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Graphene-Based Nano-Composite Material for Advanced Nuclear Reactor: A Potential Structural Material for Green Energy
Published in Uma Shanker, Manviri Rani, Liquid and Crystal Nanomaterials for Water Pollutants Remediation, 2022
Materials that are commonly used for commercial reactors, are not suitable for the Gen-IV design due to their higher operating temperature as well as higher neutron dose. For a typical example, zirconium alloys are widely used as fuel cladding and other reactor components due to their low neutron absorption cross-section, moderate mechanical and corrosion resistance at high temperature (at an operating temperature of ª350°C) and aqueous environment (Murty and Charit 2008). With an increase in the operating temperature, zirconium alloys suffer from hydrogen embrittlement due to hydride formation, oxidation, allotropic phase changes and poor creep properties. Some of the outer core components (such as pressure vessel, piping, etc.) are typically made from low alloy steel, need to replace due to their poorer mechanical response and irradiation resistance at high temperatures. Several potential material candidates are listed in Table 2 and Figure 3 exhibiting promising performance as a structural material for Gen-IV reactors.
Properties and Uses
Published in Alan Cottrell, An Introduction to Metallurgy, 2019
In water and steam cooled reactors, corrosion in high-temperature water (e.g. 280°C at 70 atm. pressure) or steam at 500°C present special problems for canning materials. Zirconium (with hafnium removed) is the favoured material, usually in the form of zircalloy (1.5% Sn, 0.15% Fe, 0.05% Ni in Zr) or other alloys (e.g. 2.5% Nb in Zr). A recent trend is to add about 0.4% tellurium, which improves the resistance of zirconium to corrosion and absorption of hydrogen, in water at high temperatures and pressures.
Thermal Sensors
Published in John Vetelino, Aravind Reghu, Introduction to Sensors, 2017
Thermistors typically have resistance between 10 Ω and 100 MΩ. Due to the large resistance range, thermistors provide excellent sensitivity and are also very stable, particularly between 100 and 300°C. Thermistors can also be manufactured very cheaply. The two commonly used thermistors are the bead type and the disc type. In the bead type one starts with a platinum wire that is drawn or stretched. Small blobs of metal oxide are then attached with a binder at intervals along the wire. The wire is then sintered at about 1,300°C to produce a series of thermistors with embedded leads. The platinum wire is then cut on each side of the metal oxide thermistor, and the resulting individual thermistors are coated with a protective coating (glass). This results in a large number of thermistors with embedded leads that can be used as temperature sensors. In the disc type thermistor the metal oxide in powder form is compressed and heated to about 1,100°C to form a disc. Silver is then deposited on the opposite sides of the disc by spraying, screen printing, vacuum deposition, or RF magnetron sputtering. The disc can then be diced to form several individual thermistors. Comparing the two techniques, the bead type thermistors are the most durable and the cheapest to produce. Finally, for applications greater than 1,000°C thermistors made with zirconium oxide are the most popular. Zirconium is unique in its ability to withstand high temperatures.
Droplet impingement method to measure the surface tension of molten zirconium oxide
Published in Journal of Nuclear Science and Technology, 2020
Toshiki Kondo, Hiroaki Muta, Yuji Ohishi
Zirconium is used widely as the material for the control rods in nuclear power plants. When severe accidents occur, it is considered that zirconium is heated in air and oxidized to zirconium oxide; subsequently, the zirconium oxide melts and falls down the reactor. Therefore, a large of amount of molten corium made from zirconia is assumed to be present at such accident sites. It has been reported that 97 at. % of the debris generated by the severe accident in the three-mile-island was oxide, and 13.75 at. % of this oxide was zirconium oxide [3,4]. Consequently, to establish the physical properties of zirconium oxide, we reported upon the viscosity and surface tension of Zr1−xOx (x = 0, 0.1, 0.2) [5] and the viscosity of molten ZrO2 [6]. However, no experimental reports have reported on the surface tension of liquid ZrO2 because of the difficulty in performing the measurement at a high temperature.
The effect of adding lanthanum nitrate on anodizing process of zirconium-niobium alloy
Published in Inorganic and Nano-Metal Chemistry, 2020
Mohsen Asadi Asadabad, Ramin Shoja Gharabagh
One of the special properties of zirconium alloys is excellent corrosion resistance, which results in their widespread use in fission nuclear reactors.[1] At room temperature an oxide film with a thickness about 2–5 nm is formed on the surface of zirconium alloys.[2,3] Although, the oxide film improves corrosion resistance, at high temperatures the reaction of zirconium oxide with steam results in problems such as release of hydrogen gas. The penetration of this gas into the fuel rod and the formation of hydride phases will have a destructive effect on the performance of the fuel complex. Increasing the thickness of the oxide layer can improve the corrosion resistance and delay the formation of the hydride phase.[4] Anodizing is an electrochemical process that thickens oxide layers formed on active metals such as aluminum, zirconium, and tantalum.
Microstructure evolution of cold-worked Zircaloy-4 under 20 mev Cu ion irradiation
Published in Philosophical Magazine Letters, 2019
S. E. Naceri, M. Izerrouken, A. Ishaq, A. Guittoum, A. Sari, A. Khereddine, M. Kadouma, O. Menchi, M. Ghamnia
Zirconium-based alloys have found their greatest use in the nuclear industry, as a cladding and structural material in nuclear reactors, owing to their favourable properties such as good corrosion resistance, relatively high hardness and a low thermal neutron absorption cross-section (0.18 barn compared with 3.1 barn for iron). However, over a long-service period, the microstructure and mechanical properties of zirconium alloys change gradually in the harsh in-pile environment including a high stress level, a highly aggressive corrosive environment and especially a high irradiation dose (principally neutron irradiation, inducing irradiation damage). In order to design more tolerant cladding materials for aggressive environments such as high fuel burn up, many studies have been carried out to investigate and characterise the radiation-induced changes in the microstructure of neutron-irradiated Zr and Zr alloys at high temperature (>300°C) [1–7]. However, to the authors’ knowledge, few data are available concerning Zircaloy irradiated under research reactor operating conditions (T < 100°C) [8].