6th Forum on New Materials - Part B

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Authors: Tammana Jayakumar, Ellappan Rajendra Kumar
Abstract: A detailed and comprehensive programme is undertaken in India to fabricate Indian Test Blanket Module (TBM) to be tested in ITER. Development of India-specific Reduced Activation Ferritic/Martensitic (RAFM) steel has been realized through melting and physical and mechanical properties characterization of several heats of 9Cr-RAFM steel with varying tungsten and tantalum contents. The RAFM steel having 1.4 wt. % tungsten and 0.06 wt. % tantalum is found to possess better combination of strength and toughness and is considered as India-specific RAFM steel. Different joining processes for fabrication of TBM have been assessed. Hot Isostatic Pressing (HIP) has been demonstrated to fabricate the first wall of TBM. To avoid channel collapse during HIPing, leachable ceramic cores were inserted in the channels. Electron Beam (EB) and Laser Welding processes are used for fabrication of breeder cassettes and hence, these welding procedures have been developed. Tungsten Inert Gas (TIG), Narrow Gap TIG (NG-TIG) and Laser Hybrid welding processes are being considered for integration of the various components such as first wall, back plate, bottom plate, breeder assembly and flow dividers into TBM. RAFM steel welding consumables have also been developed and qualified. Procedure for laser hybrid welding has also been developed. Necessary technologies for inspection and quality assurance of the fabricated TBM are also being developed. Use of ultrasonic C-Scan imaging to examine the bond integrity of the HIP joint has been demonstrated. Phased Array technique that would enable inspection of welds by longitudinal movement of the probe from an optimised lateral distance of the weld without the requirement of lateral movement has also been developed. The challenges in developing the India-specific RAFM steel and the fabrication and inspection technologies for fabrication of Indian TBM are presented.
Authors: Yan Yun Zhao, Shao Jun Liu, Chun Jing Li, Bo Yu Zhong, Gang Xu, Qun Ying Huang, Yi Can Wu
Abstract: China Low Activation Martensitic (CLAM) steel has been chosen as the structural material for China ITER Test Blanket Module (TBM). Creep-rupture and fatigue damage caused by high temperature and pulse stresses are two key issues for the final application of CLAM steel in China ITER TBM. In this paper, the research and development progress of the creep and fatigue behaviors of CLAM steel were presented. These results showed that CLAM steel possessed good high temperature mechanical properties.
Authors: Ludovic Charpentier, Marianne Balat-Pichelin
Abstract: The Gas-cooled Fast-Reactor (GFR) is one of the system developed in the frame of the 4th generation of nuclear plants. The helium coolant may contain some residual oxidizing impurities that can oxidize the cladding material. A sandwich material with a metallic layer (Ta or Nb) inserted between two sheets of SiC/SiC composite is a promising cladding system of the nuclear fuel to support mechanical strengths, to retain fission products without blocking the neutrons. Nevertheless SiC may support active oxidation with production of gaseous SiO and CO and sublimation in accidental conditions for temperatures above 2000 K, so the aim of the following study is to firstly investigate the oxidation resistance of the metallic liner in extreme conditions, alone and then covered by the SiC/SiC composite. The High Pressure and Temperature Solar Reactor (REHPTS) was implemented at the focus of the 6 kW Odeillo solar furnace in order to reproduce in helium atmosphere the sudden and huge temperature increase that can occur in the case of a nuclear accident. XRD and SEM enabled to observe that the high temperature exposure favored the (211) preferential orientation of tantalum and that the amorphization of niobium occurs above 1500 K. Such changes may impact the properties of the cladding elements. Some sandwich materials made of one plate of metal covered with a SiC/SiC composite were also tested and a carburation of the metallic plates at 2200 K was observed.
Authors: Kenji Konashi, Kunihiro Itoh, Tsugio Yokoyama, Michio Yamawaki
Abstract: Metal hydrides have high hydrogen atom density, which is equivalent to that of liquid water. An application of the hafnium hydride has been investigated as a neutron absorber in the Fast Breeder Reactors (FBRs). Fast neutrons are efficiently moderated by hydrogen in Hf hydrides and are absorbed by Hf. Since three isotopes of Hf have large cross sections, increase in the life of control rod is considered by Hf hydride. Results of design study of the core with Hf hydride control rods shows that the long lived hafnium hydride control rod is feasible in the large sodium-cooled FBR. Results of irradiation test conducted in BOR-60 has demonstrated the integrity of the capsules during irradiation. Na bonded capsule has an advantage in confinement effect of hydrogen compared with He bonded one. An application of hydride technique to transmutation target of MA was also discussed. MA hydride target is able to enhance the transmutation rate in FBR.
