Sister Rod Destructive Examinations (FY21) Appendix J: Leaching of High Burnup Used Nuclear Fuel in Deionized Water - Spent Fuel and Waste Disposition

Sasikumar, Yadu; Nuttall, William; Keever, Tamara; Skitt, Darren; Montgomery, Rose and Bevard, Bruce (2022). Sister Rod Destructive Examinations (FY21) Appendix J: Leaching of High Burnup Used Nuclear Fuel in Deionized Water - Spent Fuel and Waste Disposition. In Sister Rod Destructive Examinations (FY21) US Department of Energy, Oak Ridge National Laboratory, Tennessee, USA.



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.

As a part of the DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zirc-4 (LT Zirc-4), ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rod’s condition before dry storage and are focused on understanding overall SNF rod strength and durability. Composite fuel and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition.

The leaching experiment was a part of a larger study conducted by Yadukrishnan Sasikumar in defending his doctoral thesis at The Open University, School of Engineering and Innovation. Sasikumar’s work utilized sister rod specimens after they had been fractured during fatigue testing and were slated for disposal.

The leaching experiment aims to understand the trends in the radiolysis-enhanced dissolution of HBU SNF when exposed to water (e.g., in-reactor or in-pool cladding failures). Specimens from a baseline M5-clad rod and a heat-treated M5 rod were cut from the fractured CIRFT specimens and placed in 100 mL deionized water for a period of 128 days. Both radial and axial sections were cut to provide different surface areas of fuel in contact with the leachate. During the four-month exposure period, aliquot samples of the leachate were analyzed using gamma spectroscopy and inductively coupled plasma - mass spectrometry (ICP-MS). The analysis quantified the amount of fuel leached into the solutions and provided individual isotopic release fractions (of 30+ isotopes) which were compared as a function of time and surface area of the fuel exposed.

Consistent with existing literature, the leaching followed a trend in which isotopes of certain elements such as Cs and Mo were among the first species in the matrix to dissolve, and with the highest release rates. This was followed by a gradual matrix dissolution consisting of uranium and other actinides and a slower-than-matrix release from some isotopes, including Ru and Rh. It was also observed that in this case, leaching did not increase with the increased surface area of fuel. Unlike to the trend observed in other leaching studies, the circumferential samples having less exposed fuel surface area leached more than the axial samples for a majority of the isotopes during the timespan of the study. Previous experiments were conducted by decladding or exposing the pellet-clad surface completely, but the samples used in this study retained the cladding. One possible explanation for the higher leaching rate of the circumferential samples is that the pellet-clad interface, which has a greater density of grain boundaries and defects, may be the most vulnerable area to leaching of fuel in the presence of water.

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