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2025 ANS Winter Conference & Expo
November 9–12, 2025
Washington, DC|Washington Hilton
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Researchers use one-of-a-kind expertise and capabilities to test fuels of tomorrow
At the Idaho National Laboratory Hot Fuel Examination Facility, containment box operator Jake Maupin moves a manipulator arm into position around a pencil-thin nuclear fuel rod. He is preparing for a procedure that he and his colleagues have practiced repeatedly in anticipation of this moment in the hot cell.
Claus Petersen, Gerhard Schanz, Siegfried Leistikow
Nuclear Technology | Volume 80 | Number 1 | January 1988 | Pages 161-172
Technical Paper | Advanced Light Water Reactor / Fission Reactor | doi.org/10.13182/NT88-A35556
Articles are hosted by Taylor and Francis Online.
To investigate the potential of the austenitic 15Cr-15Ni steel DIN Material No. 1.4970 as fuel cladding material for an advanced pressurized water reactor (APWR), rod and tube samples were mechanically tested under inert and oxidizing conditions by uniaxial loading and internal pressure up to 1200°C, to receive recent information about its safety potential under emergency cooling conditions. Uniaxial strength values are not influenced by test atmosphere. The total strain is quite low up to 950°C and increases sharply above this temperature to a maximum of ∼80% at 1100°C. The uniaxial creep strength shows a transition to more pronounced temperature and time dependence at 800°C, which is due to recrystallization. Creep rupture strain, which remains around 20% below 950°C, rises above that temperature to a level of 80 to 90%. Steam oxidation slightly decreases burst creep strength, mainly due to metal consumption, and markedly decreases the circumferential strain, especially due to the pronounced tendency to localized deformation at cracks through the defective oxide scale. Even then the circumferential strain of steel tubes is not small enough to meet reactor safety considerations with respect to the emergency cooling of a densely packed APWR core.