Press Release 13-01-31 Four NRC Augmented
Team Conflicts of Interest
The DAB Safety Team:
January 31, 2013
Media Contact: Don
Leichtling (619)
296-9928 or
Ace Hoffman (760)
720-7261
Four More Statements
From NRC Region IV Augmented Inspection Team (AIT) That Require A Nuclear Reactor Regulation (NRR) Investigation And
Resolution.
The DAB Safety Team Has Transmitted
The Following Request To The Offices Of Chairman Of The NRC, The California
Attorney General and Senator Barbara Boxer’s Committee on Environment and
Public Works (EPW).
1. NRC AIT in its report
dated November 09, 2012 (Re: NRC ADAMS
Library Accession Number ML 2012010 - Unresolved Item 05000362/2012007-03, “Evaluation of Unit 3 Vibration and Loose
Parts Monitoring System Alarms (V&LPM)”) closed the referenced item by
stating that, “The inspectors determined that the licensee properly responded
to and evaluated the alarms and followed the applicable station alarm
procedures and vendor recommendations. Subsequently, the licensee requested from
the vendor an in-depth evaluation of the available acoustical data, which was
documented in Nuclear Notification NN 201818719. This evaluation established
the likely source of the alarms. The results were inconclusive because of
limitations with the monitoring system. Specifically, because of sensor
locations (lower portion of the steam generator below the tube sheet in the
support structure) and sensitivity, it was not possible to determine the exact
source of the Unit 3 alarms. Westinghouse engineering personnel performed an
evaluation (Evaluation 201818719-SPT-2) of acoustical data and determined from
the shape and intensity of the particular
responses that the acoustic source was not
likely from the upper bundle of the
replacement steam generator or related to the
tube-to-tube wear. The licensee (SCE) is considering additional sensor
locations which are not required, but may help with monitoring the upper bundle
region of the steam generator during power operation. The results of this
additional monitoring and increased sensor sensitivity may provide the licensee
with a potential means to monitor for tube-to-tube degradation.” (See Page 14
for the limitations of this improved version of V&LPM system related to NO
detection capability of tube-tube wear as claimed by AIT Team and SCE and
questioned by NRR below). According to
the December 18, 2012 SCE NRC Public meeting Press and Webcast Reports, Edison officials came under sharp questioning
about the Vibration
and Loose Parts Monitoring System monitors at a U.S. Nuclear Regulatory Commission
panel meeting in Maryland. Richard Stattel of the NRC’s Nuclear Reactor
Regulation (NRR) Instrumentation Branch told the Edison Officials in a roaring
and loud voice on an international live web cast, “The equipment could not do
the job described by the company or provide additional safety if the plant is
restarted. The instrumentation that you're proposing ... does not appear to be
capable of detecting the conditions that would lead to actual tube wear."
Edison depicted the equipment in its restart plan as an important safety
measure "but it doesn't appear to do that.” See the DAB Safety Team’s Press
Release + 12-12-28 Thirty Alarms Demonstrates SONGS Unsafe for details on this subject.
DAB Safety Team
Comments: The NRR is saying loud and clear
that both NRC AIT and SCE Engineers need to understand the basic functions of
“Safety-Grade” Instrumentation and the concept of “tube-to-tube” wear (Fluid
Elastic Instability).
Since there are no means of monitoring tube wall thinning
while the plant is in service, the risk of tube burst is wholly dependent upon
the accuracy and reliability of SCE’s “Safety-Grade” Instrumentation. The DAB Safety Team has stated
earlier that NRC AIT Report is just a replication of SCE Root Cause Evaluation
and not a true assessment by an Independent Regulator tasked with ensuring
Public Safety.
On
December 21, 2012, the US Nuclear Regulatory Commission (NRC)
blog posted a letter from Chairman Macfarlane titled, “A Visit to Japan: Reflections from the Chairman.” She said, “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.” According to the March 16, 2012 Press reports, Senators
Barbara Boxer (D-CA), Chairman of the Senate Environment and Public Works
Committee (EPW), and Dianne Feinstein (D-CA) sent a letter to the Chairman of
the Nuclear Regulatory Commission (NRC), Dr. Gregory Jaczko, calling on the NRC
to perform a thorough inspection at the San Onofre plant, located in San
Clemente. The collusion and casual relationship between NRC
AIT Team and SCE requires an Investigation by the Offices of NRC Chairman and Senator
Barbara Boxer to determine the impact on both future US reactor operations and
emergency preparedness planning. This investigation by the AIT does not
meet the NRC Chairman’s Standards.
