NRC: The Good, The Bad and the Ugly

NRC: The Good, The Bad and the Ugly

... and why it is unsafe To restart San Onofre

A GOOD NRC enforcement example:

Davis-Besse Nuclear Power Station is a nuclear power plant in Oak Harbor, Ohio. On March 5, 2002, maintenance workers discovered that corrosion had eaten a football-sized hole into the reactor vessel head of the Davis-Besse plant. Corrosion had been clogging the plant’s filters for months, requiring repeated filter replacement, but the cause was not investigated until after a worker leaned against a control rod drive mechanism, and it toppled over. Although the corrosion did not lead to an accident, this was considered to be a serious nuclear safety incident. Some observers have criticized the NRC’s Commission work as an example of regulatory capture [See Note 1] and the NRC has been accused of doing an inadequate job by the Union of Concerned Scientists.  The Nuclear Regulatory Commission kept Davis-Besse shut down until March 2004, so that FirstEnergy was able to perform all the necessary maintenance for safe operations. The NRC imposed its largest fine ever—more than $5 million—against FirstEnergy for the actions that led to the corrosion. The company paid an additional $28 million in fines under a settlement with the U.S. Department of Justice. The NRC closely monitored FENOC’s response and concluded in September 2009 that FENOC met the conditions of the 2004 order. From 2004 through 2009 the NRC reviewed 20 independent assessments conducted at the plant and verified the independent assessors’ credentials. The agency also conducted its own inspections and reviewed FENOC’s reactor vessel inspections conducted in early 2005. NRC inspectors paid particular attention to the order’s focus on safety culture and safety conscious work environment to ensure there were no new signs of weakness. The NRC task force concluded that the corrosion, occurred for several reasons:

·    NRC, Davis-Besse and the nuclear industry failed to adequately review, assess, and follow up on relevant operating experience at other nuclear power plants;

·    Davis-Besse failed to ensure that plant safety issues received appropriate attention; and

·    NRC failed to integrate available information in assessing Davis-Besse’s safety performance.

A BAD NRC enforcement example:

At San Onofre by Region IV and the NRC: The papers shown below have been obtained from Public Domain written by Dr. Joram Hopenfeld and a former SONGS Employee based on his investigation of the steam generator issues, review of the plant data and discussions with several Key SONGS Insiders. These papers confirm that Southern California Edison wants to restart unsafe Unit 2 nuclear reactor at 70% power under false pretenses, just for profits, and as an unapproved risky experiment by subverting the NRC and Federal regulatory process.  The true Root Cause (See Note 2) of the unprecedented tube-to-tube wear in Unit 3 has NOT been officially established, as required by NRC Confirmatory Letter Action 1 for restarting the defectively designed and degraded Unit 2.  NRC has not even completed their review of Unit 2 Return to Service Reports, nor have they finished Special Unit 2 Tube Inspections, nor have they (publicly?) reviewed SCE’s Response to NRC’s Requests for Additional Information (RAIs).
NOTE: NO FINES ARE MENTIONED - WHY?

An UGLY NRC enforcement example?:

Now, SCE wants the NRC to approve a new shady License Amendment, undermining public safety and they want it done without the involvement of Public Safety Experts, Attorneys and/or Citizens/Ratepayers.  After the review of the Mitsubishi Root Cause Evaluation and the Draft SCE License Amendment, it is crystal clear that the NRC needs to follow the example of their own enforcement at David Besse together with the lessons learned from Fukushima, when it comes to NOT approving this new Shady License Amendment for restarting San Onofre Unit 2’s defectively designed and degraded replacement steam generators.  In light of the unanticipated/unprecedented tube leakage at San Onofre Unit 3, the health and safety, along with the economic concerns/objections of 8.4 million Southern Californians’ MUST OVERRIDE and PREVENT the restarting of Unit 2 at ANY power level.  In a Democratic Society, truth must prevail over profit motivations, misleading propaganda of electricity service disruption and/or projected probabilistic temporary inconveniences to the public based on phony data, because America cannot afford a trillion dollar nuclear eco-disaster!

Our Safety must override SCE's profits and prevent them from restarting Unit 2.

