Environmental Engineering Reference
In-Depth Information
curve technique is convenient to evaluate the irradiation embrittlement of
steels. The growth rate of a crack can be estimated from the known rela-
tion between the crack length and applied stress. As irradiation is known
to benefi t HCF and, as the material behaviour under HCF is well under-
stood, a prudent design for longer life becomes possible. Knowledge on the
creep rate of a material alerts for corrective measures as the dimensional
changes are predictable. The activation energy for creep indicates which
metallurgical parameter is crucial in limiting the life. Resolving the yield
stress into a source hardening and frictional terms helps understanding of
the fl ow response of the material to nuclear irradiation. It is now known
that synergistic effects of neutron irradiation and DSA could lead to bene-
fi cial effects on strength and ductility in certain temperature and strain-rate
regimes. By making a judicial choice of the temperature and fl uence, a steel
can be safely used in the blue brittleness range. Understanding the metal-
lurgical treatment and the material response has helped in choosing the
right material such as SA 304L instead of CW 316 for better creep resis-
tance for baffl e plates. In Zr-2.5%Nb alloy, the stable
β
phase (80%Nb) is
seen to be less creep resistant than the
phase (35% Nb) and the pressure
tubes (in Pressurized Heavy Water Reactors (PHWRs)) can have a longer
life with this modifi cation.
Corrosion is another major problem in nuclear reactors. Uniform, nod-
ular and shadow corrosion that affects the reactor components, and which
are not infl uenced by any external stress, are controlled by modifying alloy
and water chemistries. Routine surveillance test programmes enable better
understanding of material behaviour. This has helped to substitute some of
the components which suffer from SCC with those having better resistance
(e.g. Alloy 690, 52,152). IASCC is known to occur in almost all materials and
in components at low stress levels and this phenomenon is yet to be under-
stood well to come out with effective solution.
This chapter serves as an introduction to the various materials degrada-
tion phenomena as summarized above while the subsequent chapters dwell
on various details with Part I on various fundamental phenomena, Part II
on specifi c and varied components of LWRs while Part III covers manage-
ment strategies adopted by various nuclear utilities/vendors.
β
1.6
References
1.
'Light Water Reactor Sustainability Research and Development Program Plan,
Fiscal Year 2009-2013', Idaho National Laboratory Idaho Falls, Idaho 83415,
p. 7; http://www.inl.gov; Prepared for the U.S. Department of Energy Offi ce
of Nuclear Energy Under DOE Idaho Operations Offi ce Contract DE-AC07-
05ID14517.
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