2. NRC AIT in its
report dated November 11, 2012 (Re: Unresolved
Item 05000362/2012007-03, "Evaluation of Retainer Bars Vibration during
the Original Design of the Replacement Steam Generators”) closed
the referenced item by stating that, “The inspectors determined that
the licensee’s failure to verify the adequacy of the retainer bar design as
required by SONGS Procedure SO123-XXIV-37.8.26 was of very low safety
significance (Green) based on NRC Inspection Manual Chapter 0609.04, “Phase 1 –
Initial Screening and Characterization of Findings,” and Inspection Manual
Chapter 0609, Appendix A, “The Significance Determination Process (SDP) for
Findings At-Power,” because the finding did not involve a degraded steam
generator tube condition where one tube could not sustain 3 times the
differential pressure across a tube during normal full power, steady state
operation and none of the replacement steam generators violated the “accident
leakage” performance criterion in plant Technical Specifications as a result of
the retainer bar vibrations. The licensee also implemented actions to inspect
all affected tubes in Unit 2 and 3 and remove from service all those tubes
surrounding the smaller retainer bars that could wear due to vibration of the
retainer bar. Because this violation has been determined to be of very low safety
significance (Green) and has been entered in the licensee’s corrective action
program as SONGS Nuclear Notification (NN) 201843216, it will be dispositioned
as a non-cited violation in accordance with Section 2.3.2 of the NRC’s
Enforcement Policy.”
John Large, internationally known Consulting Engineer,
Chartered Engineer, Fellow of the Institution of Mechanical Engineers, Graduate
Member of the Institution Civil Engineers, Learned Member of the Nuclear
Institute and a Fellow of the Royal Society of Arts states concerning SONGS Restart Unit 2 in
his testimony to the Atomic Safety Licensing Board, “In October 2012 MHI
reported directly to the NRC safety concerns about the retainer bars: The
Steam Generator tube wear adjacent to the retainer bars was identified as
creating a potential safety hazard.
The maximum wear depth is 90% of the tube thickness. The cause of the tube wear
has been determined to be the retainer bars' random flow-induced vibration
caused by the secondary fluid exiting the tube bundle. Since the retainer bar
has a low natural frequency, the bar
vibrates with a large amplitude. This type tube wear could have an
adverse effect on the structural integrity of the tubes, which are part of the
pressure boundary. The plugging of the tubes that are adjacent to the retainer
bars was performed. MHI has recommended to the purchaser [SCE] to remove the
retainer bars that would have the possibility of vibration with large amplitude
or to perform the plugging and
stabilizing for the associated tubes. According
to MHI, it is the lower resonance frequency of the smaller diameter retainer bars that is susceptible to
turbulent two-phase flow exciting the bar into its prime resonance or some
harmonic frequency thereof [p10, item 3].14 Whatever, a number of the tubes
capturing the retainer bar had
sustained abraded wear from interaction with it. These tubes comprised six
tubes in U2 and four tubes in U3, with seven tubes in total showing wear
greater than the 35% limit of the tube wall thickness for which isolation from
service is required by plugging with, as previously noted, an incidence site in
one of U2 RSGs having worn through 90% of its wall thickness. I agree with the
findings of MHI that the tube wear at the retainer bar localities arises
because of random flow induced (not FEI) vibration of the retainer bar itself, it being
entirely independent of any tube motion excited from other sources. However,
MHI’s advice to either plug the local tubes and/or remove the retainer bars at
risk raises two issues unique to the retainer bar and its sub-assembly: (i)
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; (ii) MHI’s
recommendation that those retainer bars at risk of large-amplitude fluid flow
excited vibration should be removed is, of course, dependent upon reliable
analysis to identify the at-risk assemblies; and, importantly, and (iii) this
restraint system probably also serves to contain the tube bundle geometry
during a main line steam break (MSLB) design basis event, so any change or
removal of the retaining bar assemblage would require a full safety
justification.”
Westinghouse states, “For most of the straight leg section of the
tube, the gap velocities at lower power levels and at 100% power are similar.
The recirculating fluid flow rate is relatively constant at all power levels. However,
in the U-bend region, the gap velocities are a strong function of power level.
The steam flow in the bundle is cumulative and increases as a function of the
power level and the bundle height which causes high fluid quality, void
fraction, and secondary fluid velocities in the upper bundle.”
