Low-Temperature Rupture Behavior of Zircaloy-Clad Pressurized Water Reactor Spent Fuel Rods Under Dry Storage Conditions-Reference-Cited by-同舟云学术

Low-Temperature Rupture Behavior of Zircaloy-Clad Pressurized Water Reactor Spent Fuel Rods Under Dry Storage Conditions

Published:1984-10 Issue:1 Volume:67 Page:107-123
ISSN:0029-5450
Container-title:Nuclear Technology
language:en
Short-container-title:Nuclear Technology

Author:

Einziger Robert E.¹,Kohli Rajiv²

Affiliation:

1. Westinghouse Hanford Company, P.O. Box 1970 Mail Stop W/A-40, Richland, Washington 99352

2. Battelle Columbus Laboratories, 505 King Avenue Columbus, O/i/o 45207

Publisher

Informa UK Limited

Subject

Condensed Matter Physics,Nuclear Energy and Engineering,Nuclear and High Energy Physics

Link

https://www.tandfonline.com/doi/pdf/10.13182/NT84-A33534

Reference35 articles.

1. L. D. BLACKBURN, D. G. FARWICK, S. R. FIELD, L. A. JAMES and R. A. MOEN, “Maximum Allowable Temperature for Storage of Spent Nuclear Reactor Fuel,” HEDL-TME 78-37, Hanford Engineering Development Laboratory, Richland, Washington (May 1978).

2. High Temperature Postirradiation Materials Performance of Spent Pressurized Water Reactor Fuel Rods under Dry Storage Conditions

3. R. S. KEMPER and D. L. ZIMMERMAN, “Neutron Irradiation Effects on the Tensile Properties of Zircaloy-2,” HW-52323, General Electric-Hanford Atomic Products Operation (Aug. 1957).

4. Rapid stress corrosion crack growth in irradiated zircaloy

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