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

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1 OVERVIEW OF WESTINGHOUSE LEAD ACCIDENT TOLERANT FUEL PROGRAM Robert Oelrich 1, Sumit Ray 2, Zeses Karoutas 2, Ed Lahoda 3, Frank Boylan 3, Peng Xu 2, Javier Romero 2, Hemant Shah 2 1 Engineering Center of Excellence, Westinghouse Electric Company, Columbia, SC , oelricrl@westinghouse.com 2 Westinghouse Electric Company, Columbia, SC Westinghouse Electric Company, Cranberry Township, PA ABSTRACT: Westinghouse is commercializing two unique accident tolerant fuel (ATF) designs; silicon carbide (SiC) cladding with uranium silicide (U 3 Si 2 ) fuel, and chromium-coated zirconium alloy cladding (CZA) with U 3 Si 2 fuel. Testing of the cladding alternatives in autoclaves has been concluded and samples have begun irradiation at the Massachusetts Institute of Technology (MIT) Reactor and the Halden Project Reactor. Uranium silicide fuel is undergoing exposure in the Advanced Test Reactor and fuel pins have been removed and are undergoing post irradiation examination at the Idaho National Laboratory. Testing results to date have confirmed anticipated results, and a business case has been developed on the value of ATF for both fuel costs and operational savings. In parallel, a licensing approach has been developed and is under discussion with the US Nuclear Regulatory Commission (NRC). This paper provides an update on these activities and a summary of results. KEYWORDS: SiC, silicon carbide, Cr, chromium, coating, zirconium, uranium, silicide, U 3 Si 2, licensing, testing. I. INTRODUCTION AND BACKGROUND The first introduction of Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program lead test rods (LTRs) planned for 2019 utilize Cr coated zirconium alloy (CZA) cladding with U 3 Si 2 high density/high thermal conductivity pellets. The lead test assembly (LTA) program, planned for 2022 insertion, will introduce SiC/SiC composite cladding along with CZA cladding and the high density/high thermal conductivity U 3 Si 2 pellet. Over the past several years, Westinghouse s ATF test program has tested the CZA and SiC claddings in autoclaves and in the MIT reactor and U 3 Si 2 pellets in the Advanced Test Reactor (ATR). A Westinghouse ATF paper on the corrosion testing of SiC can be found in the same Top Fuel meeting by Xu et. al [Ref. 9]. Westinghouse partner General Atomics also published a paper by Deck et al. on SiC cladding with an emphasis on manufacturing and characterization, which complements this paper in other key areas of the development at this stage [Ref. 8]. In addition, ultra high temperature tests at the state-of-the-art facilities in Churchill, PA have been carried out to confirm the time and temperature limits for the SiC and CZA claddings. Finally, an extensive research program to fully understand the science behind ATF materials continues with the Westinghouse lead, international Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) group and at the US national laboratories. Based on the positive results to date, fuel rod and assembly design in preparation for the LTR and LTA programs is underway as well as licensing efforts with the NRC, and accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. II. TESTING OF WESTINGHOUSE ATF CLADDINGS II.A. Autoclave Corrosion Testing of ATF Claddings Westinghouse has performed corrosion testing using the autoclave facility at the Churchill, PA site to screen various coatings and SiC preparation methods for corrosion resistance. As part of a multi-year program, over 12 types of coatings on zirconium alloys and approximately 10 versions of SiC have been tested in autoclaves. As a result of this testing, two coatings (Table I) were identified for testing in the MIT reactor. 1

