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Last original steam generators being moved from San Onofre to Utah starting tonight

Wednesday, December 5th, 2012
Issue 49, Volume 16.

SAN ONOFRE - The last of four original steam generators being moved from the San Onofre Nuclear Generating Station to Utah for storage will be trucked out beginning tonight, according to plant operators.

Southern California Edison, which operates the plant on the northern San Diego County coast, replaced the generators two years ago with equipment made by Japan's Mitsubishi Heavy Industries. Two of the generators were moved to the storage facility in Clive, Utah, last year, while a third shipped out earlier this year.

The facility has been shut down since the end of January, when a leak was discovered in a steam tube in one Advertisement
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of the reactors. The other reactor was already off-line for maintenance. Neither has been restarted.

Opponents of plans to get the reactors going again contend that regulators failed to take design changes by Mitsubishi into account.

Moving the original steam generators is no easy task. The 700,000-pound generator is fitted on a specially configured 400-foot-long truck for the three- week trip to Utah. Travel through California will take place at night, according to SCE.

The utility said the low level of radiation in the generator is roughly the same as a dental x-ray if someone were to stand within 5 to 10 feet for one hour.



Comment Profile ImageBill Hawkins
Comment #1 | Wednesday, Dec 5, 2012 at 5:06 pm
SCE’s misleading and embarrassing technical presentation at their NRC public meeting when explaining fluid elastic instability in SONGS Unit 2, creates doubts about Edison’s ability to perform a safe restart of their damaged “Radiation Steaming Crucibles”.

Due to high steam flows, high fluid velocities, narrow tube clearances, areas of top of U-tube bundle in a nuclear steam generator have no water, as seen in SONGS 3. When this happens, fluid elastic instability occurs and the thin tubes move with large sprinting amplitudes and hit the neighboring tubes with violent and repeated impacts. Poorly designed nuclear steam generator such as SONGS, have NO in-plane anti-vibration bar supports and water to protect tubes from hitting each other. Therefore, cascading tube ruptures can occur and cause a reactor Meltdown. SONGS Unit 2 “As-designed and Defective Radiation Steaming Crucibles have NO in-plane anti-vibration bar supports, have extremely low tube clearances and many unanticipated operational occurrences and MSLB even at 70% power, can cause the entire u-tube bundle to be devoid of water and cascading tube ruptures. Result is Fukushima In Southern California Backyards.

After spending almost a year, wasting hundreds of millions of dollars of rate payer’s money and hiring World’s Best Experts, Southern California Edison brought out SONGS Senior Vice President of Engineering, Tom Palmisano, on November 30, 2012 (Friday night) to give an intense technical presentation and a dazzling performance in a bid to convince the regulators they should be able to restart Unit 2 at the plant. However, Palmisano failed miserably to convince the DAB Safety Team, NRC Region IV Panel and the Public/Technical Experts, that Southern Californians will be safe from the effects of a potential nuclear radiological accident by the restart of Unit 2’s “Sick and Unsafe” Steam Generators by operating them at 70% power for a trial period of 5 months. When questioned by NRC Panel Members, the unprepared and uncomfortable Palmisano tasked with convincing the NRC, the Public and the News Media that Unit 2 is safe for restart “stumbled several times during the presentation”, then in response to the panel members questions for further technical details said sometimes irritatingly, “I will get back to you” and then finally “hesitantly” admitted, “SCE and MHI analysis are still not yet complete and will continue for several months for a detailed analysis and investigation to this problem.”

San Onofre’s “As designed and defective” replacement steam generators (aka radiation steaming crucibles) without replacement or adequate repairs (replacement of tube bundle and anti-vibration supports) are unsafe and fail to meet the Steam Generator Fundamental Tube Integrity Criteria for a Main Steam Line Break Accident and Other Unanticipated Operational transients.

According to public sources, Pete Dietrich has already announced in internal SONGS staff meetings that Unit 2 will tentatively restart on February 2, 2013. If confirmed, this is stunning news for the public and gives the clear perception of collusion between the NRC and SCE. TEPCO’s reported collusion with Japanese Regulators resulted in Fukushima’s catastrophic nuclear accident (a Trillion Dollar Japanese economic, public health & safety radiation nightmare and Eco-Disaster).

