Media Alert 13-01-28
Media
Contact: Don Leichtling (619) 296-9928 or Ace Hoffman (760)
720-7261
SONGS
UNIT 3 RSG REAL ROOT CAUSE: Lack of “Critical Questioning & Investigative
Attitude” by SCE, MHI and NRC Region IV and AIT Team.
The DAB Safety Team has transmitted the following Allegation to “Office of the Chairman of the NRC”
and “The United States Senate Committee on Environment and Public Work.”
It is the DAB Safety Team’s goal to help educate
both the NRC and the Public by providing unbiased, logical and factual
information in order to help assess the real dangers of any San Onofre Unit 2
restart. According to Press Reports and
San Onofre Insiders, Unit 2 permission for restart by the NRC is imminent yet
the REAL
Root Cause for the $1 Billion destruction of Units 2 and 3 RSGs (Including
equipment cost and expenses) has not yet even been determined. The Public
does not know the status of SCE’s ongoing cause evaluations, SCE’s response to
32 NRR’s RAI’s and NRC’s Special San Onofre Inspections. We like to remind NRC
San Onofre Special Panel, what NRC Chairman Macfarlane said during her recent
Fukushima Trip, “Regulators may need to be ‘buffered’ from political winds, but
they need to be fully subjected to the pressure of scientific and engineering
truth and cannot be allowed to make decisions or order actions that are
‘independent’ of facts.” The NRC rush to a faulty judgment
cannot be allowed to compromise Public Safety just to please SCE, as this
conflicts with President Obama's Policy, the new NRC Chairman’s Standards and
the advice of NRC retired Branch Chiefs who have also spoken out.
A NRC Branch Chief gifted with MIT
Intelligence, Intuition and a Sixth Sense, told an anonymous participant at an
Industry Conference, “Sir, to resolve any complex technical problem and
understand unclear regulations, you have to, ‘Read and reread in between the
lines’, use, ‘Critical questioning and an investigative attitude’ and ‘Solid
teamwork & alignment.”
SONGS UNIT 3 RSG ROOT CAUSE: It
appears that Complacent SCE and Inexperienced MHI Engineers did not perform
proper academic research and industry bench marking about the potential adverse
consequences of the reduction of original CE steam generator pressures from 900
psi to say, 800 psi on fluid elastic instability and flow-induced
vibrations. These lower secondary steam operating pressure (800-833 psia)
are the primary cause for shortening the life of SONGS Original Combustion
Engineering Generators due to increased tube wear and plugging caused by flow-induced
random vibrations and destruction of SONGS Unit 3 Replacement Steam Generators
due to flow-induced random vibrations, Mitsubishi flowering effects and steam
voids or steam dry-outs (AKA fluid elastic instability). In addition, SCE
Engineers prepared defective 10 CFR 50.59 Evaluation and design specifications,
which were not challenged by MHI, and adequately reviewed by NRC Region IV. MHI
at the direction of SCE Engineers made numerous untested and unanalyzed design
changes to the steam generators under the pretense of “like for like”, and Elmo
Collins said, “The guts of the machinery look …. Different.”
Therefore
the SONGS UNIT 3 RSG REAL ROOT CAUSE: Lack of “Critical Questioning &
Investigative Attitude” by SCE, MHI and NRC Region IV and AIT Team.
·
Lessons Learnt: MHI, AIT, DAB Safety Team together with other
World Nuclear and SG Experts / Manufacturers agree that, “Lower secondary steam
operating pressures (800-833 psi) are severe and can easily cause SG
flow-induced random vibrations and fluid elastic instability.” At
lower secondary steam operating pressures (800-833 psi), the utility can
generate more thermal megawatts out of the SG and hence add more power to the
grid and thereby make more money for SCE/EIX Officers and shareholders. These
lower secondary steam operating pressures (800-833 psi) were the primary cause
for shortening the life of SONGS Original Combustion Engineering Steam
Generators. The real lesson learnt is
that ALL parties must follow the NRC Chairman’s advice, pay attention to their
work and always ensure that public safety is THE overriding obligation for the regulators, licensee, its vendors
and contractors. NRC “Reasonable
Assurance” for the protection of adequate health and safety of the public from
postulated radiological accidents cannot be compromised at any time during the
design, operation, fabrication, testing, surveillance, maintenance and
inspection of a nuclear power plant. SCE and its offsite response organizations need to
demonstrate; (1) Feasibility of an Operator Action during a postulated main
steam line break with multiple Unit 2 steam generator tube ruptures, (2) They
can effectively implement emergency plans and procedures with zero
Drills/Exercise Performance indicator failures during a Fully Staffed NRC/FEMA
evaluated exercise prior to any Unit 2 Restart.
·
Comments about the NRC Augmented Inspection Team San Onofre
Report
NOTE:
We highly recommend that NRC Augmented Inspection Team and NRC San Onofre
Special Review Panel thoroughly review SONGS Unit 2 Return to Service MHI,
AREVA, Westinghouse, DAB Safety Team and John Large Reports, then carefully
examine the operational differences between Unit 2 and 3 and then update the
NRC AIT report with a FACTUAL Root cause for FEI in Unit 3 and NO FEI in Unit
2. NRC San Onofre Special Review Panel also
needs to review the SONGS Unit 2 Restart Reports (done by SCE, Westinghouse,
AREVA and MHI), SCE Unit 3 Root Cause Evaluation, NRC AIT Report, ATHOS
Modeling Results and Unit 2 Operational Data and then arrive: (1) At a
unanimous, clear and concise conclusion whether FEI occurred in Unit 2 or not,
and (2) Provide a GAP ANALYSIS (The scientific, technical and engineering
reasons why all these reports are so different) prior the February 12, 2013 NRC
Public Meeting.
