DAB Safety Team November 02, 2012
Media Contact: Don Leichtling
(619) 260-0160 or Ace Hoffman (760) 720-7261
FOR IMMEDIATE RELEASE
Fluid Elastic Instability
(FEI) is a phenomenon that can occur in poorly designed Steam Generators (SG’s)
due to very 'dry' steam (low moisture content, aka high
steam void fractions) causing the SG tubes to vibrate vigorously
along their length (called the in-plane
direction) until they hit
their neighboring tubes due to tight clearances. These
forces can cause tube-to-tube ruptures, while the tight clearances between the tubes can be attributed to
operating, poor design and or even manufacturing defects.
At the end of January
2012, a radioactive leak in SONGS RSG Unit 3, resulted in an emergency shut
down, the cause of which was later determined to have been fluid
elastic instability (FEI >1) caused by higher vapor fractions (~99.6 %). Later 8 tubes failed their “in-situ” pressure testing and leaked with a
flow > 0.5 gallons per minute at Main Steam Line Break Testing Conditions which
resulted in more than 800 additional tubes having to be plugged; which is
something that has never happened before in the USA. It is
important to note that SCE’s poorly designed RSG’s now have more damaged and or
plugged tubes than all the rest of the US
reactor fleet put together and that is with only 7% of the tubes in Unit 3
and 8% of the tubes in Unit 2 having been visually inspected to date!
Imagine what would have
happened if something like an “ordinary”
Main Steam Line Break (MSLB) occurred where the void fractions would have
reached 100% causing the vibration amplitude to increase exponentially, which
would then cause hundreds of tubes to leak and or rupture, which would have
then over-pressurized the steam generators, lifted the main steam safety valves
and released 60 tons of radioactive
coolant and steam into the Southern California environment within a matter of
minutes. This would have caused a Fukushima Type of Nuclear Reactor Meltdown
in SONGS Unit 3 Reactor, so Southern Californians were very lucky this time
(See all the DAB Safety Team Papers.).
The truth is that San Onofre escaped becoming an International Nuclear Events Scale (INES) Level 7 nuclear disaster by the
slightest of margins, unlike Fukushima!
The DAB Safety
Team assisted by several SONGS Anonymous Insiders has concluded that SONGS Unit
2 Replacement Steam Generators (RSG’s) are in worse shape now than certified by
the SCE and their three NEI Qualified, “U.S. Nuclear Plant Designers.” Even at 70% power operations, if a steam line
break outside containment were to occur in Unit 2, the depressurization of the
steam generators with the failure of a main steam isolation valve to close, it would
result in 100% void fraction in the degraded U-Tube bundle and the “straight leg portion”
between the Tube Support Plates. This
condition of ZERO Water in the steam generators would cause fluid elastic
instability (FEI) and flow-induced random vibrations, which would then result
in massive cascading SG tube failures, involving hundreds of degraded active SG
tubes, along with all the damaged inactive (all the plugged /stabilized) SG
tubes. With an undetermined amount of
simultaneous tube leaks/ruptures, approximately 60 tons of very hot
high-pressure radioactive reactor coolant would leak into the secondary
system. The release of this amount of
radioactive primary coolant, along with an additional approximately 200 tons of
steam in the first five minutes from a broken steam line would EXCEED the SONGS
NRC approved safety margins. So, in essence, the RSG’s will become loaded
guns, or a nuclear accident waiting to happen. Any failure under these conditions, would
allow significant amounts of radiation to escape to the atmosphere and a major
nuclear accident would easily result causing much wider radiological
consequences and even a potential nuclear meltdown of the reactor! Since these events would happen at an
extremely fast pace, no credit is assumed in the first 5 minutes of the main
steam line break accident for: (1) Enhanced Unit 2 Defense-In-Depth Actions -
SCE Restart Plan Enclosure 2, Item 9.0, and (2) The differential pressure
across the SG tubes necessary to cause a rupture will not occur if operators
prevent RCS re-pressurization in accordance with their Emergency Operating -
Enhanced Unit 2 Defense-In-Depth Actions - SCE Restart Plan Enclosure 2, Item
5.2,2, Probabilistic Risk analysis.
The above
statement is consistent with the conclusions and reports provided earlier on
this subject by:
1.
Fairewinds Associates
Internationally Known Nuclear Consultant Arnie Gundersen and his team of
Anonymous Industry insiders, who have had lengthy careers in steam generator
design, fabrication, and operation.
2.
Professor Daniel
Hirsch and Internationally Known Nuclear Consultant Dale Bridenbaugh.
3.