2 Testing in the MIT reactor further narrowed the options to the Cr coating (Fig. 1.). The CZA showed no signs of peeling and had minimal weight gain after taking into account the uncoated inner surface of the tube. Based on these very positive test results, Westinghouse is now exploring methods for production of full length rods for LTRs to be constructed in 2018 for inclusion in a commercial reactor in early TABLE I. Autoclave corrosion performance for the top zirconium alloy coatings Material Process Vendor Maximum Days Average Corrosion rate (mg/dm 2 /day) Std. Dev. Corrosion rate (mg/dm 2 /day) Average Zr Corrosion (mg/dm 2 /day) TiN/TiAlN PVD* PSU** Cr Cold spray UW* *Physical Vapor Deposition **Pennsylvania State University ***University of Wisconsin Corrosion Rate (microns/year) Fig. 1. Cr coated zirconium alloy tubes before and after testing in the MIT reactor (Ref. 1) Initial autoclave and reactor testing resulted in relatively high levels of SiC corrosion. Autoclave testing with hydrogen peroxide was used to simulate the more aggressive oxidation conditions of the reactor and to explore coolant conditions that would minimize SiC corrosion rates. This testing has been used to refine the manufacturing parameters of the SiC composites (Ref. 8,9) such that along with hydrogen addition to the primary coolant above 40 cc/kg (Ref. 2), the current corrosion rates for SiC meet or exceed the target 7 microns/year recession rate. For a full core of SiC cladding, this would result in a maximum of 150 kg of SiO 2 or about 350 ppm over an 18 month cycle. This is well below the solubility limit of ~700 ppm SiO 2 at the coldest steam generator conditions. Note also, that resins are commercially available that could be added to the current resins used to maintain water chemistry to remove SiO 2 on a continuous basis. Westinghouse continues to assess the potential for crud buildup on advanced ATF claddings. Limits on crud buildup on SiC cladding may be different than for coated claddings because the SiC surface may be corroding from under any potential crud buildup. Therefore, WALT loop (the high heat transfer rate heat transfer and crud deposition loop at Churchill, PA) testing is being planned for mid-2017 to study heat transfer rates and crud buildup on the SiC cladding surface. II.B. High Temperature Testing of ATF Claddings The goal of the ATF program is to develop fuels that can withstand post-accident temperatures greater than 1200 C without the cladding igniting in steam or air. Therefore a crucial part of the testing carried out by Westinghouse over the previous year was aimed at quantifying the maximum temperature at which the ATF claddings could operate without 2

3 2017 Water Reactor Fuel Performance Meeting excessive corrosion. The test apparatus (Fig. 2.) currently uses a graphite rod which is inserted into insulation and then into the test piece. This results in a very stable heating of the test pieces. CZA has now been run at up to 1400 C. This is above the Cr-Zr low melting eutectic point of 1333 C. At 1400 C, there was noticeable reaction between the Cr and the Zr. However, there was not the rapid oxidation that uncoated Zr experiences at 1200 C. At temperatures of 1300 C, however, the Cr coated zirconium alloy was stable for reasonable lengths of time. Combined with the lowering of zirconium oxidation at normal operating temperatures which vastly reduced the formation of zirconium hydrides and therefore embrittlement, the CZA provides significant performance improvements during normal operation, transients, design basis accidents and beyond design basis accidents as compared to uncoated zirconium. Similar tests were run with SiC at temperatures from 1600 C up to 1700 C. These tests were terminated only because of excessive corrosion of the heater element. At 1600 C, the SiC cladding was visually untouched. At 1700 C, there were indications of small beads on the surface (presumably SiO2 from the reaction of SiC with steam) but on the whole, no significant deterioration of the SiC. Changes are being made to the heating rod to increase the flow of He cover gas and to allow accurate weight changes to be made on the SiC rodlets so that kinetic data can be obtained. Fig. 2. SiC rodlet undergoing testing in the ultra-high temperature apparatus in steam at 1600 C at Churchill (The sample is mounted inside the shield tube that is glowing white in the photograph. The SiC tube is inside the shield tube with steam injected both above and below the sample. The steam exits through the hole that is visible in the shield tube.) III. TESTING OF WESTINGHOUSE U3Si2 ATF HIGH DENSITY FUEL U3Si2 is a revolutionary material for LWR fuel service, and so considerable data is required on the behavior of U3Si2 at LWR operating temperatures (estimated to be from 600 C and up to 1200 C during transients). To obtain the necessary data, U3Si2 fuel pellets were manufactured at INL and put into rodlets in the ATR in The first rodlets came out of ATR at the end of 2016 (Fig. 3.) and are due for destructive post irradiation examination in the summer of 2017 at INL (Ref. 3). U3Si2 was tested for air and steam oxidation and compared to UO2 using digital scanning calorimeters at both the Westinghouse Columbia facility (Ref. 4) and at Los Alamos National Laboratory (LANL) (Ref. 5). The Westinghouse results indicate that the ignition temperatures for UO2 and U3Si2 are between 400 C and 450 C (Table II and Fig. 4.). The LANL results (Fig. 5.) indicates an ignition temperature of about 400 C. The reasons for this difference are being studied. The heat and mass generated by the oxidation of the U3Si2 is considerably higher than for UO2. The effect of this difference 3

