Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation
Abstract
An out-of-reactor method for screening to predict relative in-reactor hydriding behavior of zirconium-bsed materials is disclosed. Samples of zirconium-based materials having different composition and/or fabrication are autoclaved in a relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solution at constant temperatures within the water reactor coolant temperature range (280.degree. to 316.degree. C.). Samples tested by this out-of-reactor procedure, when compared on the basis of the ratio of hydrogen weight gain to oxide weight gain, accurately predict the relative rate of hyriding for the same materials when subject to in-reactor (irradiated) corrision.
- Inventors:
-
- (Richland, WA)
- Kennewick, WA
- Richland, WA
- Issue Date:
- Research Org.:
- Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)
- OSTI Identifier:
- 867334
- Patent Number(s):
- 4916076
- Assignee:
- Battelle Memorial Institute (Richland, WA)
- Patent Classifications (CPCs):
-
G - PHYSICS G01 - MEASURING G01N - INVESTIGATING OR ANALYSING MATERIALS BY DETERMINING THEIR CHEMICAL OR PHYSICAL PROPERTIES
- DOE Contract Number:
- AC06-76RL01830
- Resource Type:
- Patent
- Country of Publication:
- United States
- Language:
- English
- Subject:
- method; predict; relative; hydriding; zirconium; alloys; nuclear; irradiation; out-of-reactor; screening; in-reactor; behavior; zirconium-bsed; materials; disclosed; samples; zirconium-based; composition; fabrication; autoclaved; relatively; concentrated; 0m; aqueous; lithium; hydroxide; solution; constant; temperatures; water; reactor; coolant; temperature; range; 280; degree; 316; tested; procedure; compared; basis; ratio; hydrogen; weight; oxide; accurately; rate; hyriding; subject; irradiated; corrision; lithium hydroxide; coolant temperature; constant temperature; water reactor; zirconium alloy; reactor coolant; temperature range; zirconium alloys; hydroxide solution; reactor method; accurately predict; oxide solution; predict relative; /436/73/376/
Citation Formats
Johnson, Jr., A. Burtron, Levy, Ira S, Trimble, Dennis J, Lanning, Donald D, and Gerber, Franna S. Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation. United States: N. p., 1990.
Web.
Johnson, Jr., A. Burtron, Levy, Ira S, Trimble, Dennis J, Lanning, Donald D, & Gerber, Franna S. Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation. United States.
Johnson, Jr., A. Burtron, Levy, Ira S, Trimble, Dennis J, Lanning, Donald D, and Gerber, Franna S. Mon .
"Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation". United States. https://www.osti.gov/servlets/purl/867334.
@article{osti_867334,
title = {Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation},
author = {Johnson, Jr., A. Burtron and Levy, Ira S and Trimble, Dennis J and Lanning, Donald D and Gerber, Franna S},
abstractNote = {An out-of-reactor method for screening to predict relative in-reactor hydriding behavior of zirconium-bsed materials is disclosed. Samples of zirconium-based materials having different composition and/or fabrication are autoclaved in a relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solution at constant temperatures within the water reactor coolant temperature range (280.degree. to 316.degree. C.). Samples tested by this out-of-reactor procedure, when compared on the basis of the ratio of hydrogen weight gain to oxide weight gain, accurately predict the relative rate of hyriding for the same materials when subject to in-reactor (irradiated) corrision.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1990},
month = {1}
}
Save to My Library
You must Sign In or Create an Account in order to save documents to your library.