The AIT inspection concluded that:
(1) SCE was adequately pursuing the causes of the unexpected steam generator
tube-to-tube degradation. In an effort to identify the causes, SCE retained a
significant number of outside industry experts, consultants, and steam
generator manufacturers, including Westinghouse and AREVA to perform
thermal-hydraulic and flow induced vibration modeling and analysis; (2) The
combination of unpredicted, adverse thermal hydraulic conditions and
insufficient contact forces in the upper tube bundle caused a phenomenon called
“fluid-elastic instability” which was a significant contributor to the tube to
tube wear resulting in the tube leak. The team concluded that the differences
in severity of the tube-to-tube wear between Unit 2 and Unit 3 may be related
to changes to the manufacturing/fabrication of the tubes and other components
which may have resulted in increased clearances between the anti-vibration bars
and the tubes; (3) Due to modeling errors, the replacement generators were not
designed with adequate thermal hydraulic margin to preclude the onset of
fluid-elastic instability. Unless changes are made to the operation or
configuration of the steam generators, high fluid velocities and high void
fractions in localized regions in the u-bend will continue to cause excessive
and accelerated tube wear that could result in tube leakage and/or tube
rupture; (4) The thermal hydraulic phenomena contributing to the fluid-elastic
instability is present in both Unit 2 and 3 steam generators; (5) Based on the
updated final safety analysis report description of the original steam
generators, the steam generators’ major design changes were appropriately
reviewed in accordance with the 10 CFR 50.59 requirements.
So
based on a review of the AIT Report and World’s Experts, the three potential
causes, which were significant contributors to the “fluid-elastic instability”
in SONGS Unit 3 and the tube-to-tube wear resulting in the tube leak are as
follows:
A. Insufficient contact tube-to-AVB
forces and differences in manufacturing or fabrication of the tubes and other
components between Units 2 & 3.
B. Due to modeling errors, the SONGS
replacement generators were not designed with adequate thermal hydraulic margin
to preclude the onset of fluid-elastic instability.
C. Differences between Unit 2 and
Unit 3’s Operational Factors.
A.
Let us now examine that whether insufficient contact tube-to-AVB forces in the
Unit 3 upper tube bundle caused “fluid-elastic instability” which was a
significant contributor to the tube-to-tube wear resulting in the tube leak.
A.1-
MHI states, “By design, U-bend support in the in-plane direction was not
provided for the SONGS SG’s”. In the design stage, MHI considered that the tube
U-bend support in the out-of-plane direction designed for “zero” tube-to-AVB
gap in hot condition was sufficient to prevent the tube from becoming
fluid-elastic unstable during operation based on the MHI experiences and
contemporary practice. MHI postulated that a “zero” gap in the hot condition
does not necessarily ensure that the support is active and that contact force
between the tube and the AVB is required for the support to be considered
active. The most likely cause of the observed tube-to-tube wear is multiple
consecutive AVB supports becoming inactive during operation. This is attributed
to redistribution of the tube-to-AVB-gaps under the fluid hydrodynamic pressure
exerted on the tubes during operation. This phenomenon is called by MHI, “tube
bundle flowering” and is postulated to result in a spreading of the tube
U-bends in the out-of-plane direction to varying degrees based on their
location in the tube bundle (the hydrodynamic pressure varies within the U
bend). This tube U-bend spreading causes an increase of the tube-to-AVB gap
sizes and decrease of tube-to-AVB contact forces rendering the AVB supports
inactive and potentially significantly contributing to tube FEI. Observations
Common to BOTH Unit-2 and Unit-3: The AVBs, end caps, and retainer bars were
manufactured according to the design. It was confirmed that there were no
significant gaps between the AVBs and tubes, which might have contributed to
excessive tube vibration because the AVBs appear to be virtually in contact
with tubes. 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. Because this high void fraction is a
potentially major cause of the tube FEI, and consequently unexpected tube wear
(as it affects both the flow velocity and the damping factors) the
correlation between the void fraction (steam quality) and the number of
tubes
with wear in a given void fraction region was investigated. From this
investigation, a
strong
correlation between the void fraction (steam quality) and the percentage of
tubes with
the
Type 1 (tube-to-tube) and Type 2 (tube-to-AVB) wear was identified. The
correlation between flow velocity and the number of tubes with wear was also
investigated. The results show that when the flow velocity is high, the
percentage of tubes with wear increases, even though this correlation is not as
strong as that between the void fraction (steam quality) and the percentage of
tubes with wear.”
A.2 – 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. The primary source
of tube-to-AVB contact forces is the restraint provided by the retaining bars
and bridges, reacting against the component dimensional dispersion of the tubes
and AVBs. Contact forces are available for both cold and hot conditions.
Contact forces significantly increase at normal operating temperature and
pressure due to diametric expansion of the tubes and thermal growth of the
AVBs. After fluid elastic instability develops, the amplitude of in-plane
motion continuously increases and the forces needed to prevent in-plane motion
at any given AVB location become relatively large. Hence shortly after
instability occurs, U-bends begin to swing in Mode 1 and overcome hindrance at
any AVB location.”
A.3 – Westinghouse states, “Test
data shows that the onset of in-plane (IP) vibration requires much higher
velocities than the onset of out-of-plane (OP) fluid-elastic excitation. Hence,
a tube that may vibrate in-plane (IP) would definitely be unstable OP. A small
AVB gap that would be considered active in the OP mode would also be active in
the IP mode because the small gap will prevent significant in-plane motion due
to lack of clearance (gap) for the combined OP and IP motions. Thus, a contact
force is not required to prevent significant IP motion. Manufacturing
Considerations: There are several potential manufacturing considerations
associated with review of the design drawings based on Westinghouse experience.
The first two are related to increased proximity potential that is likely
associated with the ECT evidence for proximity. Two others are associated with
the AVB configuration and the additional orthogonal support structure that can
interact with the first two during manufacturing. Another relates to AVB
fabrication tolerances. These potential issues include: (1) The smaller nominal
in-plane spacing between large radius U-bend tubes than comparable Westinghouse
experience, (2) The much larger relative shrinkage of two sides (cold leg and
hot leg) of each tube that can occur within the tubesheet drilling tolerances.
Differences in axial shrinkage of tube legs can change the shape of the U-bends
and reduce in-plane clearances between tubes from what was installed prior to
hydraulic expansion, (3) The potential for the ends of the lateral sets of AVBs
(designated as side narrow and side wide on the Design Anti-Vibration Bar
Assembly Drawing that are attached to the AVB support structure on the sides of
the tube bundle to become displaced from their intended positions during lower
shell assembly rotation, (4) The potential for the 13 orthogonal bridge
structure segments that are welded to the ends of AVB end cap extensions to
produce reactions inside the bundle due to weld shrinkage and added weight
during bundle rotation, and (5) Control of AVB fabrication tolerances
sufficient to avoid undesirable interactions within the bundle. If AVBs are not
flat with no twist in the unrestrained state they can tend to spread tube
columns and introduce unexpected gaps greater than nominal inside the bundle
away from the fixed weld spacing. The weight of the additional support
structure after installation could accentuate any of the above potential
issues. There is insufficient evidence to conclude that any of the listed
potential issues are directly responsible for the unexpected tube wear, but
these issues could all lead to unexpected tube/AVB fit-up conditions that would
support the amplitude limited fluid-elastic vibration mechanism. None were
extensively treated in the SCE root cause evaluation.”