Notes:

1: Regulatory capture occurs when a regulatory agency, created to act in the public interest, instead advances the commercial or special concerns of interest groups that dominate the industry or sector it is charged with regulating.  Regulatory capture is a form of government failure, as it can act as an encouragement for firms to produce negative externalities. The agencies are called "captured agencies".

2. Human performance errors resulting from the negative safety culture of production (profits) goals overriding public safety obligations.


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Additional Information:

The full DAB Safety Team's Media Alert 5 Parts:

https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?pli=1&docId=15V8BD4YK0MjwUV6gPZt6ILS_lP7CpClzgnZentLfx8U

The complete five (5) part presentation, see the eight (8) titles listed below:

Part 1, First Part  https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?pli=1&docId=1GcDuO0NvUdBBRVqCinM3Gr9bhmsJDCIbOg65rwdtjBY

Part 1, Second Part https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?pli=1&docId=1v_UcRLUlwmhtVDbftGcROPkF-cujiGa_-qjUJ7i2WIM

Part 1, Third Part https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?pli=1&docId=1yCOQ_diw8Qix1qG3j7MC6F2VfcqMoFOidxjJdIBpba0

Part 1, Fourth Part https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?pli=1

Part 2 https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?
pli=1&docId=1wkyTGWUy0pIXjBXc91JWr6_ACB3d-AIQbuVVZtDRZvI

Part 3 https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?pli=1&docId=16VmGtq9oYGS2-bpR4nBt9d-YB21E2xrjFpIJoxFJ4s4

Part 4 https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?pli=1&docId=0BweZ3c0aFXcFMC1BNlpMeDQyM2M

Part 5 https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?pli=1&docId=0BweZ3c0aFXcFX2gxRDFCRW9CVzg

San Diego screening of MOVIE: "311: Surviving Japan"

WE HAVE THIS weekend to pre-sell just 6 seats to make this screening happen in San Diego! No excuses on the weekend, just DO it: You will NOT be charged until the event is confirmed. ($12 per ticket)http://www.tugg.com/titles/311-surviving-japan?location=global&state=upcoming

Please join us for the 2nd Anniversary of the Fukushima nuclear disaaster!

San Diego screening of "311: Surviving Japan" on Monday, March 11, 2013, the 2nd Anniversary of the Fukushima nuclear disaster (7:30p). PLEASE RESERVE YOUR TICKET(S) ONLINE ASAP TO ENSURE THE FILM IS SCREENED HERE: http://www.tugg.com/titles/311-surviving-japan?location=global&state=upcoming You will NOT be charged until the event is confirmed. ($12 per ticket).  We must sell 50 tickets in advance ASAP for this film to be a go at the theater!

Where:

AMC Fashion Valley 18

7037 Friars Rd, San Diego, California 92108

YOU MUST RESERVE YOUR SEAT ONLINE NOW!  There will be no "walk up" tickets sold.

We will do a "Light Brigade" action along the busy Friars Road entrance to the Fashion Valley Mall from 6-7p, holding lighted letter signs spelling out "No More Fukushimas". Please emailmarthasullivan@mac.com to volunteer as a "Holder of the Light."

We will also do Flyering inside the Mall during the same period before the screening: 6-7p, with a special guest for the occasion. Please email marthasullivan@mac.com to volunteer for the Info Crew.

Here is a brief description of the film: "Inside story of 2011 Japanese Tsunami relief & Fukushima nuclear disaster. A critical look at how the authorities handled the nuclear crisis and Tsunami relief by an American who volunteered in the clean-up. It is in short, a documentary of the devastating events in Japan and 6 months of the after-math that followed. It features true stories from those affected by the disaster, the government and even TEPCO. It highlights the struggle in dealing with: The Tsunami clean-up, Government response to the disaster, radiation plus the future of nuclear power after the accident." (90 minutes long, plus speaker.)

NRC AIT Review Of San Onofre Requires An NRR Investigation

NRC AIT review of SCE 10CFR 50.59: The NRC AIT stated in its report, “Based on the updated final safety analysis report description of the original steam generators, the team determined that the steam generators major design changes were reviewed in accordance with the 10 CFR 50.59 requirements.  The team determined that no significant differences existed in the design requirements of Unit 2 and Unit 3 replacement steam generators.  Based on the updated final safety analysis report description of the original steam generators, the team determined that the steam generators major design changes were reviewed in accordance with the 10 CFR 50.59 requirements.  