Copyright December 5, 2012 by The DAB Safety Team. All rights reserved. This material may not be published, broadcast or redistributed without crediting the DAB Safety Team. The contents cannot be altered without the Written Permission of the DAB Safety Team Leader and/or the DAB Safety Team’s Attorneys.
Comment Profile ImageOld Fallbrook local
Comment #2 | Thursday, Dec 6, 2012 at 3:23 pm
Dollie Parton has had a Masectomy. The other "boob" is probably also diseased.
Comment Profile ImageBill Hawkins
Comment #3 | Friday, Jan 25, 2013 at 9:41 pm
Special Public Awareness Series – San Onofre Nuclear Generating Station (SONGS) Unit 2 Restart - Comments about the NRC Augmented Inspection Team San Onofre Report
Courtesy of DAB Safety Team

The DAB Safety Team’s goal is to help 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 even been determined. Public does not know the status of SCE, MHI updated and ongoing cause evaluations, SCE’s response to 32 NRR’s RAI’s and NRC’s Special San Onofre Inspections. The NRC rushing to a faulty judgment cannot be allowed to compromise Public Safety just to please profit-motivated SCE, as this conflicts with President Obama’s Policy, the new NRC Chairman’s Standards and the advice of NRC retired Branch Chief’s who have also spoken out.

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 Pane 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 an 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 the changes to the manufacturing/fabrication of the tubes and other components which may have resulted in increased clearance between the anti-vibration bars and the tubes; (3) Due to modeling errors, the SONGS 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 tube wear and accelerated 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).”
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.”
A.5 – 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. MHI states that 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 area of Unit 3 wear as confirmed by ECT and visual inspections. According to MHI Technical Document, RSG Anti-bar Vibration Structure was only designed to sustain the adverse effects in the out-of plane vibrations and not in-plane vibrations. Therefore, based on a review of MHI, AREVA and Westinghouse excerpts shown below, it is concluded, that FEI and MHI Flowering effect redistributed the tube-to-AVB gaps in Unit 3 RSG’s. It is concluded that 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 manufacturing so called defects.

B. Let us now examine of 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.”
B.4 – Fairewinde 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.”
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 – 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 1970’s. Westinghouse and Combustion Engineering (CE) have designed CE engineering replacement steam generators (RSGs) to prevent the adverse effects of fluid elastic instability since 2000’s (e.g., PVNGS).
B.7 – 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.8 – 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.9 – 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 bundle and anti-vibration bar support structure, which could very likely contribute to the tube-to-tube wear and AVB degradation at the Unit 2, 70% power and main steam line break with multiple steam generator tube ruptures.
B.10 – Conclusions on Modeling Errors: 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 SONGS 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 10CFR50, Appendix B, Quality assurance Standards and or NRC Regulations.

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 SONGS 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) cause at 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%). When this happens, the extremely hot and vibrating tubes cannot dissipate their energy and return to their original in-plane design position. In effect, one unstable tube drives its neighbor to instability through repeated violent and turbulent 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 SONGS Unit 3. So in review, due to narrow tube pitch to tube diameter, tube 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% SONGS power conditions outside the industry NORM). These high fluid velocities cause U-tubes to vibrate with very large amplitudes in the in-plane direction and literally hit other the tubes with repeated and violent impacts. Due to lower 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 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. That film is enough to help keep the tubes from vibrating with large amplitudes, hitting other tubes violently, and protect the Anti-Vibration Bar support structures and maintain the tube-to-AVB gaps and contact forces. These adverse conditions in SONGS at 70% power operation (RTP) with the present defective design and degraded of RSGs known as fluid elastic instability (Tube-to-Tube Wear, or TTW) can lead to rapid U-tubes failure from fatigue or tube-to-tube wear in Unit 2 due to a main steam line break as seen in SONGS Unit 3 RSG’s. In summary, FEI is a phenomenon where due to SONGS 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
DAB Safety Team SG Secondary U2 Pressure Range 863 -942 psi
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 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 – ATHOS Model shows no differences in Units 2 & 3
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 – DAB Safety Team Conclusions – Due to higher SG pressure (Range 863 – 942 psi) and lower thermal megawatts compared to Unit 3, FEI did not occur in Unit 2. This is consistent with the position of RCE Team Anonymous Member. NRC AIT Report, SCE, Westinghouse and AREVA conclusions on Unit 2 FEI are inconsistent, confusing and inconclusive.
C.3.7 – DAB Safety Team Request: NRC San Onofre Special Panel investigation required.
To be continued.. Part 4 – Original SONGS Combustion Engineering Steam Generators
Comment Continued : The comment above was written from the same location.
Post Continued
Comment Profile ImageBill Hawkins
Comment #4 | Tuesday, Jan 29, 2013 at 7:42 pm
Courtesy of HelpAllHurtNeverBaba US NRC Blog
January 29, 2013 at 1:04 am

To: NRC Moderator Mr. Victor Dricks, Senior Public Affairs Officer, NRC Region IV
Request for independent re-review of SONGS 50.59 Screen/Evaluation by NRC Region II – Please send me an email after you complete the review ASAP. These guys who performed the screen and evaluations are very close friends of mine and I want to make sure they were on the right track. Trying to help my friends and NRC Region IV. Thanks… HAHN Baba


Moderator January 29, 2013 at 2:21 pm

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 [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.