4 in heat release and mass on the stability of the rods is being investigated in rodlet tests in the Churchill autoclaves in the summer of It is noted however that ATF cladding surfaces are much harder than zirconium alloy cladding and grids, so it is expected that the likelihood of grid to rod fretting leakages will be greatly reduced from the current 1 to 2 ppm. The power from this reaction would also be very small compared to the power from the rod under operating conditions. Fig. 3. Neutron radiographs of 20 MWd/kgU U 3 Si 2 Pins from ATR (Note the lack of pellet cracking and distortion.)(ref. 4) Finally, LANL identified the potential for the formation of a U 3 Si 2 -H 1.8 compound in the event of a leaker. Further work reported by S. Mašková et al. (Fig. 6.) (Ref. 6) indicated that this would not likely be an issue since the operating temperature of the U 3 Si 2 fuel will be above the decomposition temperature (~525 C) of this compound. TABLE II. Comparison of Westinghouse U 3 Si 2 and UO 2 scanning calorimeter results Material UO 2 U 3 Si 2 Max. Δmass (%) 4.1% 25% Oxidation T ( o C) ~410 ~450 Reaction enthalpy (-J/g) ~200 ~3200 Heating rate 2.5 o C/min Air The oxidation is exothermic Steam. 4

5 Steam 2017 Water Reactor Fuel Performance Meeting Fig. 4. Westinghouse thermogravimetric results for U 3 Si 2 for steam versus air Ai Heating rate 2.5 C/min. Fig. 5. Thermograms depicting the sensitivity of these onset measurements to the sample geometry/form and ramp rate Fig. 6. Temperature dependent solution energy when considering the vibrational entropy of an H 2 molecule (left). Experimental observation of H 2 evolution with a peak at ~713 K from (Ref. 6). IV. ACCIDENT SCENARIO EVALUATIONS AND LICENSING IV.A. Accident Scenario Evaluations MAAP5 calculations were performed for CZA and SiC claddings along with high density fuels for the station blackout scenario and the TMI2 small break LOCA with replenishment of the primary coolant (Ref. 7). The CZA option offers modest ATF gains (~200 C) before large scale melting of the core begins in beyond design basis events such as a long term station blackout. Though it does not prevent the contamination of the PWR primary loop due to ballooning and bursting at about 800 C to 900 C, the CZA option can prevent a TMI2 type of accident from extending into the fuel melt-down phase and prevent extensive contamination of the containment and perhaps preserve the nuclear plant. This is because although the CZA may begin to fail as the temperature exceeds 1400 C due to eutectic formation, it does not rapidly oxidize as uncoated zirconium alloys do and does not provide a rapid energy input spike into the core (Fig. 7). Note that in this case, FeCrAl was used to model the performance of CZA since their temperature and oxidation performance is 5

6 about the same. The results for the station blackout scenario (Fig. 8.) indicate that fission products can be contained within SiC cladding for up to two hours longer than current Zr based cladding due to its higher temperature capability (~2545 C decomposition temperature). These two hours can be used to implement additional responses by the operators. The lower pressure in the system due to minimal hydrogen production (Fig. 9.) increases the chances that alternate means to feed cooling water to the core at about 40 gpm can result in avoidance of fuel melting, indefinitely extending the coping time as long as the water flow continues. The SiC cladding of course prevents any leakage of fission products into the primary loop since it will not balloon and burst. Due to the short timespan before coolant was re-introduced to the system, the SiC cladding would have had no adverse consequences from a TMI2 type accident (Fig. 7). Fig. 7. Hottest core node for TMI-2 accident where coolant restored at ~9900 seconds Fig. 8. Hottest core node for PWR station blackout 6

7 IV.B. LTR and LTA Design and Regulatory Approval Fig. 9. Total hydrogen generated for PWR station blackout With utility support for the Westinghouse ATF program, commercial plants have been identified in which to run lead test rods in Since CZA is a similar material to the current zircaloy cladding, it is expected that the coating can be implemented without having to make significant changes to the licensing basis. To confirm that change can be implemented without prior NRC review and approval, the functional groups will need to perform an evaluation which will include an assessment of the impact of the change as well as a demonstration that the coating is not a new material (a new material requires an exemption from 10 CFR 50.46). The lead test assembly planned for 2022, which will add SiC cladding, will require a license amendment request (LAR) for the plant. Due to the longer implementation schedule, an exemption request for an LTA of SiC may be required. There is currently ongoing rulemaking to change 10 CFR Once this rulemaking is complete an assessment will need to be completed to determine if an exemption is required for SiC LTAs. Implementation of U 3 Si 2 pellets in LTAs is expected to require an exemption from 10 CFR Additional work will need to be completed to determine if an LAR will be required to implement LTRs of U 3 Si 2. V. ECONOMIC EVALUATIONS OF WESTINGHOUSE ATF FUEL BASED ON CURRENT TESTING RESULTS V.A. Economic Evaluations The proposed Westinghouse accident-tolerant fuel cladding designs also are expected to deliver improved fuel cycle economics as a result of the higher density and thermal conductivity of the U 3 Si 2 fuel pellets. Significantly more U235 can be packed into the same volume as UO 2, enabling longer fuel cycles, while also staying below the five percent enrichment limit that is the design basis for many operating plants. These improved uranium efficiencies can save an estimated $4M per cycle through reduced uranium costs. Furthermore, other potential savings in plant operations and maintenance that could mean an addition $5M to $12M of savings per plant per year have been identified, but remain to be verified. V.B. Research and Development Activities and Required Funding Government and industry will have to support significant efforts in setting industrial standards for any of the cladding or fuel options since these options are not currently in use by the industry. The same is true for the NRC which must license these new fuels since all current regulations are oriented toward UO 2 /Zr fuel. In addition, government supported research is needed to eliminate the scientific gaps in understanding of SiC composites and U 3 Si 2 fuel. Topics include the interaction of U 3 Si 2 and SiC, effect of SiC manufacturing methods on physical properties, ability to predict the mechanical properties of 7