Later the NRR technical specialists reviewed SCE’s 10 CFR 50.59 evaluation against 10 CFR 50.59(c)(2)(viii) which requires that licensees obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change if the change would result in a departure from a method of evaluation described in the final safety analysis report (as updated) used in establishing the design bases or in the safety analyses.

The NRR technical specialists 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.” (emphasis added)  The changes were as follows:

• Reactor Coolant System Structural Integrity - Use of ABAQUS Computer Program 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.

• Main Steam Line Break Mass-Energy Blowdown Analysis & Tube Wall Thinning Analysis – Use of ANSYS Computer Program instead of STRUDL and ANSYS Computer Programs:  SCE’s 50.59 evaluation did not mention whether the method has been approved by NRC for this application. 

The NRR now needs to investigate why the NRC AIT Team displayed poor judgment in their review of SCE’s 10 CFR 50.59 Evaluation, which in effect, let SEC off the hook without even a fine, for making design changes that put all of Southern California at risk, since we came so very close to having a Fukushima-type radioactive nuclear accident in San Onofre less than a year ago!

Link to full Press Release 13-01-07 NRC AIT Review Requires An NRR Investigation

The DAB Safety Team: January 7, 2013    Supplemental To Our Press Release + 12-12-31

Copyright January 1, 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

Edison’s Claims About SONGS Unit 2 Pressures Are Erroneous

Press Release

The DAB Safety Team: December 7, 2012

Media Contact: Don Leichtling (619) 296-9928 or Ace Hoffman (760) 720-7261 

SCE’s Claims About SONGS Unit 2 Steam Generator Operating Pressures Are Erroneous Because They Conflict With SCE’s Submitted NRC Reports And SCE’s Plant Procedures (Operational Data).

Now SCE is claiming in their Unit 2 Restart documents, “Limiting power to 70% significantly reduces fluid velocity. The reduction in fluid velocity significantly reduces the potential for FEI.”  What they are not saying is that reducing power to 70% significantly increases the steam generator operating pressures, (as the NRC said in its AIT Report) which will:

·      Increase the pressure inside all the already damaged SG tubes

·      Do nothing to completely eliminate FEI from happening at any time during normal plant operations, and especially during a MSLB or similar accident, which can cause a nuclear incident or worse!

SCE’s attempt in using evasive and misleading technical inconsistencies to justifying their dangerous and possibly catastrophic restart plan cannot hide the truth, revealed in their actual plant operational data provided to the NRC and published in the NRC AIT Report.

Background History:

After the radioactive leak occurred in the San Onofre Unit 3 steam generator, Arnie Gundersen along with a team of anonymous steam generator experts were the first ones in the industry to absolutely state, “The pitch to diameter ratio of tubes in the original CE generators is dramatically different from any of the Westinghouse generators fabricated by Mitsubishi.  As water moves vertically up in a steam generator, the water content reduces as more steam is created.  With the Mitsubishi design the top of the U-tubes are almost dry in some regions. Without liquid in the mixture, there is no damping against vibration, and therefore a severe fluid-elastic instability developed.  The real problem in the replacement steam generators at San Onofre is that too much steam and too little water is causing the tubes to vibrate violently in the U-bend region. The tubes are quickly wearing themselves thin enough to completely fail pressure tests. Even if the new tubes are actively not leaking or have not ruptured, the tubes in the Mitsubishi fabrication are at risk of bursting in a main steam line accident scenario and spewing radiation into the air.”

SCE’s Restart Plan Justification Is Just Scientific Misinformation:

Based on analysis of the NRC AIT Report, Westinghouse’s Operational Assessment, SONGS procedures, operational data, plant daily briefing sheets and engineering calculations the DAB Safety Team concludes the following:

·      Secondary side lower pressures (833 psi) along with higher reactor thermal power and design deficiencies (low tube clearances) at 100% power created conditions of “ALMOST NO WATER” in certain regions of both Unit 3 steam generators tube bundles.  This resulted in fluid elastic instability, where unprecedented tube-tube wear was observed.  At the June 18, 2012 AIT presentation, the NRC said, “Throughout the US nuclear industry, this is the first time more than one steam generator tube failed pressure testing…. Eight tubes failed. The pressure testing identified that the strength of eight tubes was not adequate and structural integrity might not be maintained during an accident… this is a serious safety issue.”  Southern Californians were lucky, that SONGS Unit 3 tube leakage was detected and stopped in time.  Otherwise, this condition could have potentially caused a reactor meltdown like Fukushima in Southern Californian’s backyards. 