SCE in
its November 30, 2012, NRC Presentation stated, “Four
tubes with retainer bars wear above 35% limit in Unit 2 were plugged.” The NRC
website states, “The severity of one of the wear indications at a
Unit 2 retainer bar was significant enough (90 percent thru-wall) to warrant
in-situ pressure testing. This pressure test confirmed the structural integrity
of this tube (there was no leakage).” (Above
Bolding Added)
DAB
Safety Team Comments: Let us summarize what John Large and
Westinghouse are saying: (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 to vibrate 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/or in-service tube localities, resulting in the
so-called ‘foreign object’ tube wear, (2) For most of the straight leg section
of the tube, the gap velocities at lower power levels and at 100% power are
similar. Therefore, even at 70% power, the tube-to-retainer bar wear will
continue at the same rate as 100% power and plugging the tubes is not a
satisfactory solution in terms of reducing the active tubes rupture safety
risks. SCE is not stating the facts either in its Root Cause Evaluation nor in
its NRC Presentation. Two better questions are, “How many tubes in Unit 2 have
what amounts of fatigue cracks and why has SCE not used state-of-the-art
technology to visually examine all RSG tubes at San Onofre?” What this really means is that Southern
Californians were lucky once again, that Unit 2 just happened to be shutdown
for refueling! Otherwise, one or more
worn tubes could have leaked or failed due to a design bases accident and/or any
unanticipated transients. Almost 180 tubes had to be plugged and stabilized
in Unit 2 Replacement Steam Generators due to retainer bar design mistakes. In
addition, no reports are available to determine the extent of tube fatigue
damage or damage to the small retainer bars caused by the worn tubes and
whether the damaged retaining
bars are strong enough to restrain the movement of the anti-vibration bar
assembly during a main steam line break design basis event (Ref: NRR RAI #32). The design of the retainer bars approved by
SCE and manufactured by MHI clearly violated the Code of Federal Regulations,
10 CFR Part 50, GDC 14, “RCPB—shall have “an extremely low
probability of abnormal leakage…and gross rupture” and Appendix B, Criterion
III, “Design Control.” The DAB Safety Team’s opinion is that NRC
AIT is treating the retainer bar mistakes and its design approval by SCE just
as a routine matter like “No big deal, nothing happened, so who cares” instead
of performing the strict enforcement required of an Independent Regulator
tasked with ensuring Public Safety. This investigation by
the AIT does not meet the NRC Chairman’s Standards.
3. NRC AIT in its
report dated November 11, 2012 (Re: Unresolved
Item 05000362/2012007-08, “Non-Conservative Thermal-Hydraulic Model Results”)
states that, “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
the original Report in July 2012, the NRC AIT concluded that, “Due to
modeling errors, the SONGS replacement generators were not designed with
adequate thermal hydraulic margin to preclude the onset of fluid-elastic
instability.” NRC
John Large states, “I identify a number of issues with
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, … 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. A complete understanding of the causation of the in-plane FEI
is essential to ensure that the SONGS Unit 2 plant is acceptably safe to
restart and, once restarted, predictably safe to continue in operation over the
proposed 150 day inspection interval. To the contrary, the understanding
presented by SCE is neither comprehensive nor convincing. In my opinion, simply
sweeping the FEI issue under the carpet on the basis of (in- or out-of-plane)
FEI will not reoccur at 70% power is not only disingenuous but foolhardy.”
Arnie Gundersen states, “The AIT report indicated that the change to the FIT-III evaluation
methodology was not discussed as part of Edison’s
50.59 screening because the details of thermal hydraulic models used for the design of the OSG were not discussed in the
original FSAR. It should have been obvious to Edison that 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 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.”
Gundersen further states, “The AIT reported that
FIT-III predictions differed considerably in comparison to an Electric Power
Research Institute developed code named ATHOS. FIT-III predicted lower flow
velocities and void fractions that were not conservative compared to ATHOS. The
AIT Report neglected an analysis of the root cause of the critical differences
between FIT-III and ATHOS, and the negative impact such lax calculational
modeling had on the design, fabrication, and successful operation of the San
Onofre RSGs. Had Edison sought the required FSAR license amendment, differences
between FIT-III and ATHOS would have been identified six years ago. The AIT did
not address the possibility that the lack of conservatism in FIT-III
predictions, in addition to causing tube vibrations, could also result in
non-conservative predictions of the behavior of the steam generator pressure
vessel and associated main steam piping during accident conditions that are
required to be analyzed in the FSAR. The AIT noted that the non-conservatisms
in FIT-III are a contributor to the failure by Edison to adequately calculate
the San Onofre RSG tube vibrations. But equally important, the AIT failed to
address that FIT-III could also create non-conservative predictions of the
behavior of the steam generator pressure vessel and associated main steam
piping during accident conditions that are required to be analyzed in the FSAR.
