort to optimize other secondary uses compatible with the main goal which is the validation of materials to be used in fusion reactors. One is the production of radioisotopes for nuclear medicine, in particular, Molybdenum-99. The decay of 99Mo produces Technetium-99 metastable. 99mTc