nucreonics DATA SHEET no. 30 
Reactor Materials 
Properties of Zircaloy-2 
By L. S. RUBENSTEIN 
Bettis Plant, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania 
THE DATA COMPILATION adjoining sum- 
marizes the present knowledge of the 
corrosion, mechanical and_ physical 
properties of Zircaloy-2. 
Zircaloy-2 is an alloy of sponge zir- 
conium containing nominally 1.5 wt% 
Sn, 0.12 wt% Fe, 0.05 wt% Ni and 
0.10 wt% Cr. Its use in reactors is 
based on its excellent corrosion resist- 
ance, low thermal-neutron-absorption 
cross section and its structural stability 
at reactor operating temperatures. 
Particularly attractive for reactor- 
fuel cladding is the low thermal-neu- 
tron-absorption cross section of 0.22— 
0.24 barn of Zircaloy-2 (1a) [0.18 for 
pure zirconium, which has a scattering 
cross section of ~8 barns (10)]. 
Zircaloy-2 is vacuum-melted com- 
mercially by a double-consumable- 
electrode arc-melting technique. It 
undergoes allotropic transformation 
from the low-temperature alpha phase 
(close-packed-hexagonal) to the high- 
temperature beta phase (body-cen- 
tered-cubic). This transformation 
plays an important role in the variation 
of corrosion and mechanical properties. 
It is very easy to weld, but must be 
protected by welding in controlled 
inert-atmosphere boxes or by use of 
efficient inert-gas shields because of the 
rapid rate of contamination with 
oxygen and nitrogen when welded in 
air. Such addition or variation of im- 
purities can cause significant changes 
in mechanical properties and corrosion 
resistance. 
Zircaloy-2 has been fabricated by 
almost all of the known metalworking 
operations. Wire of all gages has 
been produced by drawing a rod formed 
by hot rolling or extrusion. Cups have 
been spun, deep-drawn, flow-turned, 
back-extruded and impact-extruded. 
Closed-die forgings have been made 
and tubing has been formed by both 
extrusion and welding. Shapes have 
also been cast using a special vacuum- 
melting and casting technique (2). 
Strip Specifications 
The important properties of Zirca- 
loy-2 can be best indicated by the 
characteristics of the nominal alloy as 
called for in the military specifications 
[Mil-Z-19859A (1959)] for acceptable 
strip as paraphrased below. 
Chemical composition. The chemi- 
cal composition shall be 1.20-1.70% 
tin, 0.07-0.20% iron, 0.05-0.15 chro- 
mium and 0.03-0.08% nickel. The 
TABLE 1—Specifications for 
Maximum Impurities 
Impurity Impurity 
level level 
Element (ppm) Element (ppm) 
Aluminum 75 Manganese 50 
Boron 0.5 Nitrogen 80 
Cadmium 0.5 Silicon 120 
Carbon 270 Sodium 20 
Cobalt 20 # Titanium 50 
Copper 50 Tungsten 100 
Hafnium 200 Uranium— 
Hydrogen 25 total 3.5 
Lead 130 Uranium- 
Magnesium 20 235 0.025 
sum of the iron, chromium and nickel 
contents determined from the average 
of all analyses made for a single ingot 
shall fall within the range 0.18—-0.38 %. 
Impurities. Unless otherwise speci- 
fied, impurity levels of the finished ma- 
terial shall not exceed the limits 
specified in Table 1. 
Mechanical properties. The aver- 
age hardness of material after anneal- 
ing shall not exceed the values speci- 
fied in Table 9 (see foldout adjoining) 
for the method used. 
The material shall conform to the 
elevated-temperature mechanical pro- 
perties specified in Table 11 (based on 
use of vacuum-melted material). 
Corrosion properties. Corrosion 
coupons, after completion of a 14-day 
test in 750° F, 1,500-psi steam, shall 
show the following: (a) a continuous 
black lustrous temper film; (b) freedom 
from white or brown corrosion products; 
(c) freedom from superficial surface 
stains, cracks, fissures, streaks and 
blisters; (d) a weight gain of 28 + 
10 mg/dm?. 
* * * 
This work was performed for the U. S. 
Atomic Energy Commission under contract 
AT-11-1-GEN-14. 
BIBLIOGRAPHY 
la. G. L. Hartfield, Bettis, personal communica- 
tion (1959). 1b. B. Lustman, F. Kerze, 
“The Metallurgy of Zirconium,’’ National 
Nuclear Energy Series Vol. VII-4 (McGraw 
Hill Book Co., New York, 1955) 
2. E. L. Richards, J. H. Hart, W. H. Friske, 
W. J. Hurford, ‘The Melting and Fabrication 
of Zircaloy,’’ 1958 Geneva Conference Paper 
No. 990 
3. R. E. Johnson, W. D. McMullen, in the 
Bettis Plant Materials Manual (1957) 
4. J. H. Keeler, 1956 educational course, ASM, 
Columbia River Basin 
6. J. Weinberg, in ‘‘Zirconium Highlights,” 
WAPD-ZH-12 (1958) 
6. J. G. Goodwin, The effect of heat treatment 
on the tensile and corrosion properties of 
Zircaloy-2, in ‘‘Zirconium  Highlights,"’ 
WAPD-ZH-5 (1958) 
7. S. Kass, WAPD-TM-97 (1957) 
8. W. L. Mudge, Jr., F. Forscher. Mechanical 
properties of Zircaloy-2, WAPD-101 (1954) 
9. R. G. Wheeler, W. 8. Kelly. Irradiation of 
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drogen, HW-39805 (1955) 
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behavior of Zircaloy-2 and Zircaloy-3, 
WAPD-TM-132 (1958) . 
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Zircaloy-2 and Zircaloy-3, in 
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mechanical properties of Zircaloy-2, WAPD- 
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communication (1958) 
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vacuum melted Zircaloy-2, in ‘Zirconium 
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Torsional properties of 
“Zirconium 
251 
