Zirconiym—What is its Future? 
Attracted by zirconium’s low thermal-neutron cross section 
and good corrosion resistance, metallurgists have developed 
better alloys and improved fabrication techniques. 
Still 
needed—cheaper metal, more mechanical and corrosion data 
THE PLACE OF ZIRCONIUM in reactor 
engineering has been put on a much 
firmer basis in the last few years (NU, 
July 53, p. 27). In spite of its several 
difficult properties, metallurgists have 
better learned how far they can extend 
zirconium’s mechanical and corrosion 
limits by processing and alloying. 
Evidence of the interest in zirconium 
is the AEC’s rapidly expanding pro- 
curement program (NU, Dec. ’55,.p. 
13). [AEC has since requested bids, 
due Feb. 1, for an additional 300,000 
lb annually, bringing the annual AEC 
demand for reactor-grade zirconium 
to 1,200,000 lb.] It has been claimed 
that the Nautilus would not have been 
feasible without it. As an instance of 
its use in power reactors, the Shipping- 
port pressurized-water reactor uses 
U-Zr-alloy fuel elements clad with 
Zircaloy-2 (NU, Sept. ’55, p. 47). 
Industry’s confidence in the future 
of zirconium is indicated by the number 
of companies devoting facilities and 
money to investigating its behavior 
under operating conditions and to 
developing fabrication techniques. A 
number of companies were surveyed to 
provide the product and price tabula- 
7,4 Alloy Developments 
The primary virtue of zirconium and 
its alloys, other than their low neutron 
cross sections, is their good corrosion 
resistance in high-temperature 
(<680° F) water systems, e.g., water- 
cooled and -moderated reactors. Zir- 
conium-alloy systems are investigated 
with the aim to: 
1. Develop more reproducible and 
predictable corrosion behavior. 
2. Delay or eliminate breakaway 
(flaking or spalling white oxide film)— 
after breakaway, corrosion is faster. 
3. Find alloying elements that confer 
corrosion resistance. 
4. Utilize less expensive Kroll-proc- 
ess sponge. 
Of the binary alloying elements, Sn, 
54 
‘November 17. 
tion on pp. 6 and 7. Fabrication 
methods must still be perfected to allow 
mass production. It can be expected 
that prices will drop sharply in the near 
future—perhaps the biggest factor will 
be cheaper base metal through less 
expensive Zr-Hf separation methods. 
To assess the present status of zir- 
conium as a reactor structural mate- 
rial, a conference was sponsored by the 
Atomic Industrial Forum in New York, 
This article is based on 
the proceedings of that conference and 
on a Nuc.Leonics survey of commercial 
products available. 
TABLE 1—Tensile Properties of Zirconium Alloys 
0.2% offset Elongation 
yield strength (lb/in?) in 1 in. (%) 
Alloying element 
(Zr base) Room temp. 500° C Room temp. 500° C 
3% Al 55,000 28,000 30 23 
3% Sn 45,000 24,000 35 40 
3% Mo 73,000 30,000 ~0 40 
3% Nb 73,000 30,000 30 45 
3% Sn-1.5% Mo 59,000 30,400 25 42 
3% Sn-3.3% Al 78,900 39,400 5 6 
Yield strength (lb/in?) 
Zircaloy-2 
Zircaloy-3 
43,000 
37,000 
Tensile strength (lb/in*) 
33,000 
26,000 
19,000 
14,000 
64,000 
60,000 