8 SiC composites based on the composite design, swelling and fission gas release of U 3 Si 2. These issues are viewed as affecting the engineering of the fuel rather than being fundamental issues with the technical viability of these materials for fuel use based on the currently available data. The development and implementation programs for ATF are likely to cost >$300M for CZA and >$500M for SiC cladding, both with U 3 Si 2 fuel. Government investment in the research and development phases of the ATF program is required to reduce risk to a level appropriate for industry. The vast majority of the technical risk is at the research and development stage which will be <20% of the total cost. Government investment at the 80% level is appropriate at this stage due to the high technical risk involved VI. CONCLUSIONS Based on results to date under continued testing of the Westinghouse ATF products as described above, there appears to be no fundamental technical reason why the introduction of Westinghouse ATF products will not be achievable as LTRs and LTAs in the anticipated 2019 and 2022 timeframe, respectively. Performance issues with SiC and CZA claddings with U 3 Si 2 fuel have been identified and overcome through testing modifications and engineering. There may be other issues that surface in the future, as with any revolutionary new product. However, the robust research and development process that Westinghouse has in place will continue to be applied to address these issues. VII. ACKNOWLEDGMENTS This material is based upon work supported by the Department of Energy 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. VIII. REFERENCES 1. GORDON KOHSE, MIT, ED LAHODA, SUMIT RAY, FRANK BOYLAN, PENG XU and RICHARD JACKO, SiC Cladding Corrosion and Mitigation, Top Fuel 2016, Boise, ID, September 11, 2016, Paper Number 17450, ANS, (2016). 3. JASON HARP, Idaho National Laboratory preliminary photographs. 4. LU CAI, PENG XU, ANDREW ATWOOD, FRANK BOYLAN and EDWARD J. LAHODA, Thermal Analysis of ATF Fuel Materials at Westinghouse, ICACC 2017, Daytona Beach, FL, January 26, E. SOOBY WOOD, J.T. WHITE and A.T. NELSON, Oxidation behavior of U-Si compounds in air from 25 to S. MAŠKOVÁ, K. MILIYANCHUK and L. HAVELA, Hydrogen absorption in U 3 Si 2 and its impact on electronic properties, Journal of Nuclear Materials, 487, pp (2017). 7. EUGENE VAN HEERDEN, CHAN Y. PAIK, SUNG JIN LEE and MARTIN G. PLYS, Modeling Of Accident Tolerant Fuel for PWR and BWR Using Maap5, Proceedings of ICAPP 2017, Fukui and Kyoto,Japan, April 24-28, CHRISTIAN P. DECK, HESHAM E. KHALIFA, GEORGE M. JACOBSEN, JON SHEEDER, JIPING ZHANG, CARLOS BACALSKI, GOKUL VASUDEVAMURTHY, CHUNGHAO SHIH, SARAH OSWALD, KIRILL SHAPOVALOV,ERIC SONG, JOSHUA STONE, EDWARD LAHODA, RICHARD JACKO, and CHRISTINA BACK, Production of Accident Tolerant Fuel Irradiation Test Rodlets With SiC-SiC Cladding, Paper 110, 2017 Water Reactor Fuel Performance Meeting, Jeju Island, Korea, September 10-14, PENG XU, ED LAHODA, RICHARD JACKO, FRANK BOYLAN, AND ROBERT OELRICH, Status of Westinghouse SiC Composite Cladding Fuel Development, 2017 Water Reactor Fuel Performance Meeting, Jeju Island, Korea, September 10-14,