atw 2018-09v3

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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 8/9 ı August/September FUEL 444 uncoated zirconium experiences at 1,200 °C. At temperatures of 1,300 °C, however, the Chromium- 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 chromium- coated zirconium 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 1,600 °C up to 1,700 °C. These tests were terminated only because of excessive corrosion of the heater element. At 1,600 °C, the SiC cladding was visually untouched. At 1,700 °C, there were indications of small beads on the surface, presumably SiO 2 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 Helium cover gas and to allow accurate weight changes to be made on the SiC rodlets so that kinetic data can be obtained. 3.3 Testing of Westinghouse U 3 Si 2 ATF high-density fuel U 3 Si 2 is a revolutionary material for LWR fuel service because its inherent thermal conductivity is much greater than existing UO 2 -based fuel, resulting in significantly lower pellet temperatures. U 3 Si 2 -based fuel can also have up to 17 percent greater uranium density than UO 2 -based fuel, so considerably more energy can be economically realized from each individual fuel assembly. However, due to these differences, considerable data is required on the behavior of U 3 Si 2 at LWR operating temperatures (estimated to be from 600 °C and up to 1,200 °C during transients). To obtain the necessary data, U3Si2 fuel pellets were manufactured at INL and put into rodlets in the ATR in 2015. The first rodlets came out of the ATR at the end of 2016 (Figure 3) and post-irradiation examination (PIE) was performed in the summer of 2017 at INL [3]. The PIE results indicate some small amount of cracking that may have been due to impurities within the U 3 Si 2 . Fission gas release and swelling were both essentially zero with an exit burnup of 20 MWd/kgU. Considering the ATR high heat generation rates (12 to 15 kW/ft), which are significantly above the average of 5 kW/ft and peak of 9 kW/ft normally found in LWRs, this was exceptionally good behavior. The next set of U 3 Si 2 pins is due out in <strong>2018</strong> and will have achieved a burnup of 40 MWd/kgU. U 3 Si 2 was tested for air and steam oxidation and compared to UO 2 using digital scanning calorimeters at both the Westinghouse Fuel Fabrication Facility in Columbia, South Carolina (USA) [4] and at Los Alamos National Laboratory (LANL) [5]. The Westinghouse test results indicate that the ignition temperatures for UO 2 and U 3 Si 2 are between 400 °C and 450 °C. The LANL results indicate 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 U 3 Si 2 is considerably higher than for UO 2 . The effect of this difference in heat release and mass on the stability of the rods was investigated in rodlet tests in the autoclaves in the Churchill facility during the summer of 2017. Unacceptable tube bulging was found and programs are now underway to increase the oxidation resistance of the U 3 Si 2 . | | Fig. 3. Neutron radiographs of 20 MWd/kgU U3Si2 pins from ATR. Note the lack of pellet cracking and distortion.(Ref. 4). 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 ppm levels. 4 Accident scenario evaluations To assess and demonstrate the performance of ATF materials in postulated accident scenarios, Modular Accident Analysis Program, Version 5 (MAAP5), calculations were performed for chromium-coated zirconium and SiC claddings along with high-density fuels for the station blackout scenario and the Three Mile Island Unit 2 (TMI2) small-break loss-of-coolant (LOCA) scenario with replenishment of the primary coolant [6]. The chromium-coated zirconium 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 would not prevent the contamination of the PWR primary loop due to ballooning and bursting at about 800 °C to 900 °C, the chromiumcoated zirconium option could prevent a TMI-2 type of accident from extending into the fuel meltdown phase and prevent extensive contamination of the containment and perhaps preserve the nuclear plant. This is because, although the Cr-coated Zr may begin to fail as the temperature exceeds 1,400 °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 (Figure 4). Note that, in this case, Iron-chromium-aluminum (FeCrAl) was used to model the performance of chromium-coated zirconium since the temperature and oxidation performance is about the same. The results for the station blackout scenario (Figure 5) 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 (~2,000 °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 (Figure 6.) 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 Fuel Westinghouse EnCore® Accident Tolerant Fuel ı Gilda Bocock, Robert Oelrich, and Sumit Ray

