Published September 1990 | Version v1
Report

Improvements in zirconium alloy corrosion resistance

  • 1. Westinghouse Electric Corp., Pittsburgh, PA (USA). Commercial Nuclear Fuel Div.
  • 2. Westinghouse Electric Corp., Pittsburgh, PA (USA). Science and Technology Center

Description

The corrosion rates of a series of Zircaloy 4 and Zr-Nb alloys were evaluated in long-term (exceeding 500 days in some cases) autoclave tests. The testing was done at various conditions including 633 K (680 F) water, 633 K (650 F) water, 633 k (680 F) lithiated water (70 PPM/0.01 molal lithium), and 673 K (750 F) steam. Materials evaluated are from the following three groups: (1) standard Zircaloy 4; (2) Zircaloy 4 with tightened controls on chemistry limits and heat-treatment history; and (3) Zr-Nb alloys. To optimize the corrosion resistance of the Zircaloy 4 material, the effects of specific chemistry controls (tighter limits on nitrogen, oxygen, silicon, carbon and tin) were evaluated. Also the effects of the thermal history, as measured by integrated annealing of ''A'' time were determined. The ''A'' times ranged from 0.1x10-18 (h) to 46x10-18 (h). A material referred to as ''Improved Zircaloy 4'', having optimized chemistry and ''A'' time levels for reduced corrosion, has been developed and tested. This material has a reduced and more uniform corrosion rate compared to the prior Zircaloy 4 material. Alternative alloys were also evaluated for potential improvement in cladding corrosion resistance. ZIRLOTM material was chosen for development and has been included in the long-term corrosion testing. Demonstration fuel assemblies using ZIRLO cladding are now operating in a commercial reactor. The results for the various test conditions and compositions are reported and the relative corrosion characteristics summarized. Based on the BR-3 data, there is a ranking correspondence between in-reactor corrosion and autoclave testing in lithiated water. In particular, the ZIRLO material has significantly improved relative corrosion resistance in the lithiated water tests. Reduced Zircaloy-4 corrosion rates are also obtained from the tighter controls on the chemistry (specifically lower tin, nitrogen, and carbon; higher silicon; and reduced oxygen variability) and ''A'' times. 6 refs, 6 figs, 12 tabs

Part of:
Fundamental aspects of corrosion on zirconium base alloys in water reactor environments

Additional details

Publishing Information

Imprint Title
Fundamental aspects of corrosion on zirconium base alloys in water reactor environments
Imprint Pagination
275 p.
Journal Page Range
p. 145-157.
Report number
IWGFPT--34

Conference

Title
Technical committee meeting on fundamental aspects of corrosion on zirconium base alloys in water reactor environments.
Dates
11-15 Sep 1989.
Place
Portland, OR (USA).

INIS

Country of Publication
International Atomic Energy Agency (IAEA)
Country of Input or Organization
International Atomic Energy Agency (IAEA)
INIS RN
22036883
Subject category
S36: MATERIALS SCIENCE; S42: ENGINEERING;
Resource subtype / Literary indicator
Conference
Descriptors DEI
ALLOY-ZR98SN-4; AUTOCLAVES; CORROSION RESISTANCE; HEAT TREATMENTS; MATERIALS TESTING; NIOBIUM ALLOYS; ZIRCONIUM BASE ALLOYS
Descriptors DEC
ALLOYS; CHROMIUM ADDITIONS; CORROSION RESISTANT ALLOYS; HEAT RESISTING ALLOYS; IRON ADDITIONS; TESTING; TIN ALLOYS; ZIRCALOY; ZIRCONIUM ALLOYS

Optional Information