A.4 - John Large,
Internationally Known Scientist and Chartered Nuclear Engineer from London says
about the SONGS Unit 2 Replacement Steam Generators (RSGs) AVB Structure, “It
impossible to reliably predict the effectiveness of the many thousands of AVB
contact points for when the tube bundle is in a hot, pressurized operational
state. The combination of the omission of the in-plane AVB restraints, the
unique in-plane activity levels of the SONGS RSGs, together the very demanding
interpretation of the remote probe data from the cold and depressurized tube
inspection, render forecasting the wear of the tubes and many thousands of
restraint components when in hot and pressurized service very challenging
indeed. Phasing of AVB-TSP Wear
-v- TTW: I reason that, overall, the tube wear process comprises two distinct
phases: First, the AVB (and TSP) -to-tube contact points wear with the result
that whatever level of effectiveness is in play declines. Then, with the U-bend
free-span sections increased by loss of intermediate AVB restraint(s), the
individual tubes in the U-bend region are rendered very susceptible to FEI
induced motion and TTW. Whereas the OAs commissioned by SCE broadly agree that
the wear mechanics comprises two phases, there are strong differences over the
cause of the first phase comprising in-plane AVB wear: AREVA claim this is
caused by in-plane FEI whereas, the contrary, Mitsubishi (and Westinghouse)
favor random perturbations in the fluid flow regime to be the tube motion
excitation cause. Put simply: i) if AREVA is correct then reducing the reactor
power to 70% will eliminate FEI, AVB effectiveness will cease to decline
further and TTW will be arrested; however, to the contrary, ii) if Mitsubishi
is right then, even at the 70% power level, the AVB restraint effectiveness
will continue to decline thereby freeing up longer free-span tube sections that
are more susceptible to TTW; or that iii) the assertion of neither party is
wholly or partly correct. I find that the AVB assembly, which features strongly in
the onset of TTW, is clearly designed to cope only with out-of-plane tube
motion since there is little designed-in resistance to movement in the in-plane
direction - because of this, it is just chance (a virtually random combination
of manufacturing variations, expansion andpressurization, etc) that determines
the in-plane effectiveness of the AVBs.”
A.5 - SCE claims, “The facts identified in this analysis
indicate that even though the Unit 3 tube bundle components (tubes and AVBs)
might have been fabricated and assembled better, the tube "to" AVB
gaps built gaps might have been in fact larger in the Unit 3 RSGs as suggested
by the ECT results. Based on this, it cannot be ruled out that the tube-to-AVB
gaps are larger and more uniform in the Unit 3 RSGs than the Unit 2 RSG’s. This
might have resulted in reduction of the tube-to-AVB contact force and consequently
in multiple consecutive AVB supports being inactive. Inactive tube support
might have resulted in "tube-to-tube" wear.
A.6 – DAB Safety Team Conclusions:
SONGS Unit 3 RSG’s were operating outside SONGS Technical Specification Limits
for Reactor Thermal Power and Current Licensing Basis for Design Basis Accident
Conditions. Let us summarize what these experts are saying: (1) AREVA is
saying, “After fluid elastic instability develops, the amplitude of in-plane
motion continuously increases and the forces needed to prevent in-plane motion
at any given AVB location become relatively large. Hence shortly after
instability occurs, U-bends begin to swing in Mode 1 and overcome hindrance at
any AVB location”, (2) Westinghouse is saying, “A small AVB gap that would be
considered active in the OP mode would also be active in the IP mode because
the small gap will prevent significant in-plane motion due to lack of clearance
(gap) for the combined OP and IP motions. Thus, a contact force is not required
to prevent significant IP motion”, (3) MHI is saying, “High steam flows and
cross-flow velocities combined with narrow tube pitch-to-diameter ratio caused
elastic deformation of the U-tube bundle from the beginning of the Unit 3
cycle, which initiated the process of tube-to-AVB wear and insufficient contact
forces between tubes and AVBs. Tube bundle distortion is considered a major
contributing cause to the mechanism of tube-to-tube/AVB/TSP wear seen in the
Unit 3 SG’s. After 11 months of wear, contact forces were virtually eliminated
between the tube and AVBs in the areas of highest wear of Unit 3 tubes as
confirmed by ECT and visual inspections. According to MHI Technical Document, the
RSG Anti-bar Vibration Structures were only designed for out-of plane
vibrations and not in-plane vibrations”, (4) John Large is saying, “Its impossible to reliably predict
the effectiveness of the many thousands of AVB contact points for when the tube
bundle is in a hot, pressurized operational state.” In essence, all the three
NEI qualified, “U.S. Nuclear Plant Designers” are saying: (a) Firstly, these
AVB structures are not designed to prevent in-plane vibrations, (b) Secondly,
once fluid elastic instability develops, the AVB is not strong enough to
prevent large in-plane amplitude of tubes,
(c) Thirdly, excessive fluid-induced random vibrations and fluid
hydrodynamic pressures (Mitsubishi Flowering Effect) cause the loss of tube-AVB
contact forces and increase the tube-to-AVB gaps and the onset of fluid elastic
instability develops, and (d) Lastly, tube-to-AVB gaps and contact forces are
irrelevant to prevent fluid elastic instability from progressing and causing
tube-to-tube and tube-to-AVB wear. Therefore,
based on a review of MHI, AREVA and Westinghouse excerpts shown above, it is
concluded, that FEI, flow-induced random vibrations and MHI Flowering effect
redistributed the tube-to-AVB gaps in Unit 3 RSG’s as measured by SCE. Westinghouse
even goes this far to state, “None of the MHI manufacturing issues were
extensively treated in the SCE root cause evaluation.” Hence, It is concluded
that the NRC and SCE claims that insufficient contact forces in Unit 3
Tube-to-AVB Gaps ALONE caused tube “to” tube wear are misleading, erroneous and
designed to put the blame on MHI for purposes of making SCE look good in the
public’s eyes and for collecting insurance money from MHI’s so-called “manufacturing
defects”.
B.