Authors: Michio Yamawaki, Yuji Arita, Takuya Yamamoto, Fumihiro Nakamori, Kazuhito Ohsawa
Abstract: Large amounts of depleted uranium kept as uranium fluoride or solid form after enrichment of natural uranium is sought to be utilized in the form of UNiAl intermetallic compound for hydrogen absorber. First principles calculation on UNiAl hydride has been performed in this study to predict the change of the crystal structure and the lattice constants with varying the hydrogen content. The results of the calculations have supported the experimental trends, suggesting that the present approach is promising in predicting the better hydrogen absorber based on depleted uranium.
Authors: Ji Yeon Park, Dae Jong Kim, Weon Ju Kim
Abstract: SiCf/SiC composites are one of the candidates for high temperature structural applications because of their high strength and corrosion resistance under severe conditions and stability under neutron irradiation [1~3]. A silicon carbide fuel cladding for the light water cooled reactors (LWRs) may allow a number of advances, including: the increased safety margins under transients and accident scenarios, such as loss of coolant accident; the improved resource utilization via a higher burn-up beyond the present limit of 62 GWd/MTU; and improved waste management [3~5]. Some components of SiCf/SiC composite will be applied as tubular geometry for the high-temperature core parts. The proposed design of an advanced LWR fuel cladding, referred to as Triplex, consists of three layers: an inner SiC monolith, a central SiCf/SiC composite, and an outer dense SiC evrionmental barrier coating. The inner SiC layer provides the strength and hermeticity to contain fission products. The SiCf/SiC composite layer fabricated by the CVI process provides a pseudo-ductile failure mode. The outer SiC thin coating layer protects against corrosion [5]. The chemical vapor deposition (CVD) technique is an effective approach for the fabrication of SiCf/SiC composite and coated SiC monolith [6]. To increase the homogeneity of the microstructure and the deposition rate of a SiC tube, the process parameters should be optimized and modified.
Authors: Leo Sannen, Sven van den Berghe, Ann Leenaers
Abstract: Historically, uranium enriched to >90% 235U has been used for many peaceful applications requiring high fission densities such as driver fuels for research reactors. However, the use of high-enriched uranium or HEU (all enrichments >20% 235U are considered HEU) for civil applications, is considered a proliferation concern. Since the 1970's, efforts are being devoted to the conversion of research reactors operating on HEU to alternative fuels using uranium with enrichment below 20% or LEU. These efforts imply the development of high-density LEU fuels to replace the low volume-density (mostly) UAlx based HEU fuels. The paper updates the present status of these developments focusing on the UMo dispersion fuel. It aims to provide an overview of the knowledge generated and the lessons learned in roughly 15 years of UMo dispersion fuel R&D in Europe through irradiation experiments and post-irradiation examinations (PIE).
Authors: Xiang Li, Ya Ting Yang, Cao Fei Fu, Qun Ying Huang, Liu Si Sheng, Zhen Qi Chang, Christophe C. Serra
Abstract: Porosity-controlled nuclear fuel microsphere is an essential material of fabricating minor actinide-bearing dispersion-type nuclear fuel with the infiltration processes. In this paper, monodisperse and size-controlled spherical oxide nuclear fuel particles with size range of 20μm to 800μm were fabricated by means of microfluidic technology combined with sol-gel process using cerium as a surrogate for plutonium. The porous CeO2 beads with the density range of 25% to 93% T.D. were successfully prepared by the addition of polyethylene glycol 6000 used as a porogen to the feed broth. The uniform U3O8 beads were also prepared at the same experimental conditions as CeO2 beads prepared, which shows the feasibility of the method for fabricating size-controlled monodisperse nuclear fuel beads.
Authors: He Fei Huang, De Hui Li, Long Yan
Abstract: The irradiation effects of a new nickel-base alloy (Ni-17Mo-7Cr) has been investigated by using 1 MeV Xe20+ and 7 MeV Xe26+ ions irradiation with displacement damage range from 0.33 to 6.6 dpa. The transmission electron microscopy and nanoindentation were employed to study respectively the microstructural evolution of thin-foil specimens and nanoindentation hardness changes of bulk specimens. In case of 0.33 dpa, high number density of nano-scale black spots were observed in thin-foil specimens. High-resolution transmission electron microscopy images revealed that these black spots are some rounded solute clusters and dislocation loops. As far as the ion dose of 3.3 and 6.6 dpa, the black spots were replaced with linear-like defects which were proved to be some Ni, Mo and Cr-enrichment regions. In addition, nanoindentation results for bulk specimens showed an evident hardening phenomenon in irradiated Ni-17Mo-7Cr alloys, compared to the unirradiated specimen. The irradiation induced defects may be responsible for the hardening of Ni-17Mo-7Cr alloys.

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