Engineers and researchers with the U.S. Oak Ridge National Laboratory (ORNL) (Oak Ridge, Tennessee, USA) and GE Global Research (Schenectady, New York, USA) have developed a novel accident-tolerant fuel (ATF) cladding utilizing a new composition of an iron-chromium-aluminum (FeCrAl) alloy.
The multi-year, multi-faceted research, development, and deployment project has produced the nuclear-grade FeCrAl ATF cladding known as IronClad, which is now available in pre-commercial product forms and has been deemed acceptable for pilot testing.
The monolithic cladding was one of 10 winners selected by a panel of corrosion industry experts for the 2019 MP Corrosion Innovation of the Year Awards, honored in March 2019 in Nashville, Tennessee, USA. Further technical details and testing results are available at the awards web site.
Limitations of Zr-Based Alloys
Since the disaster at Japan’s Fukushima Daiichi Nuclear Power Plant in March 2011, scientists and engineers have searched for solutions to help prevent future accidents of that magnitude.
The chain of events at Fukushima began when a magnitude 9.0 earthquake off the coast of Tohoku caused a tsunami, which destroyed the ability to cool fuel elements in the reactors. The fuel’s cladding, a zirconium (Zr) alloy used to contain the fuel and radioactive fission products, reacted with the boiling coolant water to form hydrogen gas, which then exploded.
Conventional nuclear fuel cores have often used Zr-based alloys as cladding around the nuclear fuel, according to Jeremy Busby, the division director of ORNL’s reactor and nuclear systems division. “The role of the cladding is to contain the radionuclides of the fuel during both normal operation and during design basis accident scenarios,” Busby explains. “The choice of Zr-based alloys for this role is primarily due to the neutronic transparency of Zr, which increases the efficiency of the core, and the acceptable performance under normal operation through decades-long alloy optimization efforts by the nuclear industry.”
According to Busby, one problem with using Zr-based alloys is limited safety margins, such as time until the onset of severe core degradation under accident scenarios. The reason for such a limited time window is the rate of oxidation under high-temperature (more than 1,000 °C) steam environments, which leads to the production of hydrogen and further energy into the system.
An initial engineering response to reduce the severity of future Fukushima-like accidents was to try and increase the oxidation resistance of the cladding, Busby says. At first glance, coatings were determined to be the lowest cost and fastest-to-market solution. However, Busby believes uncertainties remain as to the overall performance of coated Zr-based cladding technologies.
New Monolithic Cladding Proposal
At ORNL, one solution under development since 2012 has been the use of a monolithic cladding composed of an FeCrAl alloy with an addition of yttrium (Y). In initial screening experiments taken after the Fukushima disaster, these alloys showed potential as a material system to replace Zr-based alloys. The increased oxidation resistance in an FeCrAl alloy results from the alloy’s ability to form aluminum oxide (Al2O3) in high-temperature steam environments.
The problem earlier in the decade was that commercially available FeCrAl alloys were not optimized for nuclear-grade fuel cladding, owing to the expected fuel efficiency reduction associated with increased neutron absorptions.
Fuel cladding performance characteristics are multi-dimensional, Busby explains, with neutronic performance, mechanical performance, radiation tolerance, and fabricability as key variables. Preliminary investigations found that a balance of Cr and Al, as well as minor solute additions, were needed to begin meeting the stringent demands for FeCrAl use in a nuclear core.
Kevin Field, a research staff member at ORNL, notes that FeCrAl alloys exhibit a low-temperature miscibility gap at or below 500 °C, which can lead to deleterious precipitation and material embrittlement under thermal aging and irradiation. Nuclear fuel cladding operates within this miscibility gap of between 290 °C and 330 °C. The severity of the precipitation is reduced with lower Cr and increased Al contents in the alloy.
This balance is in direct competition, however, with high-temperature steam oxidation resistance—where low-Cr variants may not remain protective in accident scenarios, depending on Al content. Similar compositional dependencies exist for other performance factors, Field says. These include thin-wall tube fabrication and mechanical performance, such as fracture toughness, and welding.
According to Busby, the group’s innovation lies in optimizing both the composition and microstructure for performance in a nuclear power plant under both normal operation and accident scenarios. Recently, ORNL, in collaboration with GE’s Global Research Center and Global Nuclear Fuel venture, completed a research project to develop the new FeCrAl alloy, known commercially as IronClad, for nuclear fuel applications.