Let us now examine the effects of modeling errors, that the SONGS replacement
generators were not designed with adequate thermal hydraulic margin to preclude
the onset of fluid-elastic instability.
B.1 – NRC AIT Report states, “The
ATHOS thermal-hydraulic model predicts bulk fluid behavior based on first
principals and empirical correlations and as a result, it is not able to
evaluate mechanical, fabrication, or structural material differences or other
phenomena that may be unique to each steam generator. Therefore this analysis
cannot account for these mechanical factors and differences which could very
likely also be contributing to the tube degradation.”
B.2 – Ivan Cotton states, “Fluid
elastic instability is one of the most damaging types of instabilities
encountered in heat exchangers and steam generators and can impose a severe
economic penalty on the power and chemical industries. At present our
understanding of the mechanisms leading to fluid-elastic instability is very
limited and more experiments are needed to more fully delineate the conditions
for the onset of fluid-elastic instability.” Such experimentation should only
be done in a sealed lab, NOT our environment with the lives of eight million
local residents at stake in the outcome!
B.3 – Ishihara, Kunihiko and Kitayama
state, “Tube vibrations become large as tube thickness/diameter ratio (T/D)
increases and tube length/diameter ratio (L/D) decreases, and the tube
vibrations strongly depend on the dynamic characteristics of tubes such as the
natural frequency and the damping ability.” In the case of SanO’s replacement
SGs, the tubes, especially in the U-bend region, were too close together,
poorly restrained, poorly damped, along with too much heat flux and an inappropriate
pressure-to-flow ratio, along with other causes which resulted in FEI and FIV.
B.4
– Fairewind states, “Realistically, the 3-D steam analysis is not accurate
enough to apply to such important safety related determinations. To make such
mathematical risk 3-D analysis, a very large margin of error must be applied,
and that has not been done. For example, if the 3-D steam analysis determines
that plugging 100 tubes is a solution, then plugging ten times that number
might be the appropriate solution due to the mathematical errors in the 3-D analysis
being applied by Edison and Mitsubishi. Edison has taken many of its
mathematical/computer models as “gospel” rather than accepting their wide
margins of error, which directly affect safety, which is precisely what got
them in trouble in the first place. It should have been obvious to Edison
that MHI FIT-III has not been benchmarked, and had not been previously used in
licensing procedures showing that the use of FIT-III might have an adverse
effect on the FSAR safety analysis thus necessitating the entire license
amendment review and public hearing process. As noted by the AIT, Edison
approved the use of FIT-III code even though the code was not benchmarked, nor
identified as acceptable in the FSAR. Consequently, Edison operated San Onofre
without knowing the uncertainties in the Replacement Steam Generators’
performance characteristics. Predicted liquid levels, pressure drops,
vibrations, and temperatures at both Units 2 and 3 were all subject to unknown
uncertainties during both normal and abnormal operations. In my opinion, by approving the use of an un-benchmarked
and untested design tool like FIT-III, Edison did not did not meet the
requirements expected from a nuclear licensee.
Use of an un-benchmarked computer code that is not included in
the FSAR protocol demands a formal
FSAR license amendment process including the requisite public hearings.”
B.5
– Mitra, V.K. Dhir, I. Catton state, “Flow induced vibrations in heat exchanger
tubes have led to numerous accidents and economic losses in the past. Efforts
have been made to systematically study the cause of these vibrations and
develop remedial design criteria for their avoidance. Instability was clearly
seen in single phase and two-phase flow and the critical flow velocity was
found to be proportional to tube mass. It is also found that nucleate boiling
on the tube surface is also found to have a stabilizing effect on fluid-elastic
instability.”
B.6 - John Large, says “Factual Issues v) & vi) –
SONGS SG Comparison to Other Operating SGs: I identify a number of issues with
the representation of Figures 4-3 and 5-1 of the AREVA Tube-to-Tube Report,
including: i) it is not exactly clear which properties are being represented on
the spider diagram for comparison with the other operational SGs; even so ii)
since it is most unlikely that AREVA has undertaken a comprehensive (ATHOS)
simulation of each of the five nominated SGs, the comparisons drawn are likely
to be between aggregate or bulk flows within the entire tube bundle of each SG;
iii) as acknowledged by AREVA, the SONGS RSGs are dominated by in-plane flow
regimes whereas all other SGs are characterized by out-of-plane flow regimes;
and iv) none of the comparative SGs has been identified. In other words, unless the spider diagrams of
Figure 4-3 and 5-1 somehow, and I cannot reason how, are making a direct
comparison of the complex two-phase fluid cross-flow situation in the SONGS and
other five comparative plant steam generators, then these figures only provide
the bases of a somewhat meaningless comparisons.”
B.7 – SCE states that SONGS Unit 3
Damage (FEI) was caused due to outdated MHI Thermal-Hydraulic Computer Models.
According to NRC AIT Report, SONGS did not specify the value of FEI in its
Design and Performance Specifications SO23-617-1. Academic Researchers have
discussed and warned about the adverse effects of fluid elastic instability
(tube-to-tube wear) in nuclear steam generators since the 1970’s. Westinghouse
and Combustion Engineering (CE) have designed CE replacement steam generators
(RSGs) to prevent the adverse effects of fluid elastic instability since 2000’s
(e.g., PVNGS).
B.8 – The NRC AIT Report dated
November 9, 2012 states, “the FIT-III thermal-hydraulic model was still
in-progress at the time of the inspection and no final conclusions were reached
for the cause of the non-conservative flow velocities, which were used as inputs
in the tube vibration analysis and resulted in non-conservative stability
ratios. Since the licensee had not completed the cause evaluation for this
unresolved item, the inspectors were not able to make a final determination of
whether a performance deficiency or violation of NRC requirements occurred. The
inspectors were informed that Mitsubishi was performing an evaluation of
the potential factors that contributed to the low flow velocities in
FIT-III relative to the velocities calculated by the ATHOS model developed
after the tube leak event in Unit 3. This evaluation was included in Document
SO23-617-1-M1530, Revision 1, which also intended to demonstrate the validity
of FIT-III results for the original tube vibration analysis. This evaluation
was still being finalized and not yet approved by Edison. The licensee and
Mitsubishi continued to evaluate this unresolved item and no final conclusions
were reached at the time of the inspection. The NRC is continuing to perform
independent reviews of existing information, and will conduct additional
reviews as new information becomes available.” In another related finding, NRC
inspectors stated, “SCE Engineers did not meet Procedure SO123-XXIV-37.8.26
requirements to ensure the design of the retainer bar was adequate with respect
to the certified design specification. Specifically, the licensee failed to
ensure that there was sufficient analytical effort in the design methodology of
the anti-vibration bar assembly to support the conclusion that tube wear would
not occur, as a result of contact with the retainer bars due to flow-induced
vibration. The inspectors determined that the requirements for flow-induced
vibration in the certified design specification, along with the expectations in
Procedure SO123-XXIV-37.8.26, provided sufficient information to reasonably
foresee the inadequate design of the retainer bars during the review and
approval of design Calculations SO23-617-1-C749 and SO23-617-1-C157, including
the associated design drawings provided by Mitsubishi.”
