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| Fig. 1: A simple light-water reactor. (Source: Wikimedia Commons) |
The light-water reactor (Fig. 1) is the most common type of thermal-neutron reactor. The design was conceived nearly six decades ago, and involves normal water, as opposed to water containing heavy isotopes of hydrogen like deuterium, as its coolant and neutron moderator. [1] Now that the current fleet of nuclear plants are reaching the end of their 40-year initial lifespans and are looking toward 20- to 40-year extensions, scientists and engineers are responsible for closely examining how radiation has affected the function and safety of core materials in light-water reactors.
The stainless steel used as welds in light-water reactors are iron-chromium-nickel alloys. At standard temperatures and weight percentages of chromium, the steel has a dual-phase microstructure of ferrite and austenite. Main thermal aging effects associated with this steel is the decomposition of ferrite into iron-rich and copper-rich areas, and the growth of unwanted carbide molecules at the grain boundaries between phases. [2] This decomposition of steel drastically decreases the fracture toughness, or amount of mechanical energy able to be absorbed by the material before it breaks. For example, an ER316L stainless steel weld loses 43% of its ductile fracture toughness after being aged at 400 °C for 5000 hours. [2] Researchers must determine the extent of thermal aging in reactor steels before clearing the plants for extended lifetimes.
Typical steels for in-core materials in light-water reactors provide high creep strength, sufficient fracture toughness, and limited corrosion in the temperature window of the reaction. However, due to irradiation by neutrons released during the fission reaction, there is an induced shift of the ductile-to-brittle transition temperature that negatively impacts the lifetime of the material. [3]
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| Fig. 2: Caption goes here. (Source: Wikimedia Commons) |
When neutrons, ions, or helium bombard the steel, atoms within this steel are displaced; Fig. 2 illustrates the collision cascade mechanism by which a single incoming neutron can cause many more collisions. This creates vacancies and interstitial defects in the crystal lattice. These defects then agglomerate and extend to in order for the material to achieve the lowest energy configuration of the atoms. This interruption of long-range of crystallinity is responsible for hardening of the material because when these defects interact, they become pinned together and the material is, in a sense, trapped in its current crystal configuration. It loses ductility and is prone to fracture. [3]
Cesium and iodine are released as fission products in light-water reactors. The iodine then reacts with exposed zirconium in Zircaloy cladding, forming ZrI4. Then, through a phenomenon called the Van-Arkel vapor transport process, zirconium throughout the alloy is is transported as vapor, forming pits that lead to cracks. [4] Like other defects, these cracks reduce the toughness of the materials in the reactor. [4]
There are several other modes of degradation due to neutron irradiation and thermal aging. However, there are limited studies and contradictory results. In order to confidently increase nuclear power plant lifetimes, more research must be conducted into the aging of core materials.
© Teresa Dayrit. 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] D. Squarer et al., "High Performance Light Water Reactor," Nucl. Eng. Des. 221, 167 (2003).
[2] S. H. Hong et al., "Evaluation of the Effects of Thermal Aging of Austenitic Stainless Steel Welds Using Small Punch Test," Procedia Engineer. 130, 1010 (2015).
[3] K. Elrich et al., "Materials for High Performance Light Water Reactors," J. Nucl. Mater. 327, 140 (2004).
[4] P. S. Sidky, "Iodine Stress Corrosion Cracking of Zircaloy Reactor Cladding: Iodine Chemistry (A Review)," J. Nucl. Mater. 256, 1 (1998).
