2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

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1 FRESH WATER QUENCHING OF ALLOYS OF NUCLEAR INTEREST INCLUDING FeCrAl FOR ACCIDENT TOLERANT FUEL CLADDING Michael Schuster 1, Cole Crawford 1, and Raul B. Rebak 1 1 GE Global Research: 1 Research Circle, Schenectady, NY USA, rebak@ge.com ABSTRACT: In the US, the Department of Energy is funding research performed at commercial fuel vendors to develop advanced technology fuel (ATF) also known as accident tolerant fuel (ATF). General Electric and Oak Ridge National Laboratory are evaluating the simplest and most near term idea of replacing the current zirconium based cladding using an iron-chrome-aluminum (FeCrAl) alloy. The current work investigates the behavior of FeCrAl alloy and other nuclear alloys of interest under fresh water reflooding conditions following a high temperature excursion. It is important to determine the integrity of the cladding material after being quenched in water. Tests were performed for six alloys (Zircaloy-2, austenitic type 304SS, and Alloy 600, and ferritic 4C54, T91 and APMT) which were exposed for 2 h at 1200 C in air, argon or steam and then quenched in ambient temperature water. Results show that argon only produced transformation in the zirconium alloy. When tested in air and steam several alloys suffered extensive damage. APMT was resistance to attack by the development of a protective alumina scale on the surface, which remained adherent upon quenching. KEYWORDS: Accident Tolerant Fuel, FeCrAl, Quench, Oxidation. I. INTRODUCTION Following the March 2011 earthquake and tsunami in Japan, which caused plant black out in several of the stations of the Fukushima plant, the international community decided to find alternatives to the zirconium alloy cladding for uranium dioxide fuel in the light water reactors. In the Fukushima accident, the lack of cooling caused the zirconium in the cladding of the fuel to react with water to form large amounts of heat and combustible hydrogen gas. Zr + 2H 2O ===== ZrO 2 + 2H 2 (gas) + Heat The US Department of Energy is funding research at several commercial nuclear fuel vendors. 1 Several concepts are under consideration. The simplest and more near term concept is to use iron-chrome-aluminum alloy (FeCrAl) to replace zirconium alloys as cladding material, keeping the same uranium dioxide fuel. The main benefit of the FeCrAl alloys is their outstanding resistance to oxidation in a case of an accident, that it, it would generate little heat of reaction and little combustible hydrogen gas. The slower oxidation rate of FeCrAl compared with zirconium alloys would increase the time allowed for the utilities to cool it (quench), probably by flooding the reactor with fresh water. Therefore, it is important that the alloy of the cladding would withstand the insertion of fresh water after being exposed to high temperatures for some time. It is important that during the flooding with fresh water the cladding will contain the fuel and will not release radionuclides to the water and the environment. FeCrAl Alloys Properties The objective of the GE-DOE cost shared project is to develop an iron-chromium-aluminum (FeCrAl) fuel cladding for current design light water power reactors. 2 The idea of using FeCrAl alloys as cladding for current UO 2 fuel is also supported by Oak Ridge National Laboratory (ORNL), who developed the alloy C26M, which is a current candidate with Kanthal APMT. Besides Fe, Cr, and Al, the cladding may contain other elements such as molybdenum, yttrium, hafnium, zirconium, titanium, etc. The nominal composition of APMT is Fe + 21 Cr + 5 Al + 3 Mo + traces of other elements and the nominal 1

2 composition of C26M is Fe + 12 Cr + 6 Al + 2 Mo + traces. APMT is manufactured using powder metallurgy and C26M is manufactured by traditional melting. It has been known since the 1930s that FeCrAl alloys are incredibly resistant to high temperature oxidation by air and steam. Figure 1 shows the process of how this alloy resists attack by superheated steam. Under normal operation conditions and up to 1000 C the protection to the alloy is given by the formation of a chromium rich oxide on the surface. However, as the temperature increases beyond 1000 C, an aluminum oxide layer (alumina) forms between the metal and the chromium oxide layer. Eventually, as the temperature increases further in the presence of steam, the chromium oxide layer volatilizes and the alumina layer remains on the surface protecting the alloy from additional oxidation up to the melting point (~1500 C) of the alloy. Figure 2 shows the presence of an approximately one micron thick layer of alumina on the surface of Aluchrom YHf after exposure for 4 h at 1200 C in 100% steam. The composition of Aluchrom YHf is nominally Fe + 20Cr + 6Al + 0.5Si (max) + 0.1Hf (max) + 0.1Y (max). The oxide in Figure 2 does not contain chromium or iron. Some particles are rich in Hf, Zr, Si and Y. Fig. 1. Oxidation behavior of FeCrAl alloys in presence of steam. Fig. 2. Alumina layer formed on the surface of Aluchrom YHf after exposure to 100% steam at 1200 C for 4 h. 2