Such a conclusion implies that damage to the steam generator pressure vessel
itself, and not just the tubes, might have occurred at San Onofre and remains
unanalyzed by either Edison or the NRC. The probability of an accident
exceeding the plant’s Current Design Basis is increased by the radically
different Edison Replacement Steam Generators. Hence, the risks involved in
operating the San Onofre RSGs should have been addressed as part of an FSAR
license amendment and hearing process. It is my professional opinion that
Edison should have applied for the 50.59 process so that the FSAR license
amendment evaluation and public hearings would have occurred six years ago,
prior to creating an accident scenario and facing losses that by the end of
this process will easily total more than $1 Billion. The seriousness of the
licensing and safety impact of the damaged RSGs at San Onofre cannot be
overstated or underestimated. Any Design Basis Accident (DBA) as defined in the
FSAR needs to be accurately modeled in order to protect public health and
safety. The FSAR’s DBA analysis including the extent of tube leakage in the
event of a Main Steam Line Break significantly impacts the design and implementation
of Emergency Evacuation Plans. In the event of a steam line break accident in
the San Onofre Replacement Steam Generators with the degraded condition of the
tubes, an accident would have occurred that is more severe than any design
basis accident scenario previously analyzed by Edison in the FSAR. Such a DBA
steam line break accident would render the San Onofre emergency plan totally
inadequate and most likely cause a
permanent evacuation of a large portion of Southern California.”
DAB Safety
Team Comments: After the June 18, 2012 public Meeting, the
NRC AIT Team Chief announced to the world, "The computer simulation used by Mitsubishi during the design of the
steam generators had under-predicted velocities of steam and water inside the
steam generators by factors of three to four times." Now, six months later, the AIT Team is saying
the matter is unresolved. The AIT Team is
just repeating what SCE says or is not sure what they said four months
ago. ATHOS Modeling results are not
reliable, because the results by NRC AIT Team, Westinghouse, MHI, AREVA and
Independent Experts show that fluid elastic instability occurred both in Units
3 and 2. The investigations in the Root cause of SONGS Unit 3 FEI
regarding computer modeling have not been completed by NRC AIT Team, SCE and
MHI. FEI did not occur in Unit 2 according to DAB Safety Team and
Westinghouse. As also shown in other DAB Safety Team reports, FEI was not
caused in Unit 3 by tube-to AVB gaps as claimed by NRC AIT Team and SCE.
This is consistent with the findings of Westinghouse, AREVA, MHI, John Large
and SONGS Anonymous Insiders. The AIT Team is hurting its own credibility by issuing
contradicting and conflicting statements. This investigation by
the AIT does not meet the NRC Chairman’s Standards.
performed by Mitsubishi in
determining flow and heat transfer by Mitsubishi in d
4. NRC AIT report dated
November 11, 2012 (Re: Unresolved Item
05000362/2012007-10, “Evaluation of Departure of Methods of Evaluation for 10
CFR 50.59 Processes”) closed the referenced item by stating: (a) The change from ANSYS to ABAQUS did not
require a license amendment prior to implementing the change, so with respect
to section 2.10.D.6 of the NRC Enforcement Manual, there is no reasonable
likelihood that the change from ANSYS to ABAQUS would ever require NRC
approval. Therefore, in accordance with the NRC Enforcement Manual, the
inspectors determined that the licensee’s change from ANSYS to ABAQUS was a
minor violation of 10 CFR 50.59(d)(1), and (b)n
Based on this, the inspectors determined that the licensee had changed from
using ANSYS and STRUDL to analyze several events for the original steam
generators, to using only ANSYS to analyze a single limiting event for the
replacement steam generators. Therefore, because the licensee did not change
the method described in the Updated Final Safety Analysis Report, the
inspectors concluded that the licensee did not need to obtain a license
amendment prior to implementing that change. In the original Report in July 2012, the NRR technical specialist
reviewed SCE’s 10 CFR 50.59 evaluation and found two instances that failed to
adequately address whether the change involved a departure of the method of
evaluation described in the updated final safety analysis report: (a) Use of
ABAQUS instead of ANSYS: The SCE’s 50.59
evaluation incorrectly determined that using the ABAQUS instead of ANSYS was a
change to an element of the method described in the updated final safety
analysis report did not constitute changing from a method described in the
updated final safety analysis report to another method, and as such, did not
mention whether ABAQUS has been approved by the NRC for this application.
(b) Use of ANSYS instead of STRUDL and
ANSYS: While SCE’s 50.59 evaluation
correctly considered this a change from a method described in the FSAR to
another method, the 50.59 evaluation did not mention whether the method has
been approved by NRC for this application.