·      Secondary side higher pressures in Unit 2 (864-942 psi) at 100% power negated the effects of  “low tube clearances” and prevented steam “dry-out” (high void fractions) in the Unit 2 tube bundle region, where no fluid elastic instability (tube-tube wear) was observed. 

The DAB Safety Team’s findings are summarized as follows:

·      DAB Safety Team “Strongly Agrees” with Arnie Gundersen and his team of anonymous steam generator experts and with MHI on the causes of fluid elastic instability in Unit 3.  What did SCE do, instead of thanking Arnie Gundersen, who first identified the real cause of the problem, tried to discredit him by implying, “What does he know about steam generators, he is just a high school math teacher.”

·      DAB Safety Team “Agrees” with Westinghouse, why fluid elastic instability did not occur in Unit 2.

·      DAB Safety Team “Strongly Disagrees” with both SCE’s conclusions “that fluid elastic instability Most Likely Occurred in Unit 2” and “secondary side operating parameters were similar in the U3 and U2 SGs”. 

·      DAB Safety Team “Strongly Disagrees” with NRC that the differences in the actual operation between units and/or individual steam generators had an insignificant impact on the results and in fact, the NRC AIT team did not identify any changes in steam velocities or void fractions that could account for the differences in tube wear between the units or steam generators.  Discussions with two of the NRC panel members gives us the perception that the NRC panel members disagree amongst themselves and also with SCE on the effect of operational parameters on fluid elastic instability in Unit 2 Steam Generator E-089.

Adverse operational conditions, such as larger reactor thermal power and lower steam generator pressures (e.g., 833 psia) and design deficiencies (low tube clearances and no-in-plane fluid elastic instability structural protection) cause areas in the U-tube bundle of a nuclear steam generator to have “ALMOST NO WATER” as observed in SONGS Unit 3 steam generators.  When this happens, fluid elastic instability occurs and the thin steam generator tubes carrying radioactive coolant move with large sprinting amplitudes and hit the neighboring tubes with violent and repeated impacts.  Therefore, multiple tube failures can occur, as was observed in SONGS Unit 3 at main steam line break testing conditions.

MHI states, “The higher than typical void fraction is a result of a very large and tightly packed tube bundle, particularly in the U-bend, with high heat flux in the hot leg side. This high void fraction is a potentially major cause of the tube FEI, and consequently unexpected tube-to-tube wear (as it affects both the flow velocity and the damping factors). In general, larger thermal power is more severe for vibration, because the steam flow rate increases. At constant thermal power, lower steam pressure is more severe for vibration than higher pressure.” MHI is indirectly saying that steam generator pressures of 833 psia created fluid elastic instability in Unit 3, where unprecedented tube-to-tube wear was observed.  AREVA states, “At 100% power, the thermal-hydraulic conditions in the U-bend region of the SONGS replacement steam generators exceeded the past successful operational envelope for U-bend nuclear steam generators based on presently available data.” MHI has officially notified the NRC that all SONGS damaged RSG Tubes subject to tube-to-tube wear (FEI) should be plugged and or stabilized.  SCE cannot certify this as having been done, since they have not inspected the majority of Unit 2’s RSG tubes using the most advanced technology, as indicated in HMI’s official notice to the NRC.  Again SCE is caught guessing about the amount of tube fatigue damage, which directly affects the RSG tube structural integrity; all RSG tubes are subject to tube-to-tube wear, extreme pressure variations and other stresses during a MSLB or other unanticipated operational transients.

NRC AIT Report states, “The team performed a number of different thermal-hydraulic analysis of Units 2 and 3 steam generators. The output of the various analyses runs were then compared and reviewed to determine if those differences could have contributed to the significant change in steam generator tube wear. It was noted that Unit 3 ran with slightly higher primary temperatures, about 4°F higher than Unit 2. The result of the independent NRC thermal-hydraulic analysis indicated that differences in the actual operation between units and/or individual steam generators had an insignificant impact on the results and in fact, the team did not identify any changes in steam velocities or void fractions that could attribute to the differences in tube wear between the units or steam generators. It should be noted that increases in primary temperature and steam generator pressures has the effect of reducing void fractions and peak steam velocities, which slightly decreases the conditions necessary for fluid elastic instability and fluid-induced vibration. The analysis included the varying of steam generator pressures from 833 to 942 psia.”