3 Figure 3 shows the presence of an approximately one micron thick layer of alumina on the surface of a APMT tube after exposure for 4 h at 1200 C in 100% steam. The oxide in Figure 3 does not contain chromium or iron. Some particles are rich in Ti, Hf, Zr, Si and Y. Fig. 3. Alumina layer formed on the surface of APMT Tube, 0.8 mm wall thickness and 9.5 mm OD after exposure to 100% steam at 1200 C for 4 h. Figure 4 shows the presence of an approximately one micron thick layer of alumina on the surface of a APMT plate after exposure for 8 h at 1050 C in air. This treatment was performed in the Kanthal plant in Sweden. The oxide in Figure 4 contains a thin layer of chromia in the outer part of the alumina layer. The image in Figure 4 is the typical treatment performed by Kanthal as a pre-oxidized product. II. EXPERIMENTAL RESULTS Table I shows the six nuclear materials tested in the same conditions to determine behavior under fresh water quenching TABLE I. Alloys Tested for 2 h at 1200 C in (1) Argon, (2) Air, and (3) Steam + Water Quench Alloys Nominal Composition (wt %) Zirc-2 >99% Zr 600 Ni + 16Cr + 9Fe APMT Fe + 21Cr + 5Al + 3Mo T91 Fe + 9Cr + 1Mo 4C54 Fe Cr + 0.2N 304SS Fe + 19Cr + 10Ni 3

4 Fig. 4. Alumina layer formed on the surface of APMT plate after pre-oxidation treatment in air at 1050 C for 8 h at the production plant. Tensile specimens were exposed side by side with witness coupons for determination of oxidation behavior. The tensile specimens were mechanically tested at ambient temperature after the pre-exposure in argon, air or steam for 2 h at 1200 C. Figure 5 shows a typical temperature profile of the treatment underwent by each tensile specimen and witness coupon. Figure 6 shows the geometry of Zircaloy-2 and APMT specimens tested in air, before and after testing. The mechanical properties of the materials after quenching is reported elsewhere. 3 Figure 7 shows the tensile specimen after failure both for the as-received conditions (left) and the ones pre-exposed in air for 2 h at 1200 C plus water quenched (right). Figure 7 shows in a comparative manner the total elongation for each alloy post quenching as compared to a material which did not undergo high temperature exposure. Figure 7 shows that the austenitic materials (type 304 SS and Alloy 600) have higher elongation to failure than the ferritic (4C54, T91 and APMT) materials. Table II shows the appearance of the cross section of the witness coupons for the six alloys tested in argon, air and steam. In the argon atmosphere, the only material that suffered some type of degradation was Zircaloy-2, which had a small amount of oxide on the surface and a layer of metal phase transformation probably due to the diffusion of trace oxygen into the metal. Also, type 304 SS had some small surface degradation (Table II). In the air atmosphere, the zircaloy-2 coupon was almost completely consumed by oxidation. Other coupons which suffered some damage were T91 (low Cr content), type 304SS and 4C54. APMT was free of external attack in air and in argon. In the steam atmosphere, Zircaloy-2, type 304SS and T91 suffered severe attack. The attack front in Zircaloy-2 can be separated in several regions, two oxide layers separated by a string of precipitates (Figure 8), a phase transformed zone containing voids, and a phase transformed zone without voids (Table II). Table II shows a conspicuous anomalous gap on the surface of APMT exposed to steam and quenched. More studies are being performed to determine if this occurrence was the results or an uneven bimodal grain distribution or string of precipitates (Figure 9). The microstructure of APMT corresponding to a fabricated tube is more uniform, finer grained. 4

5 High SEM magnification of the APMT exposed in steam (Figure 9) shows a compact layer of alumina protecting the material from the environment (this is expected). That is the alumina remains adherent to the coupon upon quenching. Fig. 5. Typical temperature profile for the tensile specimen and witness coupon in argon, air and steam. Zircaloy-2-7 Specimens exposed to air for 2 h at 1200 C and then water quenched. APMT-4 Specimens exposed to air for 2 h at 1200 C and then water quenched. Fig. 6. Specimens and coupons used for the quench tests. 5