B.9 – Arnie Gundersen states, “Not
only is Mitsubishi unfamiliar with the tightly packed CE design, but Edison’s
engineers added so many untested variables to the new fabrication that this new
design had a significantly increased risk of failure. As a result of the very
tight pitch to diameter ratios used in the original CE steam generators,
Mitsubishi fabricated a broached plate design that allows almost no water to
reach the top of the steam generator.
The maximum quality of the
water/steam mixture at the top of the steam generator in the U-Bend region
should be approximately 40 to 50 percent, i.e. half water and half steam. 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.
Because
of the Edison/Mitsubishi steam generator changes, the top of the new steam
generator is starved for water therefore making tube vibration inevitable.
Furthermore, the problem appears to be exacerbated by Mitsubishi’s
three-dimensional thermal-hydraulic analysis determining how the steam and
water mix at the top of the tubes that has been benchmarked against the
Westinghouse design but not the original CE design.
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 testing. 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.”
B.10 – Comment on Limitations
of ATHOS thermal-hydraulic Models: The ATHOS thermal-hydraulic model predicts
bulk fluid behavior based on first principals and empirical correlations and as
a result, it is not able to evaluate mechanical, fabrication, or structural
material differences or other phenomena that may be unique to each steam
generator. Furthermore, the combination of the omission of the in-plane AVB
restraints, the unique in-plane activity levels of the SONGS RSGs, together
with the very demanding interpretation of the remote probe data from the cold
and depressurized tube inspection, render forecasting the wear of the tubes and
many thousands of restraint components when in hot and pressurized service very
challenging indeed. ATHOS
thermal-hydraulic Models used for 70% power have not been benchmarked, and
tested against SONGS Unit 2 RSG degraded tube bundles performance for several
cycles of depressurized/pressurized operation.
Hence, ATHOS analyses cannot accurately predict the behavior of pressurized
degraded SG tube bundles and their interaction with their anti-vibration bar
support structure, which could very likely contribute to unknown amounts of
tube-to-tube wear and/or AVB degradation in Unit 2, at 70% or 100% power during
a main steam line break accident resulting in a cascade of multiple steam
generator tube ruptures.
B.11 – Conclusions on Modeling
Errors: The NRC AIT Report dated November 9, 2012 states, “the FIT-III
thermal-hydraulic model was still in-progress at the time of the inspection and
no final conclusions were reached for the cause of the non-conservative flow
velocities, which were used as inputs in the tube vibration analysis and
resulted in non-conservative stability ratios. Since the licensee had not
completed the cause evaluation for this unresolved item, the inspectors were
not able to make a final determination of whether a performance deficiency or
violation of NRC requirements occurred. SCE and MHI are both negligent because
they did a very poor job of Industry and Academic Research benchmarking
regarding the applicability of thermal-hydraulic computer models during the
redesign of San Onofre’s original CE SGs. SCE is negligent because they did not
check the results of MHI’s outdated Thermal-Hydraulic Computer Models to meet
their specification and procedure requirements. This does not meet the NRC
Chairman’s Standards. Therefore, it is concluded that SCE claims as stated
above are not factual. SCE engineers did not check the work of MHI with a
critical and questioning attitude and did not meet the San Onofre Design
Procedures, 10CFR50, Appendix B, Quality assurance Standards and or NRC
Regulations. NRC AIT Team jumped the
boat by putting all the blame on MHI in July 2012 and now they are retreating
in November 2012 by stating with a sunken face, “The inspectors were not able
to make a final determination of whether a performance deficiency or violation
of NRC requirements occurred.”
C.
Let us now examine the other differences between Unit 2 and Unit 3’s
Operational Factors, which were significant contributors to the “fluid-elastic
instability” in San Onofre Unit 3 and the tube-to-tube wear resulting in the
tube leak.
C.1 – Adverse Design/Operational
Factors responsible for Fluid Elastic Instability: Low steam generator
pressures (SONGS RSGs range 800-850 psi, the primary cause of the onset of severe
vibrations) allow the onset of FEI, whereby U-tube bundle tubes start vibrating
with very large amplitudes in the in-plane directions. Extremely hot and
vibrating tubes need a little amount of water (aka damping, 1.5% water,
steam-water mixture vapor Fraction 99.5%). Without the water, the extremely hot
and vibrating tubes cannot dissipate their energy. In effect, one unstable tube
drives its neighbor to instability through repeated violent impact events which
causes tube leakage, tube failures at MSLB test conditions and/or unprecedented
tube-tube wear, Tube-to-AVB/Tube Support Plates wear, as we saw in San Onofre
Unit 3. So in review, due to narrow tube pitch to tube diameter, tube natural frequency,
low tube clearances, in certain portions of the RSGs U-tubes bundle, fluid
velocities exceed the critical velocities due to extremely high steam flows
(100% power conditions). These high fluid velocities cause U-tubes to vibrate
with very large amplitudes in the in-plane direction and literally hit other
tubes with repeated and violent impacts. Due to lower secondary steam operating
pressures (required to generate more heat, electricity and profits) and
excessive pressure drops due to high flows and velocities, steam saturation
temperature drops. This lowering of steam saturation temperature combined with
high heat flux in the hot leg side of the U-tube bundle causes steam dry-outs
to form (Vapor fraction >99%), known as “NO Effective Thin Tube Film
Damping.” Thin film damping refers to the tendency of the steam inside the
generators to create a thin film of water between the RSG tubes and the support
structures and each other. That film is enough to help keep the tubes from
vibrating with large amplitudes, hitting other tubes violently, and to protect
the Anti-Vibration Bar support structures and maintain the tube-to-AVB gaps and
contact forces. These adverse conditions in Unit 2 at 70% power operation (RTP)
with the present defective design and degraded RSGs, known as fluid elastic
instability (Tube-to-Tube Wear, or TTW) can lead to rapid U-tube failure from
fatigue or tube-to-tube wear in Unit 2 due to a main steam line break as seen
in Unit 3’s RSG’s. In summary, FEI is a phenomenon where due to San Onofre RSGs
design intended for high steam flows causes the tubes to vibrate with
increasingly larger amplitudes due to the fluid effective flow velocity
exceeding its specific limit (critical velocity) for a given tube and its
supporting conditions and a given thermal hydraulic environment. This occurs
when the amount of energy imparted on the tube by the fluid is greater than the
amount of energy that the tube can dissipate back to the fluid and to the
supports. The lack of Nucleate boiling on the tube surface or absence of water
is found to have a destabilizing
effect on fluid-elastic stability.