SCE says in their Root Cause Analysis, “Secondary side operating parameters were similar in the U3 and U2 SGs and well within their design limits (e.g., steam generator pressures, 833 psia).”  Note, NO mention varying the pressure to 942 psia at all…

Copyright December 7, 2012 by The DAB Safety Team. All rights reserved. This material may not be published, broadcast or redistributed without crediting the DAB Safety Team. The contents cannot be altered without the Written Permission of the DAB Safety Team Leader and/or the DAB Safety Team’s Attorneys.

San Onofre: Too Close To Having Two Nuclear Accidents

Decommission San Onofre

SONGS Insider Secret: Southern California came close to having a level 3 nuclear incident (or worse, a nuclear disaster), at both SONGS Unit 2 and Unit 3.

A radioactive disaster was narrowly averted because of the off chance discovery that one of Unit 2 SG tubes had experienced a loss of 90 % percent of its wall thickness!  This was discovered before the tube had a chance to fail (like the eight tubes that failed in-situ testing in Unit 3) because Unit 2 happened to be shut down for a scheduled refueling outage, so its SG tubing were also inspected after the radioactive SG tubing leakage problems were discovered in Unit 3. Note: These SG tubes have a wall thickness (wear) plugging limit of 35%, so this single tube remained in use long after its safety limitation had been exceeded by almost double (55%)!   A single tube failure does not sound important, unless you realize it contains highly radioactive reactor core coolant that is under both very high pressure and temperature; then any leak is very bad (think massive radioactive atmospheric releases). A single tube rupture can also lead to a cascade of tube ruptures.

A Perfect Example of highly radioactive reactor core coolant: in 1991, leakage of about 55 tons of primary coolant occurred due to the failure of one SG tube in a steam generator built by Mitsubishi in the No. 2 pressurized water reactor at the Mihama nuclear power station in Japan.  At the same time, water pressure in the core had dropped drastically and the ECCS kicked in, flooding the reactor and shutting it down.  If the core had been left exposed, a meltdown -- an overheating of the fuel that can, if uncontrolled, lead to a large release of radioactivity -- could have occurred. In the following week an estimated 7 million Bq was released into the sea and an estimated 5 billion Bq of radioactive gas was released into the atmosphere.  This tube rupture caused the first International Nuclear Events Scale (INES) level 3 nuclear incident in Japan and ignited social concerns all over Japan because it shattered the nuclear industry’s myth of 100% safe reactors! SONGS MSLB  Analysis + 12-11-19, page 14

To date, there have been too many guesses, too many probabilities built around gamed safety margins and too many already damaged tubes (thousands of which have not been visually inspected) that any one or combination of, could cause a major nuclear accident even without something like a Main Steam Line Break (MSLB) or a big quake!

Given the documented UNSAFE condition of Unit 2, there is no excuse for SCE to even consider seeking a restart of Unit 2 except greed.  The NRC needs to make that crystal clear to SCE, otherwise they, like SCE, have failed the public’s trust! – The DAB Safety Team

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Nuclear Safety Is No Accident

The DAB Safety Team along with the support of an ever-growing number of SONGS Concerned Insiders and Whistleblowers have prepared this analysis, which is consistent with the conclusions presented in the publicly available reports provided earlier on this subject by:

·       Internationally Known Nuclear Consultant Arnie Gundersen and his team of Anonymous Industry insiders, who had lengthy careers in the design, fabrication, and operation of nuclear steam generators.

·       Professor Daniel Hirsch and Internationally Known Nuclear Consultant Dale Bridenbaugh.

·       Publicly available posted documentation by Dr. Joram Hopenfeld, a retired engineer from the Office of Nuclear Regulatory Research and NRC's Advisory Committee, and

·       David A. Lochbaum, Director of the Nuclear Safety Project for the Union of Concerned Scientists (UCS).