6 Fig. 7. Post tensile test specimens for the six alloys. As received (left) and after exposure to air at 1200 C for 2 h plus water quenching (right). III. DISCUSSION The goal of the U.S. Department of Energy Accident Tolerant Fuel (ATF) program is to develop a fuel that will be tolerant to accidents as the one at the Fukushima site in The US DOE is funding research on ATF at three fuel vendors, national laboratories, and at several universities. GE and Oak Ridge National Laboratory are proposing a simple concept of replacing the current zirconium alloy using an FeCrAl alloy such as C26M or APMT. Since 2013 extensive characterization has been performed on the behavior of FeCrAl in five specific areas outlined by DOE. 1 These areas included material behavior under normal operation condition, under design basis accident, beyond design basis, fabrication and commercialization and used fuel disposition or reprocessing. The current work is investigating what would happen in the case of a loss of coolant accident in which the fuel rods may overheat and are then flooded with fresh water. It is a requirement that the cladding of the rods remain sound and leak free to avoid the release of radionuclides into the coolant or the environment. Current studies show that FeCrAl alloys can accept quenching while remaining cracking free. Moreover, the alumina layer which forms on the high temperature excursion remains adherent and protective of the FeCrAl alloy. The studies in the quenching area are continuing, with the addition of the C26M alloy and APMT with a tubing microstructure. 6

7 TABLE II. Witness coupons after testing for 2 h at 1200 C in (1) Argon, (2) Air, and (3) Steam + water quenched. The magnification of the images may vary from one to another. Alloy Argon 2 h at 1200 C + WQ Air 2 h at 1200 C + WQ Steam 2 h at 1200 C + WQ Zirc APMT T91 4C54 304SS 7

8 Fig. 8. Specimens and coupons used for the quench tests. APMT coupon exposed to steam at 1200 C for 2 h and water quenched. A protective 1 µm thick alumina film is observed on the surface (insert). Bimodal microstructure of the APMT square billet used to fabricate the coupons and tensile specimens. The microstructure of the tubing will be finer grained and more uniform (equiaxed). Fig. 9. Oxidation Characteristics of the APMT coupon exposed to steam at 1200 C for 2 h and water quenched. 8

9 IV. CONCLUSIONS The goal of the U.S. Department of Energy Accident Tolerant Fuel Program for Light Water Reactors is to identify alternative fuel technologies that enhance the safety, competitiveness and economics of nuclear power. It is expected that accident tolerant fuels will endure loss of cooling in the reactor core for longer time than the current generation of fuel. The ATF design concept proposed by General Electric and Oak Ridge National Laboratory utilizes an FeCrAl alloy (such as APMT or C26M) for fuel rod cladding in combination with the current generation uranium dioxide fuel pellets. Currently six alloys were exposed for 2 h at 1200 C in three environments (argon, air and steam) and then quenched in water. All the tested materials (even T91) showed in general a superior behavior than the zirconium based alloy. APMT showed a high resistance to oxidation by the formation of an alumina layer. This alumina layer on the surface remained adherent upon quenching. More studies are in progress to determine that APMT and C26M can accept the sudden presence of fresh water without suffering embrittlement (i.e. will retain fission products within the cladding). ACKNOWLEDGMENTS The authors would like to thank the characterization lab at GE Global Research, including Mike Larsen, Jae-Hyuk Her and Ian Spinelli. Great thanks to Lisa Sciubba at Lucideon (Schenectady, NY) for metallographic cross sections. The funding support from GE Hitachi is gratefully acknowledged. This material is based upon work supported by the Dept. of Energy [National Nuclear Security Administration] under Award Number DE-NE This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. REFERENCES 1. S. M. BRAGG-SITTON, M. TODOSOW, R. MONTGOMERY, C. R. STANEK, R. MONTGOMERY, AND W. J. CARMACK, Metrics for the Technical Performance Evaluation of Light Water Reactor Accident Tolerant Fuel, Nuclear Technology, 195(2), p , August R. E. STACHOWSKI, R. FAWCETT, R. B. REBAK, W. P. GASSMANN, J. B. WILLIAMS, K. A. TERRANI, WRFPM 2017, Jeju Island, Korea, September M. SCHUSTER, C. CRAWFORD, AND R. B. REBAK, 18 th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors, August 13-17, 2017, Marriott Portland Downtown Waterfront Portland, Oregon, USA. 9