C.2 – Unit 2
FEI Conflicting Operational Data
·
NRC AIT Report SG Secondary U2/3 Pressure
Range 833 - 942 psi
·
SCE RCE SG Secondary U2/3 Pressure - 833 psi
·
RCE Team Anonymous Member - Unit 2 SG
Secondary Pressure 863 psi
·
SONGS SG System Description Unit 2 SG
Pressure Range 892 - 942 psi
·
Westinghouse OA SG Secondary U2/3 Pressure ~
838 psi, Void Fraction 99.55%
·
SCE Enclosure 2, MHI ATHOS results - U2/3
Void Fraction 99.6%
·
SCE Enclosure 2, Independent Expert results -
ATHOS U2/3 Void Fraction 99.4%
·
DAB Safety Team SG Secondary U2 Pressure 863
-942 psi, Void Fraction 96-98%
·
SONGS Plant Daily Briefing Unit 3 Electrical
Generation – 1186 MWe
·
SONGS Plant Daily Briefing Unit 2 Electrical
Generation – 1183 MWe
C.3 – Unit 2
FEI Conclusions
C.3.1 - NRC AIT
Report - Operational Differences between U2/3 - 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.
C.3.2 - SCE
Unit 2 Restart Report Enclosure 2 Conclusions - Because of the similarities in
design between the Unit 2 and 3 RSGs, it was concluded that FEI in the in-plane
direction was also the cause of the TTW in Unit 2.
C.3.3 - SCE U2
FEI SONGS RCE Team Anonymous Member Conclusions – FEI did not occur in Unit 2
C.3.4 -
Westinghouse OA Conclusions: (a) An evaluation of the tube-to-tube wear
reported in two tubes in SG 2E089 showed that, most likely, the wear did not
result from in-plane vibration of the tubes since all available eddy current
data clearly support the analytical results that in-plane vibration could not
have occurred in these tubes, and (b) Operational data – Westinghouse ATHOS
Model shows no operational differences in Units 2 & 3 (void fraction
~99.6%) and then Westinghouse says in (a) above that FEI did not occur in Unit
2. Westinghouse
is contradicting its own statement.
C.3.5 - AREVA
OA Conclusions - Based on the extremely comprehensive evaluation of both Units,
supplemented by thermal hydraulic and FIV analysis, assuming, a priori, that
TTW via in-plane fluid-elastic instability cannot develop in Unit 2 would be
inappropriate.
C.3.6 - John Large States, “I note here that there
are three clear conflicts of findings between the OAs: From AREVA that
AVB-to-tube and TTW result from in-plane FEI, contrasted to Westinghouse that
there is no in-plane FEI but most probably it was out-of-plane FEI, and from
MHI that certain AVB-to-tube wear results in the absence of in-plane FEI from
just turbulent flow. My opinion is that such conflicting disagreement over the
cause of TTW reflects poorly on the depth of understanding of the crucially
important FEI issue by each of these SCE consultants and the
designer/manufacturer of the RSGs.”
C.3.7 - DAB
Safety Team Conclusions – Due to higher SG pressure (Range 863 – 942 psi) and
lower thermal megawatts as compared to Unit 3, FEI did not occur in Unit 2. This is consistent with the position of RCE
Team Anonymous Member. The NRC AIT
Report, SCE, Westinghouse, MHI, SCE chosen “Independent” Experts and AREVA
conclusions on Unit 2 FEI are inconsistent, confusing and inconclusive. The NRC
AIT Report, SCE and AREVA conclusions on Unit 2 FEI are unacceptable.
C.3.8 - The NRC San Onofre Special Review Panel should direct other
branches within the NRC (NRC-RES and/or the ACRS) to review the above data
without any prior “turf” bias and present their findings to the public for
review and comment prior to any restart decision being made by the NRC.
Dangers of SAN
ONOFRE Unit 2 Restart
1 - Chairman
Allison Macfarlane said Unit 2 would not be permitted to restart unless the NRC
has reasonable assurance it can be operated safely. Let us examine that scenario
below and determine whether NRC can have that reasonable assurance or not:
2 - Let us
examine what John Large says, “There is no account of the changes that have
been made in the evaluation of the tube structural and leakage integrity, that
is from the stage of predicting those tubes at risk of TTW and other forms of
wear, the tube thinning wear rates, through to the nature of the tube failure
being unique to the type and extent of the wear pattern and tube thinning; and
the methods of deducing, mainly by unproven inference, from the probe
inspection results particularly to determine the in-plane AVB effectiveness,
includes unacceptably large elements of test and experimentation that are
inconsistent with the analyses and descriptions of the FSAR. I provide a number
of explicit examples where I consider that the circumstances and risks accompanying
the proposed restart of Unit 2 will result in unacceptable levels of test and
experiment.” What he is saying is that these degraded tube bundles cannot
prevent multiple tube ruptures from fluid elastic instability as we saw by the
failure of 8 tubes in Unit 3 RSGs under Main Steam Line Break (MSLB) test
conditions.
3 - Let us
examine what Arnie Gundersen says, “Eight replacement steam generator tubes
failed their pressure tests in 2012 and more than 1,000 others have been
plugged. Therefore, a review of the evidence makes it clear that the San Onofre
Replacement Steam Generator tube damage discovered in 2012 was so severe and
extensive that both reactors have been operating in violation of their NRC FSAR
license design basis as defined in their Technical Specifications. The Main Steam
Line Break with radiological leakage through the steam generator tubes is one
of the bounding conditions in emergency plan evaluation and the extent of steam
generator tube failures directly impacts the FSAR analysis. The Replacement
Steam Generator (RSG) modifications at San Onofre increased both the likelihood
of equipment failure and the radiological consequence of such failure and
therefore directly affect the FSAR Current Design Basis.”