Victor Dricks

HelpAllHurtNeverBaba January 29, 2013 at 8:34 pm Your comment is awaiting moderation.

Mr. Dricks, Respectfully, Along with Arnie Gundersen and John Large, I totally disagree with the NRC assessments on SONGS 10 CFR 50.59 RSG Evaluations. I was qualified SONGS 50.59 Screener/Evaluator for a decade besides being qualified at several other nuclear power plants. I have performed numerous 50.59 changes and reviews at SONGS. The changes shown below were claimed by Edison to be in the conservative direction and improvements.

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 and then you go back to your peers for more soul searching/research and provide more arguments and we will go from there:

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.

Change Number 2: Operating Secondary Pressure – OSGs: 900 psi, RSG: 833 psi ~ 10% change

Change Number 3: Tube wall thickness was reduced from 0.048 inches to 0.043 to pump more reactor coolant through the tubes > 10% change

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 did not either provided in-plane supports, or 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….

NOTE: 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.

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.”

Problems in SONGS Original CE Steam Generators: In the Original 2001 Power Uprate Application (NRC ADAMS Accession Number ML010950020), “Proposed Change Number NPF-10115-514 Increase in Reactor Power to 3438 MWt San Onofre Nuclear Generating Station Units 2 and 3”, SCE stated “ By the above reference Southern California Edison (SCE) submitted Amendment Application Numbers 207 and 192 to the facility operating licenses for the San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, respectively, to increase the licensed reactor thermal power level to 3438 MWt. At 100% power operation, steam generator pressures typically vary between 800 psia and 815 psia, compared to the original nominal design operating pressure of 900 psia. Wear at tube support structures is a known degradation mechanism at SONGS. At SONGS, rapid wear was observed on tubes surrounding the stay cylinder in the center of the steam generator during the first cycle of operation. Many tubes in the most susceptible region around the stay cylinder have been preventively plugged. The first preventive plugging was done after 0.7 EFPY of operation. The preventively plugged region was expanded during the Cycle 3 outage. Typical active wear in CE designed steam generators has occurred at the support structures in the upper bundle region of the steam generator. These supports consist of diagonal straps (frequently called bat wings) and vertical strap supports. This currently active wear mechanism is influenced by both flow velocities and tube to support gap wear. The variable influenced by the proposed uprate is the inner bundle flow velocities. The hydrodynamic stability of a steam generator is characterized by the damping factor. A negative value of this parameter indicates a stable unit, i.e., small perturbations of steam pressure or circulation ratio will diminish rather than grow in amplitude. The damping factors remain highly negative, at a level comparable to the current design, for all cases. Thus, the steam generators remain hydrodynamically stable for all uprate cases. Based on a projected increase of 2.3% in the secondary side fluid velocity, normal operation flow induced vibration analysis is impacted by the velocity increase. Current analysis considered that tubes with more than one consecutive inactive eggcrate were staked and plugged, and two nonconsecutive inactive eggcrates are acceptable. The Stability Ratio (SR) is defined as: SR = Veff/Vcr, where, Veff= effective velocity, Vcr = critical velocity; and Values of SR < 1 (NO FEI) are considered acceptable. Accordingly, wear growth rates will be managed by existing steam generator programs. The proposed power uprate has no direct effect on steam generator tube integrity.”

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.”

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.”

So I conclude 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.

Palo Verde made similar changes to their RSGs under a 50.90 License Amendment. PVNGS Generators are running after 10 years with very little tube plugging whereas the above changes in SONGS RSGs destroyed Unit 3 and crippled Unit 2 RSGs. Because of these adverse design changes, everybody is on the run: NRC Region IV, SCE, Mitsubishi, California Public Utilities Commission, Senator Barbara Boxer and Senator Dianne Feinstein. NRC Region IV, Westinghouse, AREVA, MHI, World’s Experts, SCE (Except DAB Safety Team SONGS Anonymous Insiders) are not sure whether fluid elastic instability in Unit 2 occurred or not. Southern Californians Ratepayers have lost $1 Billion in this game without electricity and now are faced with the trauma of restart of defectively-designed and degraded Unit 2 due to SCE’s continued mistakes. I am just trying to help, so please, wake up NRC Region IV and San Onofre Special Panel, Your charter is public safety and not whether SCE looses or makes money. I guarantee that SCE will make more money by admitting their mistakes and win NRC/Public Confidence by correcting their mistakes and using “Critical Questioning & Investigative Attitude” in the future. Remember, Mr. Dricks, Truth always prevails….. HAHN BABA

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