Press Release
The DAB Safety Team: December 7, 2012
SCE’s Claims About SONGS Unit 2 Steam
Generator Operating Pressures Are Erroneous Because They Conflict With SCE’s Submitted
NRC Reports And SCE’s Plant Procedures (Operational Data).
Now SCE is claiming in their Unit 2 Restart documents, “Limiting
power to 70% significantly reduces fluid velocity. The reduction in fluid
velocity significantly reduces the potential for FEI.” What they are not
saying is that reducing power to 70% significantly increases the steam generator
operating pressures, (as the NRC said in its AIT Report) which will:
· Increase
the pressure inside all the already damaged SG tubes
· Do nothing
to completely eliminate FEI from happening at any time during normal plant
operations, and especially during a MSLB or similar accident, which can cause a
nuclear incident or worse!
SCE’s attempt in using evasive and misleading technical
inconsistencies to justifying their dangerous and possibly catastrophic restart
plan cannot hide the truth, revealed in their actual plant operational data
provided to the NRC and published in the NRC AIT Report.
Background History:
After the radioactive leak occurred in the San Onofre Unit 3
steam generator, Arnie Gundersen along with a team of anonymous steam generator
experts were the first ones in the industry to absolutely state, “The pitch to
diameter ratio of tubes in the original CE generators is dramatically different
from any of the Westinghouse generators fabricated by Mitsubishi. As water moves vertically up in a steam
generator, the water content reduces as more steam is created. With the Mitsubishi design the top of the
U-tubes are almost dry in some regions. Without liquid in the mixture, there is
no damping against vibration, and therefore a severe fluid-elastic instability
developed. The real problem in the
replacement steam generators at San Onofre is that too much steam and too
little water is causing the tubes to vibrate violently in the U-bend region.
The tubes are quickly wearing themselves thin enough to completely fail
pressure tests. Even if the new tubes are actively not leaking or have not
ruptured, the tubes in the Mitsubishi fabrication are at risk of bursting in a
main steam line accident scenario and spewing radiation into the air.”
SCE’s Restart Plan Justification Is Just
Scientific Misinformation:
Based on analysis of the NRC AIT Report, Westinghouse’s
Operational Assessment, SONGS procedures, operational data, plant daily
briefing sheets and engineering calculations the DAB Safety Team concludes the
following:
·
Secondary
side lower pressures (833
psi) along with higher reactor thermal power and design deficiencies
(low tube clearances) at 100% power created conditions of “ALMOST NO WATER” in certain
regions of both Unit 3 steam generators tube bundles. This resulted in fluid elastic instability,
where unprecedented tube-tube wear was observed. At the June 18, 2012 AIT presentation, the
NRC said, “Throughout the US nuclear industry, this is the first time more than
one steam generator tube failed pressure testing…. Eight tubes failed. The
pressure testing identified that the strength of eight tubes was not adequate
and structural integrity might not be maintained during an accident… this is a serious
safety issue.” Southern Californians
were lucky, that SONGS Unit 3 tube leakage was detected and stopped in
time. Otherwise, this condition could
have potentially caused a reactor meltdown like Fukushima in Southern
Californian’s backyards.
·
Secondary
side higher pressures in Unit 2 (864-942 psi) at 100% power negated the effects of “low tube clearances” and prevented steam
“dry-out” (high void fractions) in the Unit 2 tube bundle region, where no
fluid elastic instability (tube-tube wear) was observed.
The DAB Safety Team’s findings are
summarized as follows:
·
DAB
Safety Team “Strongly Agrees” with Arnie Gundersen and his team of anonymous
steam generator experts and with MHI on the causes of fluid elastic instability
in Unit 3. What did SCE do, instead of
thanking Arnie Gundersen, who first identified the real cause of the problem, tried
to discredit him by implying, “What does he know about steam generators, he is
just a high school math teacher.”
·
DAB
Safety Team “Agrees” with Westinghouse, why fluid elastic instability did not
occur in Unit 2.
·
DAB
Safety Team “Strongly Disagrees” with both SCE’s conclusions “that fluid
elastic instability Most Likely Occurred in Unit 2” and “secondary side
operating parameters were similar in the U3 and
U2 SGs”.