NRC AIT Report states, “For the Unit 2 and
Unit 3 replacement steam generators, the licensee determined that the proposed
activity did not adversely affect a design function, or the method of
performing or controlling a design function described in the updated final
safety analysis report. The licensee evaluated the following updated final
safety analysis report design functions in the 50.59 screening: Steam Generator
Design Functions. Let us examine the effect of these changes on Steam Generator
Design Functions: The design functions of the steam generators tubes and tube supports
are to: (1.) Limit tube flow-induced vibration to acceptable levels
during normal operating conditions, and (2) Prevent a tube rupture concurrent
with other accidents.
Change
Number 1:
105,000 square feet tube heat transfer area in OSGs; 116,100 square feet tube
heat transfer area in RSGs; 11.1% increase in heat transfer area, which is more
than a minimal change of 10% in the non-conservative direction. Change
accomplished by addition of 377 tubes in the central region by removal of stay
cylinder and increasing the length of 9727 tubes by > 7 inches in each of
the four RSGs.
Change
Number 2:
Operating Secondary Pressure – OSGs: 900 psi, RSG: 833 psi ~ 10% change – A catastrophic
change for onset and ongoing exponential fluid elastic instability.
Change
Number 3:
Tube wall thickness was reduced from 0.048 inches to 0.043 to pump more reactor
coolant through the tubes > 11.6% change.
The latest academic research indicates that the tube vibrations become
large as T/D decreases and L/D increases, because the in-plane tube vibrations
strongly depend on the dynamic characteristics of tubes such as the natural
frequency and the damping ability.
Four other changes: Moisture content was
reduced from 0.2% to 0.1% to improve SG performance, RCS Volume was increased
from 1895 cubic feet to 2003 cubic feet, RCS Flow was increased from 198,000
gpm to 209,000 gpm, feedwater flow was increased from 7.4 million pound per hour
to 7.6 million pound per hour and AVBs were not designed to prevent against
adverse effects of fluid elastic instability (In-plane vibrations, Tube-to-Tube
wear, steam dry-outs). These unapproved and unanalyzed changes were
claimed to be a conservative decision and improvements in the RSGs from OSGs
were presented as a "like for Like" change. No mixing baffles
were added in the SONGS RSGs to improve the T/H Performance in the RSGs.
FEI and SR Values were not provided by SCE in the RSG Design Specifications.
SCE told MHI to avoid the NRC Approval… MHI neither provided in-plane
supports, nor provided the operational criteria to prevent FEI in one of the
largest steam generators with such high steam flows. MHI did not
benchmark CE SG Computer codes or design details, neither did SCE, nor did SCE
check the work of MHI. And Dr. McFarlane says, “SCE is responsible for
the work of its vendors and contractors. Look at Palo Verde RSGs, a
Success Story and SONGS RSGs, a $ Billion Blunder…. For complete list of
changes, as identified by Arnie Gundersen and the DAB Safety Team see The Big
#1 Attachment.
NRC AIT
Report states,
“The licensee’s bid specification required that the stay cylinder feature of
the original steam generators be eliminated to maximize the number of tubes
that could be installed in the replacement steam generators and to mitigate
past problems with tube wear at tube supports caused by relatively cool water
and high flow velocities in the central part of the tube bundle. Mitsubishi
employed a broached trefoil tube support plates instead of the egg crate
supports in the original design. In addition to providing for better control of
tube to support plate gaps and easier assembly, the broached tube support
plates were intended to address past problems with the egg crate supports by
providing less line of contact and faster flow between the tubes and support
plates, reducing the potential for deposit buildup and corrosion.”
Arnie
Gundersen states,
“As the NRC confirmed in its AIT report, a large steam void has developed near
where the additional tubes were added in the Replacement Steam Generators
(called fluid elastic instability) that allows many types of excess vibrations
to occur. Fairewinds review of Edison’s Condition Report clearly shows that the
location within the steam generators where the steam “fluid elastic
instability” has developed is precisely the region where the extra heat created
by the 400 new tubes would create an excess of steam and various vibrational
modes.”
NRC AIT
report states,
“Mitsubishi’s preliminary explanation of the failure mechanism started with the
combination of two factors: (1) a relatively small tube pitch to tube diameter
ratio (P/D), and (2) high void fraction in the tube bundle area where the
tube-to-tube wear was identified. The small pitch to diameter ratio was a fixed
parameter in the replacement steam generators established by the nominal
center-to-center distance between adjacent tubes (P) and the nominal outside
diameter of the tubes (D). The high void fraction was identified from the
results of Mitsubishi’s thermal-hydraulic model for the secondary side of the
replacement steam generators. Mitsubishi considered that the combination of
these two factors may have resulted in favorable conditions for in-plane tube
vibration based, in part, on the results of recent studies in fluid-elastic
instability.” Mitsubishi also states, “Low secondary pressures are severe for
vibration.”
