The safe, reliable, and economic operation of the nation's nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry's success. Enhancing the accident tolerance of light water reactors (LWRs) became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident-tolerant fuel (ATF) for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-IIIϩ), fuels with enhanced accident tolerance would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance. Research and development of ATF in the United States is conducted under the U.S. Department of Energy (DOE) Fuel Cycle Research and Development Advanced Fuels Campaign. The DOE is sponsoring multiple teams to develop ATF concepts within multiple national laboratories, universities, and the nuclear industry. Concepts under investigation offer both evolutionary and revolutionary changes to the current nuclear fuel system. This paper summarizes the technical evaluation methodology proposed in the United States to aid in the optimization and prioritization of candidate ATF designs.

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This report documents work performed under the Spent Fuel and Waste Disposition's Spent Fuel and Waste Science and Technology program for the US Department of Energy (DOE) Office of Nuclear Energy (NE). This work was performed to fulfill Level 2 Milestone M2SF-22OR010201042, "FY2021 ORNL Report on High Burnup Sibling Pin Testing Results," within work package SF-22OR01020104 and is an update to the work reported in M2SF-21OR010201032, M2SF-19ORO010201026 and M2SF-19OR010201028. Sister Rod Destructive Examinations March 31, 2022 v Planned DE Status / Applicable Appendix Comments Five more tests with unpressurized specimens are expected to be completed in FY22. Tests using pressurized segments are planned to follow the unpressurized tests. DE.01 Measure internal pressure of five baseline and three heattreated rods Complete / Appendix CThe rod internal pressure and the void volume available inside the rod were measured for eight sister rods at room temperature (RT), and all pressures are within the publicly available database envelope. There is a clear correlation between the post-irradiated rod internal pressure and the as-designed fill pressure. The fission gas partial pressure trends well with the rod average burnup. The pressure and void volumes measured are consistent for rods from the same fuel vendor. The product of the partial pressure of the fission gas and the void volume, P f V, is consistent from lab to lab for sister rods from the same assembly, except for the two rods from assembly F35. A comparison of P f V indicates that the ZIRLO-clad rods might have experienced some change in pressure, void volume, or both due to the heat treatment applied, but the M5-clad rods do not exhibit the same effects.Comparisons with predictions from fuel rod performance codes FAST and BISON indicate a tendency for FAST to underpredict pressure and BISON to overpredict pressure.

This test plan describes the experimental work to be implemented by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) to characterize high burnup (HBU) spent nuclear fuel (SNF) in conjunction with the High Burnup Dry Storage Cask Research and Development Project [1] and serves to coordinate and integrate the multi-year experimental program to collect and develop data regarding the continued storage and eventual transport of HBU (i.e., >45 GWd/MTU) SNF. The work scope involves the development, performance, technical integration, and oversight of measurements and collection of relevant data, guided by analyses and demonstration of need.

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