High-Temperature Behavior of CrNi-Steel DIN Material No. 1.4970 Cladding Material with Respect to Advanced Pressurized Water Reactor Safety Considerations

1988 ◽  
Vol 80 (1) ◽  
pp. 161-172 ◽  
Author(s):  
Claus Petersen ◽  
Gerhard Schanz ◽  
Siegfried Leistikow
Keyword(s):  
High Temperature ◽  
Temperature Behavior ◽  
Reactor Safety ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
High Temperature Behavior ◽  
Safety Considerations ◽  
Cladding Material

High Temperature Oxidation of the Hot Rolled ZrNbMoGe Alloy for Fuel Cladding Material of Pressurized Water Reactor

2015 ◽  
Vol 1123 ◽  
pp. 356-359 ◽  
Author(s):  
Bandriyana ◽  
A.K. Rivai ◽  
J.H. Prajitno ◽  
A. Dimyati
Keyword(s):  
High Temperature ◽  
Oxide Layer ◽  
High Temperature Oxidation ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Pressurized Water ◽  
Temperature Oxidation ◽  
Water Reactor ◽  
Hot Rolled ◽  
Cladding Material

The Zr-2.5Nb-0.5Mo-0.1Ge alloy were developed for application as fuel cladding material in an advanced Pressurized Water Reactor (PWR) with higher burn-up at higher service temperature. Oxidation behaviour of the alloy at high temperature similar to condition related to the operating with Lost of Cooling Accident (LOCA) conditions was investigated. The sample ingot was synthesized in an arc melting furnace followed by hot rolling at 850 °C down to 3 mm of thickness. Each one alloy ingot and one alloy sheet were subjected to the high temperature oxidation test in the Magnetic Suspension Balance (MSB) at 500 °C. Light optical microscope (LIOM) and Scanning Electron Microscope (SEM) equipped with X-ray Diffraction Spectroscopy (EDS) were used to characterize the microstructure of the oxide layer. The Vickers hardness tester was used to evaluate the hardness of the alloy matrix before and after oxidation processes. The results concluded that both samples showed oxidation rate characteristic which follows the parabolic phenomenon. However the hot rolled sample had lower rate. The oxide layer was indicated as ZrO2 .

European pressurized water reactor: safety objectives and principles

1999 ◽  
Vol 187 (1) ◽  
pp. 25-32 ◽  
Author(s):  
J. Czech ◽  
J. Wirkner ◽  
M. Yvon ◽  
M. Serret ◽  
U. Krugmann ◽  
...  
Keyword(s):  
Reactor Safety ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

Simulated Environmental Tests for Selected ATF Cladding Solutions

2021 ◽  
Author(s):  
Sami Penttila ◽  
Juha-Matti Autio ◽  
Jari Lydman ◽  
Aki Toivonen ◽  
Seppo Peltonen ◽  
...  
Keyword(s):  
High Temperature ◽  
Base Material ◽  
Operating Conditions ◽  
Advanced Technology ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
High Temperature Furnace ◽  
Normal Light ◽  
C 30

Abstract Current development on advanced technology fuel (ATF) claddings is aiming at improved high temperature integrity of new cladding solutions that are based on the existing zirconium claddings. To assess their performance for commercial use, their thorough characterization is essential. The primary requirement for the cladding materials is the ability to tolerate loss of cooling for a significant period without failing. The tests in this work were performed on different types of coated Zr-alloys in a high temperature furnace in flowing steam conditions at 1100 °C/ 60 min, 1200 °C/ 30 min and 1300 °C / 5 min. In addition, exposures were performed in pressurized water reactor (PWR) water chemistry to confirm the material viability in normal light water reactor (LWR) operating conditions. After PWR and steam tests, the exposed specimens were studied using a Zeiss Crossbeam 540 field emission gunscanning electron microscope (FEG-SEM) equipped with a semi-quantitative energy dispersive X-ray spectrometer (EDS). Most of the tested specimens indicated detached coating layer. Varying amounts of cracking in the coatings were present. Some of the cracks extended into the base material. Based on this study, further development is needed.

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