John
Large states,
“Referring to the short section of the FSAR provided to me by SCE, which I
understand is not to be amended for the Unit 2 restart: (a) 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 (b) 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.”
05000362/2012007-10, “Evaluation
of Departure of Method of
John Large states, “SCE’s assertion that reducing
power to 70% will at the best alleviate, but not eliminate, the TTW and other
modes of tube and component wear is little more than hypothesis - the
supporting Operational Assessments and analyses have not proven it to be otherwise.
I am of the opinion that trialling this hypothesis by putting the SONGS Unit 2
back into service will, because of the uncertainties and unresolved issues
involved, embrace a great deal of change, test and experiment. The terms of the Confirmatory Action
Letter of March 11 2012, are versed such that to meet compliance the
response of SCE via its Return to Service Report,11 must include
considerable changes of conditions and procedures that are outside the
reference bounds of the present FSAR – this is because the physical condition
of the RSGs, and the means by which this is evaluated and projected into future
in-service operation, have substantially and irrevocably changed since the
current FSAR was approved. The fact that SCE fails to satisfy the requirements
of the CAL is neither here nor there, although it illustrates the scope and
complexity of the response required. At the time of preparing the CAL, the NRC
being well-versed in the failures at the San Onofre nuclear plant, surely must
have known that the only satisfactory response to the CAL would indeed require
considerable changes, tests and experiments to be
implemented.”
DAB Safety Team Comments: Therefore, the DAB Safety Team concludes that
the changes in design functions of the RSGs tubes and tube supports described
above definitely: a) did not limit tube flow-induced vibration to acceptable
levels during normal operating conditions and, b) involved a significant
reduction in a margin of safety – Failure of 8 Unit 3 SG Tubes under MSLB test
conditions and significant TTW > 35% of ~381 tubes in Unit 3 RSGs. A multiple tube failure event, if actually
would have occurred during a MSLB would have resulted in a significant increase
in the off-site radiological consequences over the single tube burst event, if currently
considered in the SONGS approved FSAR by NRC Region IV. 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. The AIT has no business
contradicting conclusions made earlier by the NRR technical
specialist. This investigation by
the AIT does not meet the NRC Chairman’s Standards.
NRC
Region IV Response to DAB Safety Team Analysis of SONGS 10 CFR 50.59 Evaluation
Comments: The
NRC has already conducted several reviews of the 10 CFR 50.59 documents
associated with the replacement of the steam generators at SONGS. These reviews
involved NRC inspectors from multiple offices including Region IV, Region II
and the Office of Nuclear Reactor Regulation at NRC headquarters. The results
of these reviews are contained in NRC two inspection reports that are available
at http://www.nrc.gov/info-finder/reactor/songs/tube-degradation.html. [see the
Augmented Inspection Team Report dated July 18, 2012, and the Augmented
Inspection Team Follow-Up Report dated November 9, 2012]. It is worthy of note
that the NRC staff is currently reviewing 10 CFR 50.59 documents associated
with the licensee’s proposed restart activities. The results of the ongoing
review will be documented in a future inspection report. NRC
Comments from Mel Silberberg [NRC-RES, Retired [Chief, Severe Accident Research
Branch; Waste Management Branch] to Region IV: I am
disappointed in the composition of the special panel! Where is the
representation from NRC-RES? The issues at SONGS involve thermal hydraulics and
material science. The NRC-RES and its contractors are experts in these areas.
The Office of Research was created by the Congress for such situations. Two RES
staff covering these disciplines and one or two consultants, serving as
peer-reviewers. Perhaps there needs to be a separate peer review. Public
confidence can only be gained using logical, informed measures as I described
above. Inspection Reports are only one
facet of the problem, no question. However, understanding the reasons for the
fluid instability, possible cavitation corrosion effects, etc. are phenomena
which require evaluation by T/H as well as materials experts, with appropriate
oversight by the ACRS. The SCE, the nuclear industry, the NRC and the public
need assurance, not educated guesses. I have not seen a bona fide attempt to
understand resolve the issue such that all can be alert to potential problems.
