b'Nuclear Energy University Project (NEUP) AwardInvestigation of Degradation Mechanisms of Cr-coated Zirconium Alloy Cladding in Reactivity Initiated AccidentsPrincipal Investigator: WooHyun Jung, University of Wisconsin-Madison (UWM)Team Members/Collaborators: Hwasung Yeom (Pohang University of Science and Technology [PUST]), Kumar Sridharan (UWM), Brent Heuser (University of Illinois Urbana-Champaign [UIUC]), James Corson (Nuclear Regulatory Commission)T he Cr-coated Zr-alloy isProject Description:expected to provide addi- The objective of the research is to tional safety margins byinvestigate the thermal, mechanical, virtue of its excellent corrosion/ and irradiation responses of oxidation resistance at accidentCr-coated Zr-alloy claddings under conditions as a near-term accidentRIA conditions, in comparison to tolerant fuel (ATF) cladding design.uncoated Zr-alloy cladding. The Recently, there was a data gapobjective is achieved by a pulse-type reported for licensing activitiesnuclear heat deposition on the and performance criteria includingCr-coated cladding/UO 2fuel system degradation of the coated claddingfollowed by comprehensive PIE with above the Cr-Zr eutectic tempera- two different Cr coating methods: the ture, post-quench ductility, andCS and PVD processes. The proposed coating integrity during swelling/ TREAT experiment focuses on rupture. This research investigatesdemonstration of various cladding the thermal, mechanical, andfailure modes at the later phase tran-irradiation response of Cr-coatedsient (post-departure from nucleate cladding tubes under prototypicalboiling), including ballooning/burst, reactivity-initiated accident (RIA)severe high temperature oxidation, conditions, compared to uncoatedpartial melting due to the Cr-Zr cladding. Two different Cr-coatingeutectic, or a combination thereof. methods have been investigated,In-situ monitoring during the the cold spray (CS) process and thetesting includes water and cladding physical vapor deposition (PVD)temperature, fuel pellet temperature, process. The RIA tests have beencladding internal pressure, and performed at the Transient Reactorcladding elongation. Through this Test Facility (TREAT) at Idahowork, the hypothesis is tested that National Laboratory (INL). The testsdegradation or failure mechanism impose power excursions on theof the Cr-coated Zr-alloy cladding Cr-coated cladding/UO 2fuel systemis different from that of uncoated followed by comprehensive post- conventional Zr-alloy cladding in the irradiation examination (PIE). Thishigh temperature phases of RIA tran-work will extend our knowledgesients. For example, the CS process of the Cr-coated Zr-alloy claddingrelies on severe plastic deformation under the design-basis accidents. of powder particles and substrate surface, resulting in compressive stress in both the coating and the underlying substrate. These compressive stresses may influence 120 2023|AFC ACCOMPLISHMENTS'