Publications
Showing 29 results for Author: Yadu Sasikumar
Mar, 2026
Conference Paper
Hydride Reprecipitation in Prototypical 17×17 PWR Fuel Rods after Simulated Vacuum Drying Conditions
Light water reactor (LWR) fuel rod cladding picks up hydrogen as it oxidizes in reactor. Once the hydrogen concentration exceeds the solubility limit, zirconium hydride platelets precipitate, generally in the circumferential direction. In preparation for dry storage, elevated temperatures can dissolve some of the circumferential hydrides, which during cooling can then repr…
Mar, 2026
Conference Paper
Considerations for Thermal Annealing of Zr-Alloy Cladding During Dry Storage
In January 2025, the U.S. Department of Energy (DOE) sponsored a technical workshop with fuel vendors that provided a forum for increased collaboration on backend fuel cycle research. The workshop reviewed observations of thermal annealing of irradiated zirconium alloys in simulated dry storage conditions and discussed implications to the industry’s interest in reducing we…
Nov, 2025
ORNL Report
Prioritization Plan for High Burnup Research Cask Rods
This activity established a strategic framework to prioritize fuel rods from the High Burnup (HBU) Research Cask for testing and characterization. With over 8,400 rods available, selections were guided by data value, relevance to aging degradation mechanisms, and alignment with prior Sibling Pin baseline tests. Rods were categorized as high, medium, or low priority based o…
Aug, 2025
Journal
Experimental and computational characterization of a modified Sioutas Cascade impactor for respirable radioactive aerosols
Oak Ridge National Laboratory is collecting and characterizing aerosols released when spent nuclear fuel (SNF) rods are fractured in bending. An aerosol collection system was designed and tested to collect respirable sized (
Jan, 2025
ORNL Report
Sister Rod Destructive Examinations (FY23) Appendix I: SNF Aerosols Released During Rod Fracture
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 3 Milestone M3SF-23OR010201026, “FY23 M3 draft report on results from testing in FY23,” within work package SF-23OR01020102 and is…
Jan, 2025
ORNL Report
Sister Rod Destructive Examinations (FY23) Appendix B: Segmentation, Defueling, Metallographic Data and Total Cladding Hydrogen
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…
Aug, 2024
ORNL Report
Sister Rod Destructive Examinations (FY23) Appendix J: Leaching of High Burnup Used Nuclear Fuel in Deionized Water
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 section…
Aug, 2024
ORNL Report
Fatigue Testing and Characterization of Pre-hydrided Zircaloy-4 Cladding Tubes
The solubility of hydrogen in Zircaloy-4 (Zry-4) at the cladding operating temperature is near 100 wt. ppm. Above this solubility limit, excess hydrogen precipitates as -hydride platelets in the cladding material. Because of the effect of the thermal gradient across the cladding thickness on the migration and precipitation of hydrogen, hydride rims are often observed at t…
May, 2024
Journal
Assessing the release, transport, and retention of radioactive aerosols from hypothetical breaches in spent fuel storage canisters
Interim dry storage of spent nuclear fuel involves storing the fuel in welded stainless-steel canisters. Under certain conditions, the canisters could be subjected to environments that may promote stress corrosion cracking leading to a risk of breach and release of aerosol-sized particulate from the interior of the canister to the external environment through the crack. Re…