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| Fig. 1: Reactor vessel being shipped to Dresden Generating Station. (Source: Wikimedia Commons) |
One form of radiation that can cause material damage is neutron radiation, which is particularly problematic for specialty steels used in a fission reactor pressure vessel, like the one shown in Fig. 1. The vessel body consists of low alloy carbon steel, most commonly Mn-Mo-Ni, Mn-Mo-Cr, and Cr-Mo-V steels, which are all known for their high fracture resistance and reasonable ductility. [1] The interior surface of this vessel body is usually covered by another layer of steel, a three to ten millimeter thick layer of austenitic stainless steel, which is meant to prevent corrosion and contains chromium and nickel. [1] All of these steels are intended to last at least 40 years. [2]
Reactor vessel walls are primarily hit by slow neutrons, but they are also hit by fast neutrons, which are responsible for material degradation. [3] The incident neutron transfers energy to a metal atom, displacing it from its usual lattice site. [4] This atom then moves through the lattice and causes more atoms to become displaced, resulting in the creation of a displacement cascade. [4] Ultimately, point defects - vacancies and interstitials - are formed. [4]
Over the course of 40 years, the wall of a pressurized water reactor vessel undergoes a displacement change of around 0.05 displacements per atom (dpa). [5] For reference, one dpa signifies that every atom has been displaced once from its lattice site. [5] Most displaced atoms recombine with vacancies, and the remaining point defects affect the material's microstructure and mechanical properties. [4]
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| Fig. 2: Void swelling behavior of Ti-modified 316 stainless steel. (Data from Zinkle. [7] Source: A. Ravi) |
Point defects can pin dislocations in a metal, making it more difficult for the material to plastically deform and causing hardening. [4] As a result of hardening, the ductile-to-brittle transition temperature shifts, which can lower the material's lifetime. [3,6] At low temperatures, neutron irradiation can also lower fracture toughness. [5] At higher temperatures, neutron irradiation can induce the formation of multiple new phases in a material that is at first single-phase, like austenitic stainless steel. [5] Such phase formation is made possible by the diffusion of point defects, and the resulting redistribution of atoms in the metal is energetically favored. [3,4] On a related note, neutron radiation's ability to induce segregation of alloying elements can affect the material's corrosion resistance. [1] For example, Cr, an element known to improve steels' corrosion resistance, can be depleted in some areas of the steel due to radiation-induced segregation; these areas are prone to corrosion attack. [1] Yet another mode of neutron irradiation degradation is the swelling of voids (or vacancy clusters) in the metal. [5] As depicted in Fig. 2, void swelling tends to increase with displacement dose for Ti-modified type 316, an austenitic stainless steel containing Ti, Mo, Cr, and Ni. [7]
The mechanisms of neutron irradiation degradation are not completely understood, as there are limited studies on them. [3,6] Thus, future research should further investigate these mechanisms in an effort to extend the service life of fission reactor vessels. [3,6]
© Ajay Ravi. The author warrants that the work is the author's own and that Stanford University provided no input other than typesetting and referencing guidelines. The author grants permission to copy, distribute and display this work in unaltered form, with attribution to the author, for noncommercial purposes only. All other rights, including commercial rights, are reserved to the author.
[1] P. G. Tipping, Understanding and Mitigating Ageing in Nuclear Power Plants (Elsevier Science, 2010).
[2] C. K. Gupta, Materials in Nuclear Energy Applications: Volume I (CRC Press, 2017).
[3] G. R. Odette and G. E. Lucas, "Embrittlement of Nuclear Reactor Pressure Vessels," JOM 53, 18 (2001).
[4] J. Pu, "Radiation Embrittlement," Physics 241, Stanford University, Winter 2013.
[5] S. J. Zinkle and G. S. Was, "Materials Challenges in Nuclear Energy," Acta Mater. 61, 735 (2013).
[6] T. Dayrit, "Materials Challenges in Light-Water Reactors," Physics 241, Stanford University, Winter 2019.
[7] S. J. Zinkle, "Microstructures and Mechanical Properties of Irradiated Materials," in Materials Issues for Generation IV Systems, edited by V. Ghetta et al. (Springer, 2008).
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