·
DAB
Safety Team “Strongly Disagrees” with NRC that the differences in the actual
operation between units and/or individual steam generators had an insignificant
impact on the results and in fact, the NRC AIT team did not identify any changes
in steam velocities or void fractions that could account for the differences in
tube wear between the units or steam generators. Discussions with two of the NRC panel members
gives us the perception that the NRC panel members disagree amongst themselves
and also with SCE on the effect of operational parameters on fluid elastic
instability in Unit 2 Steam Generator E-089.
Adverse operational
conditions, such as larger reactor thermal power and lower steam generator
pressures (e.g., 833 psia)
and design deficiencies (low tube clearances and no-in-plane fluid elastic
instability structural protection) cause areas in the U-tube bundle of a
nuclear steam generator to have “ALMOST NO WATER” as observed in SONGS Unit 3
steam generators. When this happens,
fluid elastic instability occurs and the thin steam generator tubes carrying
radioactive coolant move with large sprinting amplitudes and hit the
neighboring tubes with violent and repeated impacts. Therefore, multiple tube
failures can occur, as was observed in SONGS Unit 3 at main steam line break
testing conditions.
MHI states, “The higher than typical void fraction
is a result of a very large and tightly packed tube bundle, particularly in the
U-bend, with high heat flux in the hot leg side. This high void fraction is a
potentially major cause of the tube FEI, and consequently unexpected tube-to-tube
wear (as it affects both the flow velocity and the damping factors). In
general, larger thermal power is more severe for vibration, because the steam
flow rate increases. At constant thermal power, lower steam pressure is more
severe for vibration than higher pressure.” MHI is indirectly saying that steam
generator pressures of 833
psia created fluid elastic instability in Unit 3, where unprecedented
tube-to-tube wear was observed. AREVA
states, “At 100% power, the thermal-hydraulic conditions in the U-bend region
of the SONGS replacement steam generators exceeded the past successful
operational envelope for U-bend nuclear steam generators based on presently
available data.” MHI has officially notified the NRC that all SONGS damaged RSG
Tubes subject to tube-to-tube wear (FEI) should be plugged and or stabilized. SCE cannot certify this as having been done,
since they have not inspected the majority of Unit 2’s RSG tubes using the most
advanced technology, as indicated in HMI’s official notice to the NRC. Again SCE is caught guessing about the amount
of tube fatigue damage, which directly affects the RSG tube structural
integrity; all RSG tubes are subject to tube-to-tube wear, extreme pressure
variations and other stresses during a MSLB or other unanticipated operational
transients.
NRC AIT Report states, “The team performed a number of
different thermal-hydraulic analysis of Units 2 and 3 steam generators. The
output of the various analyses runs were then compared and reviewed to
determine if those differences could have contributed to the significant change
in steam generator tube wear. It was noted that Unit 3 ran with slightly higher
primary temperatures, about 4°F higher than Unit 2. The result of the
independent NRC thermal-hydraulic analysis indicated that differences in the
actual operation between units and/or individual steam generators had an
insignificant impact on the results and in fact, the team did not identify any
changes in steam velocities or void fractions that could attribute to the
differences in tube wear between the units or steam generators. It should be
noted that increases in primary temperature and steam generator pressures has
the effect of reducing void fractions and peak steam velocities, which slightly
decreases the conditions necessary for fluid elastic instability and fluid-induced
vibration. The analysis
included the varying of steam generator pressures from 833 to 942 psia.”
SCE says in their Root Cause Analysis, “Secondary side operating parameters
were similar in the U3 and U2 SGs
and well within their design limits (e.g., steam generator pressures, 833 psia).” Note, NO mention varying
the pressure to 942 psia at all…
Copyright December 7, 2012 by The DAB Safety Team. All rights
reserved. This material may not be published, broadcast or redistributed
without crediting the DAB Safety Team. The contents cannot be altered without
the Written Permission of the DAB Safety Team Leader and/or the DAB Safety
Team’s Attorneys.