I still remain puzzled as to why the ACRS [at least one of the
Subcommittees]. I am trying to reach the
ACRS Exec. Director to discuss this point. Thank you.” reviews of the 10 CFR
50.59 documents associated with the replacement of the steam generators at
SONGS. These reviews involved NRC inspectors from multiple offices including
Region IV, Region II and the Office of Nuclear Reactor R
DAB Safety Team Comments:
NRC Region IV Inspectors need to be re-trained in interpretation of
significance of 10 CFR 50.59 Evaluation rules and meaning of changes in design
function on safety evaluations. Simply sweeping the 10 CFR 50.59 mistakes under the carpet on
the basis of meaningless statements, “The NRC has already conducted several
reviews of the 10 CFR 50.59 documents associated with the replacement of the
steam generators at SONGS. The present
SONGS NRC approved for the total S/G tube leakage assumes a limit of 1 gpm for
all S/Gs, which ensures that the dosage contribution from the tube leakage will
be limited to a small fraction of 10CFR100 limits in the event of either a S/G
tube rupture or steam line break. The 1 gpm limit is consistent with the
assumptions used in the analysis of these accidents.
The
0.5 gpm (720 gpd) leakage limit per S/G ensures that S/G tube integrity is
maintained in the event of a main steam line rupture or under LOCA conditions.”
These reviews of 10 CFR 50.59 and SONGS FSAR S/G tube rupture limits from NRC
inspectors from multiple offices including Region IV, Region II and the Office
of Nuclear Reactor Regulation at NRC headquarters” are not only disingenuous but
foolhardy. A single tube leakage and/or rupture could
result in a nuclear incident or accident with tube leakages assumed in the
current SONGS FSAR as shown in the Table below. A multiple tube failure event
for all phases of the reactor in-core fuel cycle, would result in a significant
increase in the off-site radiological consequences (e.g., Fukushima, Chernobyl,
etc.) over the single tube burst event currently considered in the FSAR. The
rapid and extraordinarily severe wear that resulted in the 2012 failures of all
of Edison’s San Onofre Replacement Steam Generators was the result of Edison’s
2005 decision to radically change the RSG design and to claim that the Part
50.59 licensing process did not apply. Arnie Gundersen and
DAB Safety Team have stated consistently that 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. While
the NRC Augmented Inspection Team (AIT) briefly described how Edison addressed
its 50.59 requirements, the evidence shows that Edison did not comply with the
NEI guidelines for implementing 50.59. Published reports indicate that the
strategic decision made by Edison that the 50.59 process would not be applied
to the RSGs was made by corporate officials before any engineering
personnel had actually performed the 50.59 engineering analysis. Consequently,
Edison made a management decision to claim that the 50.59 process did not apply
and therefore San Onofre was not required to seek NRC approval for the proposed
changes at San Onofre Units 2 and 3. These unlicensed unapproved design
changes to the containment boundary violated Federal Regulations and therefore
the FSAR must be amended prior Unit 2 Restart to reflect multiple steam
generator tube ruptures with MSLB plus DBE due to Edison’s significant untested
and unanalyzed modifications.
History of Steam Generator Tube Ruptures
Year
|
Plant
|
Location
|
Flow Rate
|
Rupture or Leak
|
1975
|
Point Beach 1
|
Wisconsin
|
125 gal/min
|
Rupture
|
1976
|
Surry 2
|
Virginia
|
330 gal/min
|
Rupture
|
1979
|
Prairie Island 1
|
Minnesota
|
390 gal/min
|
Rupture
|
1982
|
Ginna
|
New York
|
630 gal/min
|
Rupture
|
1987
|
North Anna 1
|
Virginia
|
600 gal/min
|
Rupture
|
1989
|
McGuire 1
|
North Carolina
|
500 gal/min
|
Rupture
|
1991
|
Mihama 2
|
Japan
|
Undetermined
|
Rupture Level 3 INES Incident –
Offsite Releases
|
1993
|
Palo Verde 2
|
Arizona
|
240 gal/min
|
Rupture
|
2000
|
Indian Point 2
|
New York
|
90 gal/min
|
Rupture
|
2004
|
Crauss NP
|
France
|
Unknown
|
Leak
|
2005
|
||||
2006
|
||||
2012
|
SONGS 3
+ SONGS 2
had a tube with 90% Wear
|
California
|
625-750 gal/min
(See Note 1 below)
|
Leak, almost 8+ Ruptures
|
8 + Ruptures ~=
|
||||
5000 - 6000 gal/min1
|
NOTE 1: Arnie Gundersen
states, “Edison dramatically increased the radiation risk to the public as a
result of San Onofre with Replacement Steam Generators that were extremely
flawed beginning with their original design. The fact that 8 tubes failed the
pressure tests in Unit 3 indicates that those tubes would have failed during a
main steam line break (MSLB). It is uncertain if a reactor operator would have
been able to shut the plant down without melting the core. A simultaneous
rupture of 8 tubes would have caused a primary to secondary leak of radioactive
coolant of about 5000-6000 gallons per minute. This leakage would have begun to
drain the nuclear core as well as releasing radioactive primary coolant to the
atmosphere. The ability of a reactor operator to control the water level in the
affected steam generator with this high leakage rate and keep the nuclear
reactor core cooled has never been analyzed or tested. An accident of this
magnitude is outside ANY reactor’s Current Design Basis (CDB).”