Key Experimental Findings
The group explains that the new alloy has a balanced level of alloying additions and optimized processing. This has allowed them to commercially produce cladding deemed “accident tolerant,” while still satisfying the normal operation demands of nuclear fuel cladding.
Busby explains that several key experimental findings underpinned the development of their alloy:
1.) Protective alumina surfaces can be formed in elevated oxidizing environments on FeCrAl alloys only when the weight percent of Al and Cr are between 3 to 6 wt% and greater than 12 wt%, respectively. Al, with the addition of Y, helps to establish and stabilize the alumina surface under oxidizing environments, with a composition promoting a fully ferritic microstructure.
2.) The unique Cr and Al concentration volume means that at lower temperatures (~300 °C), the FeCrAl alloy is a chromia former and not an alumina former. This promotes protective chromia formation in water chemistry environments typical of light water reactors, thus preventing long-term, aqueous-based corrosion of the cladding.
3.) Good radiation tolerance was achieved by lowering the Cr to 12 wt% and increasing the Al content to 6 wt%. Fabrication steps were taken to promote a fine grain structure, thereby enhancing the alloy’s radiation tolerance.
4.) The composition control for the new alloy also enables the ability of the material to be warm drawn and produced into final form factors suitable for commercial use. According to the group, elevated Cr or Al contents would lead to cracking under scenarios such as drawing or welding.
According to Busby, most prior FeCrAl alloys with Y additions did not have the appropriate weighting to be suitable for both oxidation resistance at high temperatures and the mechanical properties and ease of fabrication needed during normal operations.
“The research effort found that the leaner compositions could not form alumina in steam at 1,200 °C,” Busby says. “However, for radiation resistance, low Cr contents were desirable. The alloy development effort identified that above 6 wt% Al, the alloy was too brittle to be fabricated into thin tubing. Thus, the current composition of 12 wt% Cr and 6 wt% Al was unlike any previous FeCrAl composition and found to possess the best oxidation resistance.”
Remarkably, Busby says the new alloy forms protective alumina at up to 1,500 °C during accident scenario simulations. “The alloy melts at ~1,520 °C, so its oxidation resistance in steam is outstanding,” he explains. “The composition is also corrosion resistant in the normal operating conditions of a light water reactor of pressurized water at ~300 °C.”
In contrast to austenitic steels, researchers say that ferritic steels like IronClad are not susceptible to stress corrosion cracking and have been tested in autoclaves for more than one year to prove their durability during normal operation.
Test Results Appear Promising
The alloy was extensively tested in laboratory and simulated field trials, with strong results found in performance factors such as mechanical strength, radiation tolerance, and fabrication.
To test the alloy on a larger scale, a test involving more than 60 m of the fuel cladding rods was performed at Karlsruhe Institute of Technology’s QUENCH facility in Karlsruhe, Germany. There, the test assembly was heated using electrical heaters inside the cladding to simulate heat from radioactive decay, all while also being subjected to high-temperature flowing steam. The test parameters were conducted to mimic multiple previous tests on Zr-based cladding and were subsequently extended into accident regimes far more severe than previous Zr-based tests.
In a direct comparison to Zr-based cladding, the new material experienced almost no discernable oxidation, according to Kurt Terrani, lead on the QUENCH test and ORNL's senior staff scientist. Even after extending the test into power and time regimes well beyond what is possible with Zr-based cladding, the IronClad material experienced limited oxidation and produced significantly less hydrogen gas.
The pre-commercial product form is now being tested in both simulated environments in materials test reactors and within commercially operating nuclear power reactors, with successful results reported thus far. (Additional testing results are available at the awards web site.)
The cladding can be retrofitted into the otherwise normal production of fuel assemblies or customized for a given application, according to Busby, who adds that the product has been designed to have no significant impacts on a reactor’s performance.
The IronClad product is scheduled for continued deployment within commercially operating nuclear reactors as a pilot program for full-scale deployment by the global nuclear fleet.
Sources: Oak Ridge National Laboratory, www.ornl.gov; GE Global Research/Global Nuclear Fuel, www.ge.com/research.