Environmental Engineering Reference
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challenging issue over the previous 8 years for operating BWR plants
where ~2-year cycles are normal and Zircaloy-2 is the standard channel
material. The primary reason for this was the unaccounted channel dis-
tortion caused by differential hydrogen across the channel resulting from
shadow corrosion on the blade side (known as shadow corrosion-induced
bow). Zircaloy-2 is particularly susceptible to this distortion mechanism
because it has a high hydrogen pickup fraction (HPUF) that increases
with exposure.
Several strategies have been developed to combat bow. As an interme-
diate resolution to this issue Zircaloy-4 has been reintroduced because it
is effectively resistant to shadow corrosion-induced bow and has similar
irradiation growth and creep performance to Zircaloy-2. The one disadvan-
tage of Zircaloy-4 is that it has less corrosion resistance than Zircaloy-2.
However, based on the extensive experience with Zircaloy-4 channels both
in the United States and Japan (plus processing improvements have been
made specifi cally to enhance corrosion resistance), the corrosion perfor-
mance of Zircaloy-4 is claimed to be adequate for channel applications.
Other examples of global nuclear fuel (GNF) publications on channel bow
are described by Mahmood
et al
. (2007) and Cantonwine
et al
. ( 2009 ).
Other reasons for BWR fuel channel bowing are (Strasser
et al
., 2010a ):
Fast neutron fl ux gradients from a variety of causes including the fl ux
gradient at the core periphery (see Fig. 5.2).
Non-uniform metallurgical structure (e.g. texture difference between the
two opposing channel sides leading to difference in irradiation growth
rate) or composition.
Non-uniform wall thickness.
As-fabricated bow.
The bowing may result in diffi culties in inserting the control rods (a safety
issue) and/or in a decrease in thermal margins, the latter from two possible
causes. First a departure from nucleate boiling (DNB) value: if the fuel rod
surface heat fl ux becomes large enough, the water fi lm adjacent to the fuel
rod will convert into a steam fi lm with a much lower thermal conductivity
resulting in a rapid large increase in the fuel cladding temperature which, in
turn, will accelerate the oxidation and embrittlement of the fuel cladding.
The maximum heat fl ux at which the water is converted into a steam fi lm is
referred to as the DNB value. Second, a loss of coolant accident (LOCA)
could, for example be caused by a coolant pipe break in the primary circula-
tion system since larger water gaps between assemblies may exist in the core
than is accounted for in the core nuclear design. To ensure that the LOCA
licensing criteria are met, the fuel rod surface heat fl ux must be limited.
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