Dr. Joram Hopenfeld,
a retired engineer from the Office of Nuclear Regulatory Research and NRC's
Advisory Committee on Reactor Safeguards (ACRS) report issued in February 2001,
which substantiated many of Dr. Hopenfeld's concerns,
The
Operational Assessments reports prepared by AREVA, and Westinghouse “conflict
and contradict” * with MHI’s
Technical Report and Press Statements, on the causes and extent of degradation
pertaining to the SONGS Unit 2 Steam Generator Replacement Generators. The DAB Safety Team Expert Panel and SONGS Concerned Insiders opinion
is that these reports are not comprehensive and fail to arrive at a concise and
clear conclusion, because:
(1)
SCE Engineers have
either not provided, or they are withholding all the information to these
parties because of “The consequences of
being Wrong, Terminated or Fired”,
(2)
Due to competing and
proprietary interests between the three NEI qualified, “US Nuclear Plant
Designers”, these reports have not been openly and candidly discussed,
(3)
Time/Pressure exerted
by SCE on these parties to prepare Operational Assessments in order to rush to
Restart Unit 2 have led to incomplete conclusions,
(4)
Since nobody really
knows, what really happened, all the Parties have a shared interest to “Operate
Unit 2 at reduced power as a “Test Lab
to conduct Nuclear Experiments “ to determine, “What really went wrong with
unit 3, so SCE can determine the Root Cause, corrective actions, repair and
test plans to return both units 2 and 3 to full power operations.”
*NOTES: Just some examples
of the conflicting and contradicting statements are shown below:
1. Independent Expert 1
states, “U-tube out-of-plane direction is more susceptible to flow-induced
excitation than the in-plane direction due to lower U-bend natural frequency in
the out-of-plane direction. U-tube FEI in the in-plane direction has never been
observed in the U-tube SGs before its occurrence in the SONGS SGs. However,
recent academic studies report (2005) that FEI may also occur in the in-plane
direction, if tube motion in the in-plane direction is possible (no tube
in-plane supports or low tube contact forces with the out-of-plane supports). “
2. Independent Expert 2 states,
“Out-of-plane fluid-elastic instability has been observed in nuclear steam
generators in the past and has led to tube bursts at normal operating
conditions. However, the observation of in-plane fluid-elastic instability in
steam generators in a nuclear power plant is a true paradigm shift.”
DAB Safety
Team Comment to items 1 & 2: FEI in the in-plane direction has been
identified as early as 1983 by Academic Scholars and Palo Verde Replacement
Steam Generator manufactured in the early 2000s are designed for FEI. Weaver
and Schneider in 1983 examined the flow induced response of heat exchanger
U-tubes with flat bar supports. It is worth quoting the first conclusion of
their paper: “The effect of flat bar supports with small clearance is to act as
apparent nodal points for flow-induced tube response. They not only prevented
the out-of-plane mode as expected but also the in-plane modes. No in-plane
instabilities were observed, even when the flow velocity was increased to three
times that expected to cause instability in the apparently unsupported first
in-plane mode.”
3. Independent Expert 1 states,
“ECT-based AVB locations are compared with design-based locations. It is
evaluated that AVB insertion depth in actual SG is not changed compared with
the design-based location. There is some Pattern-1 wear identified by visual inspection,
for which Bobbin ECT was not able to detect as this type of wear.”
4. Independent Expert 2 states, "It
should be noted that because of field spread effects the bobbin probe typically
overestimates wear scar lengths." Even though no evidence of elongated
wear scars is evident in Unit 2, it doesn’t necessarily rule out undetected
in-plane instability. Wear scars at AVB locations may be too shallow to
evaluate properly and AVB wear scar lengths may be shortened by a contact
length that is small because of the presence of AVB twist. The best evidence of
in-plane instability is the detection of TTW, not the detection of elongated
AVB wear scars. Extensive inspections of the regions of interest with the +Pt™
probe show that possible undetected TTW would be less than 5 %TW. It is
unreasonable to expect detectable elongation of AVB wear scars without the
detection of TTW. The significance of elongated AVB wear scars is that the
amount of elongation reveals the extent of unstable tube motion in-plane.
5. Independent Expert 3 states that he does
not have access to the assembly procedures. The 0.12 to 0.14 dimensions are
anecdotal (based on personal observation, case study reports, or random investigations rather than systematic scientific
evaluation) without verification.
DAB Safety
Team Comments to items 3 & 4 & 5: Will be
provided later…
Copyright
November 02, 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.
AVB: Anti Vibration Bar
CPUC: California Public Utilities
Commission
DBA: Design Basis Accident
ECT: Eddy Current Testing
FEI: Fluid Elastic Instability
MHI: Mitsubishi Heavy Industry
MSLB: Main Steam
Line Break
NRC: Nuclear
Regulatory Commission
SCE:
Southern California Edison
TTW: Tube-to-Tube Wear