<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 8/9 ı August/September FUEL 445 | | Fig. 4. Hottest core node for TMI-2 accident where coolant is restored at ~9,900 seconds. | | Fig. 5. Hottest core node for PWR station blackout. ­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 TMI-2 type accident (Figure 4). 5 Transition cycle analysis for optimum ATF implementation in current PWRs 5.1 U 3 Si 2 fuel As previously noted, one of the ­primary benefits of U 3 Si 2 is that it increases the uranium density by up to 17 percent as compared to UO 2 . This yields an effective enrichment of 0.84 weight percent U-235 as compared to 0.71 weight percent U-235 found in natural uranium. This increase in density will support improved fuel cycle economics and reduce the total number of fuel bundles that need to be inserted into a reactor, resulting in significant savings. Because of the increased density, the use of U 3 Si 2 also extends the energy output and cycle length capability for PWR fuel assemblies, while remaining below the 5 weight percent enrichment limit for commercial fuel. The Westinghouse ATF can thus either decrease the fuel cycle cost of 18-month cycles by reducing the number of feed assemblies and increasing fuel utilization, or it can make 24-month cycles economical for today’s uprated, high-power density PWRs. Economic analysis shows that the Westinghouse EnCore Fuel has very favorable economics, not only at the ATF equilibrium cycle, but also during the transition cycles from UO 2 to ATF. This is especially applicable when transitioning to a 24-month cycle operational regime, which thus represents the recommended path forward for implementation. The higher thermal conductivity of the U 3 Si 2 also provides a very high tolerance for transients while operating at higher linear heat generation rates than is possible for UO 2 – which will increase plant operating margin. In addition, the higher uranium density can extend the core operating capability compared to current fuels, while maintaining the current 5 weight percent 235U enrichment limit for commercial fuel; yet enable economically competitive fuel management schemes for the longer cycles. In particular, the introduction of ATF in a current 18-month cycle high-power density PWR to accomplish a transition from UO 2 to ATF by either maintaining the currently predominant 18-month cycle operational regime, or extending it to a 24-month cycle has been analyzed. Implementing the Westinghouse ATF to achieve a more cost effective 18-month cycle will deliver fuel cost savings due to fewer fresh assemblies per reload and improved fuel utilization. Implementing the Westinghouse ATF in conjunction with a transition to 24-month cycle will yield economic benefits due to the resulting reduced number of outages and related savings, which offset the slightly higher fuel costs (as compared to 18-month cycle fuel costs). Analyses have shown that the economic impact of the transition cycles to implement a 24-month cycle operation with ATF is significantly better than the economic impact of transition cycles which implement ATF and maintain an 18-month cycle operation. It is anticipated that the fabrication costs to make the U 3 Si 2 powder could increase as compared to existing UO 2 fabrication. However, after the powder is made, only minor cost increases are expected to occur in the rest of the fuel manufacturing process. Therefore, the overall cost increase is anticipated to be offset by the safety, economic and operational benefits. | | Fig. 6. Total hydrogen generated for PWR station blackout. 5.2 Chromium-coated zirconium cladding Chromium-coated zirconium alloy offers a higher accident temperature capability, compared to uncoated zirconium alloy cladding, of between 1,300 ˚C and 1,400 ˚C. The coated cladding also reduces corrosion and hydrogen pickup. Resistance to rod wear is another benefit of this cladding type. The potential for exothermic reactions is greatly reduced during LOCA or transient events that lead to high-temperature fuel transients. These attributes provide both safety and economic benefits that support licenseability and economically viable transition scenarios. 5.3 SiC cladding SiC cladding provides 25 percent lower thermal neutron cross-sections than current Zr cladding. This would afford even greater neutron economy. Additionally, the fuel and cladding would be able to withstand temperatures ~2,000 ˚C in the event of a beyond design basis accident. This temperature increase could result in a rise in design basis operating margins. 6 Licensing To get EnCore Fuel licensed and loaded into commercial reactor cores in region quantities by 2027, Westinghouse has initiated a program to ­significantly compress the licensing timeframe from initial testing to Fuel Westinghouse EnCore® Accident Tolerant Fuel ı Gilda Bocock, Robert Oelrich, and Sumit Ray