4 – Unit 2 Main Steam Line Break (MSLB)
Scenario: The most severe design basis accident to meet the San Onofre Unit 2
TS 5.5.2.11.b.1 steam generator structural integrity is a MSLB at the first
weld outside containment. This assumption minimizes the flow resistance between
the break and the affected SG and maximizes the mass & energy (M&E)
release. The analyses focus on M&E releases at licensed Rated Thermal Power
(RTP or 100% Power). The outside containment case includes the assumption that
the main steam isolation valve (MSIV) in the steam line with the least flow
resistance fails to close following the main steam isolation signal (MSIS).
This assumption maximizes the M&E release during a MSLB outside of the
containment. Super-heating within the SG initiates upon U-tube uncovery as
specified in the NRC Information Notice 84-90. The turbine stop valves are
assumed to close instantaneously at the time of the reactor trip. This
assumption is conservative for a MSLB event because the entire steam inventory
at the time of reactor trip is assumed to be forced out the break in 300
seconds (5 minutes). No Operator action outside Control Room can be assumed, if
it takes less than 30 minutes. Westinghouse states, “it should be understood that there
is more mass of secondary coolant in the steam generator at no load than at
full power. Therefore, no load is the worst case for steam line break analyses.”
The
depressurization of the non-isolable steam generator would result in 100% void
fractions in the degraded Unit 2 U-Tube bundle due to instant flashing of the
sub-cooled 440 degrees Fahrenheit feedwater into steam. This condition of ZERO
Water in the steam generators would cause fluid elastic instability (FEI),
flow-induced random vibrations and excessive hydrodynamic pressures (Mitsubishi
Flowering Effect). The force of the flashing steam would create high-energy
jets, lifting loose parts and debris present in the steam generator, which
would do additional damage by cutting holes into the already degraded tubes and
creating additional loading (See Note 1 below) on the tube support plates
(TSPs) due to heavy build-up of deposits on trefoil/quadrifoil-shaped holes
from SG blowdown and crack the high cycle fatigued U-bend tubes not supported
by Anti-Vibration Bars (AVB). These cumulative adverse conditions in all
likelihood would result in a massive cascading of RSG’s tube failures (tubes
would excessively rattle or vibrate, hitting other tubes with violent impacts)
due to extremely low tube-to-tube clearances and no effective or non-existent
in-plane anti-vibration bar support protection system. This jackhammering
effect would involve hundreds of degraded active SG tubes along with all the
inactive (plugged /unstabilized) tubes causing a catastrophic amount of
simultaneous tube leaks/ruptures. Under this adverse scenario, 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 to
fifteen minutes from a broken steam line would EXCEED the SONGS NRC approved
offsite radiological release doses safety margins based on assumption of a
single tube rupture in the SONGS FSAR. So, in essence, these RSG’s are like
loaded guns, or a Fukushima-type nuclear accident, waiting to happen. Any
failure under these conditions would allow significant amounts of radiation to
escape to the atmosphere and a major Loss of Coolant Accident (LOCA) could
easily result causing much wider radiological consequences and even a potential
nuclear meltdown of the reactor.
SCE states, “A
MSLB alone does not generate sufficient differential pressure to cause tube
rupture. The differential pressure across the SG tubes necessary to cause a
rupture will not occur if operators (See Note 2 below) prevent RCS
re-pressurization in accordance with Emergency Operating Instructions.” SCE’s
suggested DID Actions and proven unreliable operator actions to detect a leak
and/or to re-pressurize the steam generators as claimed by Edison are not
practical to stop a major nuclear accident from occurring in Unit 2 in the
first 5-15 minutes of a MSLB during the proposed 5-month trial period.
NOTES:
1. Plugging of
the at-risk tubes is not a satisfactory solution because it is the retainer bar
that vibrates via random fluid flow processes at sub FEI critical velocity
levels – these are likely to continue in play or, indeed, exacerbate at the
proposed U2 restart at 70% power, leading to through-tube abrasion, the
detachment of tube fragments, lodging at other unplugged and in-service tube
localities, resulting in the so-called ‘foreign object’ tube wear. This
additional loading would exceed: (1) the safety factor of 3.0 against burst
under normal steady state full power operation primary-to-secondary pressure
differential and a safety factor of 1.4 against burst applied to the design basis
accident primary-to-secondary pressure differentials, and (2) significantly
affect burst or collapse pressures determined and assessed in combination with
the loads due to a safety factor of 1.2 on the combined primary loads and 1.0
on axial secondary loads.
2. SCE’s
suggested “defense-in-depth” actions are insufficient to stop multiple tube
ruptures due to the short duration of a main steam line beak event. Human
performance weaknesses, such as mis-diagnoses, substantial delays in isolating
the faulted steam generator, communication errors and delayed initiation of the
residual heat removal system, have been identified in past events at SONGS and
other US Nuclear Power Plants. The events also involved unnecessary radiation
releases, lack of RCS subcooled margin, excessive RCS cooldown rates, and
overfilling the SG because of human or procedural problems.
CONCLUSIONS: Until the NRC can determine that San
Onofre is 100% safe to operate at its approved rated power, granting any Unit 2
Restart testing is unacceptable, because if a nuclear accident occurred during testing
who would be held liable, the Nuclear Utilities, the Insurance Carriers, the Federal
Government, the State of California, the CPUC, the NRC Commissioners, NRC
Region IV, EIX/SCE Shareholders & Employees or just the millions of
affected southern Californians? The
DAB Safety Team believes that once the true amount of existing tube fatigue and
all other associated damage is quantified, anything short of a total SG rebuild
and/or SG replacement will be unacceptable prior to any restart being
authorized by the NRC.
Special
Comments about San Onofre
- San Onofre
is rated by the Institute of Nuclear Operations (INPO) as an INPO 4 Plant
(The Worst Nuclear Plant Rating) and it should also should be rated
in NRC Region IV Response Column V (Worst rating) and not in the NRC
Response Column I (Best Nuclear Plant Rating).
