PLENTIFUL ENERGY

in 1987; Germany completed the construction of SNR-300 (762 MWth/327 MWe)
in 1985 but never operated it; and the U.S. cancelled the Clinch River Breeder
Reactor (1,000 MWth/380 MWe) while its NRC licensing process was nearing a
satisfactory completion.
Fast reactors listed in Table 14-7 have had a mixed record of operation. Largely
these were first-of-kind demonstration plants built in each country. Design mistakes
were made occasionally; there were component failures, particularly in the nonnuclear
portion of plants, a few sodium leaks and fires, and, in EBR-I and Fermi-1,
the partial core meltdowns we have described earlier. On the other hand, the
decades-long success of EBR-II, operating on a shoestring budget, and the lessons
taken from the mistakes made elsewhere, lend confidence that fast reactors, when
properly designed and operated, and with the experience accumulated by a more
mature industry, should be safe, reliable, and easy to operate and maintain.
Sodium difficulties included various leaks (mainly in the secondary systems and
in the steam generator systems), contamination of the coolant itself, and the effects
of aerosol deposits. Sodium leaks in the secondary system piping are easily detected
and the resulting smokes or fires are extinguished without undue difficulty or
consequence. As discussed in Chapter 7, the largest sodium fire was in the Japanese
demonstration reactor, Monju. There was a long delay in restart. It was not
technical in nature. An unrelated fire in a low-level waste plant at Tokai, an
unrelated criticality accident in a subcontractor‘s operation, a high court ruling
nullifying the original licensing base, opposition of local communities, and so on,
compounded over the years to delay restart for a prolonged interval. Monju finally
restarted its operation in May 2010.
A few reactors were plagued with leaks in steam generators during the early
years of their operation. Most leaks were at the tube-to-tube sheet welds, a
significant fraction due to manufacturing (welding) defects. Some were due to
material defects and lack of post-weld heat treatment. Manufacturing techniques are
as important as material selection to reliable components. Some of the component
failures were due to the selection of advanced materials that had little testing in
sodium at elevated temperature. Early steam generators of stainless steel
experienced stress corrosion cracking. To date, 2-1/4Cr-1Mo has been the dominant
choice for steam generators and has had a very satisfactory operating experience.
The 9Cr-1Mo is thought to be an improvement over 2-1/4Cr-1Mo for future
applications and has received the ASME code qualification. For the primary
system, both stainless steel types 316 and 304 have been successfully used. For fuel
assemblies, ferritic HT-9 has proven to be the most reliable, low-swelling material
for cladding and hexagonal ducts.
EBR-II stands out in all of this as a well-designed reactor which operated very
successfully for a long time, and could have gone on operating with no identifiable
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