Environmental Engineering Reference
In-Depth Information
Test load - 100 grams
Zipcaloy-2 (288°C)
250
326 °C
Zipcaloy-2
200
343°C
Low-oxygen sponge Zirconium
150
Crystal bar
Zirconium
100
343 °C
Irradiation
Temperature (°C)
326°C
~ 288 371-374
50
Zircaloy-2
Low-oxygen sponge Zr
Crystal bar Zr
0
0
1
2
3 4
Neutron fluence,
E
> 1 MeV (10
21
n/cm
2
)
5
6
7
8
4.26
Knoop microhardness vs fast neutron fl uence for zirconium and
Zircaloy-2 (Tucker & Adamson, 1984).
characteristics anneal out at temperatures above about 400°C, although the
rate of annealing is more sluggish for Nb-containing alloys.
4.4.3 Effects of hydrides on ductility
A brief summary of hydride effects is given here to provide background for
pellet-cladding mechanical interaction (PCMI) type failures. All zirconium
alloy reactor components absorb hydrogen during reactor service through the
corrosion reaction between zirconium and water. Basics of these phenomena
are given in ZIRAT Special Topical Reports (Cox & Rudling, 2000; Adamson
et al
., 2006 ; Strasser
et al
., 2008). Hydrides tend to embrittle zirconium alloys
and therefore their effects are important for in-reactor normal service, for
ex-reactor handling operations and for accident and transient scenarios such
as LOCA and RIA. It is thought that individual hydrides themselves are actu-
ally brittle at all normal reactor temperatures (Simpson & Cann, 1979; Shi &
Puls, 1999); and it is clear that high concentrations of hydrides (5000-16 000
ppm) are very brittle, as in hydride blisters or rims.
Under normal conditions, hydride platelets form in the circumferential
direction in fuel cladding illustrated in Fig. 4.27a, but under some circum-
stances such as during long term storage or during power transients they
Search WWH ::
Custom Search