2. San Onofre is the worst nuclear plant
in the country with the worst safety record, worst retaliation record, an INPO
4 rating and it is a mockery to place it in NRC Response Column I. NRC
Region IV by listing San Onofre in NRC Response Column I, is putting its
credibility on line and is displaying clear trends of collusion with SCE. It would
be informative to learn who made the decision on San Onofre’s current ranking
and why…
Definitions of
NRC Response Columns [Source: NRC Inspection Manual Chapter 0305]
Column I – All
Assessment Inputs (Performance Indicators (PIs) and Inspection Findings) Green;
Cornerstone Objectives Fully Met
Column II – One or Two White Inputs (in different cornerstones) in a Strategic
Performance Area; Cornerstone Objectives Fully Met.
Column III – One Degraded Cornerstone (2 White Inputs or 1 Yellow Input) or any
3 White Inputs in a Strategic Performance Area; Cornerstone Objectives Met with
Moderate Degradation in Safety Performance
Column IV – Repetitive Degraded
Cornerstone, Multiple Degraded Cornerstones, Multiple Yellow Inputs, or 1 Red
Input; Cornerstone Objectives Met with
Longstanding Issues or Significant Degradation in Safety Performance
Column V. Overall Unacceptable Performance; Plants Not Permitted to Operate
Within this Band, Unacceptable Margin to Safety Licensed Activities
###
Definitions, Abbreviations
and Acronyms
·
10 CFR 50.59 Safety Evaluation -
Section 50.59 of Title 10 of the Code of Federal Regulations (10 CFR 50.59)
defines the conditions under which reactor licensees may make changes to their
facilities or make changes to procedures or to conduct tests or experiments
without prior NRC approval. In general, such changes, tests, or experiments may
be carried out unless they would involve a change to the technical
specifications or an unreviewed safety question (as defined in 50.59(a)(2)). [Source: World Wide Web]
·
AIT: NRC’s Augmented Inspection Team
·
AREVA: Nuclear engineering firm
owned by French Atomic Energy Commission
·
ATHOS is a three-dimensional
computational fluid dynamics (CFD) code for analyzing steam generator (SG)
thermal-hydraulic (TH) performance characteristics [Source: Westinghouse OA]
·
AVB: Anti Vibration Bar
·
CFR: Code of Federal Regulations
·
CPUC: California Public Utilities
Commission
·
DBA: Design Basis Accident
·
ECT: Eddy Current Testing
·
EIX – Edison International
·
EPRI: Electric Power Research
Institute
·
FEI: Fluid Elastic Instability is a
phenomenon where the tubes vibrate with increasingly larger amplitudes due to
the fluid effective flow velocity exceeding its specific limit (critical
velocity) for a given tube and its supporting conditions and a given thermal
hydraulic environment. This occurs when the amount of energy imparted on the
tube by the fluid is greater than the amount of energy that the tube can
dissipate back to the fluid and to the supports, because of lack of squeeze
film damping, nucleate boiling or presence of localized regions in U-Tube
bundles with high vapor fractions (e.g., >99.6%, steam voids, steam
dry-outs, etc.) [Source: MHI/DAB]
·
FIRV: Flow-Induced Random Vibrations
is a phenomenon where the tubes vibrate due to forces created by turbulent flow
as a result of fluid velocity and density fluctuations. Vibration amplitudes
due to random vibration are generally small (smaller than those due to tube
fluid-elastic instability). [Source: MHI]
·
FSAR: Final Safety Analysis Report
·
FSM:
Fluid elastic Stability Margin
·
FWLB:
Feed-Water Line Break
·
John Large is a Chartered Engineer,
a Consulting Engineer, a Fellow of the Institution of Mechanical Engineers,
Graduate Member of the Institution of Civil Engineers and Fellow of the Royal
Society of Arts. [World Wide Web]
·
LOCA: Loss Of Coolant Accident
·
MHI: Mitsubishi Heavy Industry
·
MSIV: Main Steam (line) Isolation
Valve
·
MSLB: Main Steam Line Break
·
MWt: Mega-Watts Thermal
·
NOPD: Normal Operating Pressure
Differential
·
NRC: Nuclear Regulatory Commission
·
NRC Reasonable Assurance is the
recognition that “adequate protective measures can and will be taken in the
event of a radiological emergency.” Reasonable assurance is based on licensees
complying with NRC regulations and guidance, as well as licensees and offsite
response organizations demonstrating that they can effectively implement
emergency plans and procedures during periodic evaluated exercises. [Source:
www.nrc.gov]
·
NRC San Onofre Special Review Panel
- The NRC has established a special panel to coordinate the agency’s evaluation
of Southern California Edison Co.’s proposed plan for restarting its Unit 2
reactor and ensuring that the root causes of problems with the plant’s steam
generators are identified, and addressed. [Source: US NRC Blog]
·
NRR: NRC’s Office of Nuclear Reactor
Regulations
·
OD: Outer Diameter
·
P/D: Pitch to Diameter ratio
·
OSG: Original Steam Generator
·
RCE: Root Cause Evaluation
·
RCPB: Reactor Coolant Pressure
Boundary
·
RCS: Reactor Coolant System
·
RSG: Replacement Steam Generator
·
RWST: Refueling Water Storage Tank
·
SCE: Southern
California Edison
·
SG: Steam Generator
·
SGTR: Steam
Generator Tube Rupture
·
SM:
Stability Ratio
·
SONGS: San Onofre
Nuclear (Waste) Generating Station (alternate abbreviation: SONWGS)
·
TSP: Tube Support
Plate
·
TTS: Top-of-Tube
Sheet
·
TTW: Tube-to-Tube Wear
·
TW: Tube Wear
·
Westinghouse -
Westinghouse Electric Company provides fuel, services, technology, plant
design, and equipment for the commercial nuclear electric power industry.
The DAB Safety Team: Don, Ace and a BATTERY of safety-conscious San Onofre
insiders plus industry experts from
around the world who wish to remain anonymous. These volunteers assist
the DAB Safety Team by sharing knowledge, opinions and insight but are not
responsible for the contents of the DAB Safety Team's reports. We
continue to work together as a Safety Team to prepare additional: DAB Safety Team
Documents, which explain in detail why a
SONGS restart is unsafe at any power level without a Full/Thorough/Transparent
NRC 50.90 License Amendment and Evidentiary Public Hearings. For more
information from The DAB Safety Team, please visit the link above.
Our Mission: To prevent a Trillion Dollar Eco-Disaster like Fukushima,
from happening in the USA.
Copyright January 28, 2013 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 Attorney