ABBREVIATIONS AND
ACRONYMS
·
ACRS:
NRC’s Advisory Committee on Reactor Safeguards
·
ADAMS:
NRC’s Agencywide Documents Access and Management System
·
AIT:
NRC’s Augmented Inspection Team
·
AREVA:
Nuclear engineering firm owned by French Atomic Energy Commission
·
AVB: Anti
Vibration Bar
·
CAL: Confirmatory Action
Letter
·
CFR: Code
of Federal Regulations
·
CPUC:
California Public Utilities Commission
·
DBA: Design Basis Accident
·
ECT: Eddy Current Testing
·
ECCS: Emergency Core Cooling System
·
EDF: French nuclear parts manufacturing company, also owns transmission
lines in France, etc.
·
EPRI: Electric Power Research Institute
·
FEI: Fluid Elastic Instability
·
FIRV: Flow-Induced Random Vibrations
·
FSAR: Final Safety Analysis Report
·
FSM:
Fluid elastic Stability Margin
·
FWLB: Feed-Water Line Break
·
GDC: General Design Criteria
·
GSI: Generic Safety Issue
·
ID: Inner Diameter
·
INES: International Nuclear Events Scale
·
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
·
NPP: Nuclear Power Plant
·
NRC: Nuclear Regulatory Commission
·
NRR: NRC’s Office of Nuclear Reactor Regulations
·
OD: Outer Diameter
·
P/D: Pitch to Diameter ratio
·
OSG: Original Steam Generator
·
PRA: Probabilistic Risk Assessment
·
PVNGS: Palo Verde Nuclear Generating Station
·
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 Generating Station
·
SR: The specific mechanism for the flow-induced vibration has been
determined to be a fluid-elastic instability.
The fluid-elastic mechanism has a significant effect on tube response in
cases where the fluid-elastic stability ratio equals or exceeds 1.0. The stability ratio, SR, is defined as: SR = V eff / V c, where V eff is the
effective cross-flow velocity and V c is the critical velocity beyond which the
displacement response to the tube increases rapidly.
·
TS: Technical Specifications (for operation of a NPP)
·
TSP: Tube Support Plate
·
TTS: Top-of-Tube Sheet
·
TTW: Tube-to-Tube Wear
·
TW: Tube Wear
·
UT: Ultrasonic Testing
·
V&LPM: The Vibration and Loose Parts Monitor - consists of Vibration
and Loose Parts Channels. Each system consists of piezoelectric sensors,
preamplifiers, a signal processor unit and other peripheral equipment. The
vibration and loose parts monitoring system is designed to provide continuous
monitoring and conditioning of loose parts accelerometer signals. Two separate
accelerometers are installed on each of the steam generators. The location of
these instruments are on the steam generators’ lower supporting structures and
provide acoustic information about loose parts impacts specifically on the
reactor coolant or primary side of the steam generators. The vibration and
loose parts monitoring system real time functions consist mainly of impact
alarm validation of suspected loose part events and recording acoustic data. The improved Westinghouse DMIMS-DX™
systems are installed at Millstone 3, Krsko, Diablo Canyon 1 & 2, Wolf
Creek, and Beaver Valley 1, which are all Westinghouse NSSS plants, and at
Crystal River 3, which is a B&W plant. DMIMS-DX™ provides fast, reliable
detection of loose part impacts within the Reactor Coolant System (RCS), while
minimizing the generation of false alarms. This monitoring system is a greatly
enhanced version of the previous Westinghouse DMIMS system, employing the latest
digital technology and offering significant operational advantages to our
customers. Loose parts monitoring is
based on listening for the impact of loose parts against fixed components
within the primary system as they are propelled by the coolant flow. This
application appears simple on the surface, but its effective implementation is
not an easy task. The noises typical of an operating plant can generate false
alarms that reduce operator confidence, interfere with normal operations, and
cause unnecessary expense. The Westinghouse DMIMS System uses a patented
algorithm to determine the metallic characteristics typical of loose parts.
This algorithm and the associated alarm algorithms, together, minimize the
generation of false alarms and have established a reputation for reliability
within the 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 31,
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
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