Background

Quote No 1: SCE is ultimately responsible for the work done by their vendors and contractors.  NRC Chairman Allison Macfarlane

Quote No 2: Both San Onofre units will remain shut down until repairs are made and we and the Nuclear Regulatory Commission are satisfied it is safe to operate. Pete Dietrich

Quote No 3:  Running the reactor at a 30 percent reduction in power may not fix the problems but rather make them worse or shift the damage to another part of the generators. It’s a real gamble to restart either unit without undertaking repairs or replacing the damaged equipment. Arnie Gundersen

Quote No 4: AREVA States, "The nominal distance between extrados and intrados locations of neighboring U-bends in the same plane ranges from 0.25 inches to 0.325 inches due to the tube indexing. There are 36 U-bends in Unit 2 SG E-088 and 34 in SG E-089 with a separation less than or equal to 0.050 inches. The U-bends with the smaller separation distances are much better candidates for wear by rubbing yet do not exhibit TTW.”

Based on SONGS Unit 3 experience, this behavior can change because of plugged/staked tubes due to shifting of localized steam-dry out regions in the hot leg U-Tube Bundle region of these U-bends with these extremely low clearances.  These extremely low clearances can cause very high steam velocities, which can then result in fluid elastic instability, EVEN during the proposed 150-day monitoring period EVEN at reduced power levels and EVEN  at higher steam pressures.

The DAB Safety Team Solve the Big Mystery Behind The Destruction of SONGS Unit 3 Replacement Steam Generators (RSG’s) And The Limited Damage to Unit 2 RSG’s:

Start With Some Basic Thermodynamics: At lower steam pressures (~833 psi), steam-water mixture has a saturation temperature of 523⁰F and more internal energy aka HEAT (1198.2 Btu/lb.) as compared with higher steam pressure (~942 psi), steam-water mixture has saturation temperature of 537.5⁰F and less internal energy aka HEAT (1194.7 Btu/lb.).  Therefore, you can generate more Megawatts out of the steam generator running it at lower steam pressure, if you supply more reactor thermal power (HEAT) to the steam generator tubes from the reactor.  Lower steam pressures combined with high reactor thermal power (HEAT), high steam velocities and narrow tube clearances also promotes bad things, like localized steam dry-out regions (vapor fraction >99%) in the hot leg side of the U-Tube bundle, which can cause fluid elastic instability, flow-induced random and severe vibrations, and excessive hydrodynamic pressures (aka Mitsubishi Flowering Effect).  These adverse effects can, in turn cause excessive tube-to-tube fretting wear leading to cascading tube leakages and or ruptures, increased tube support clearances and/or tube-to-AVB gaps, and even deformation of the anti-vibration bar structure.  Tube-tube/AVB wear and deformation (aka Mitsubishi Flowering Effect) of the floating anti-vibration bar structure without structural beams and lateral/mid-span supports causes redistribution of tube-to-tube/AVB gaps and clearances, when the steam generator reverts back from hot operating to cold conditions.  This redistribution of gaps can lead to differences in measurements during steam generator refueling inspections compared with the original manufacturing/design gaps in the cold condition and can lead to wrong projections and misleading conclusions regarding past and future cycles operating conditions.

Add Some Operational History: The original SONGS CE Model 3410 steam generators (OSG’s) were rated for 1705 MWt reactor thermal power.  Over the years of operation of the SONGS OSG’s, it became evident that the steam generator tubes, made predominantly of Alloy 600, were susceptible to primary water stress corrosion cracking (PWSSC). This corrosion mechanism resulted in tube degradation necessitating plugging large numbers of tubes after each tube inspection. In addition, the SONGS OSG’s design had shown to be susceptible to tube through-wall wear and severe corrosion of the tube supports. Continuing to operate with these highly degraded steam generators involved substantial economic risks from forced outages, extended refueling outages, as well as the direct costs of inspections and repairs.  SCE’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 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.  Mitsubishi selected a u-bend configuration for the upper part of the tube bundle instead of the square bend design of the original steam generators based on its experience that u-bends were easier to fabricate and support and were easier to inspect.

The original steam generators installed throughout the domestic fleet of pressurized water  reactors, including SONGS, experienced widespread corrosion of the tubes and tube support plates, stress corrosion cracking of the tubes, and wear at tube supports.  These problems led to the replacement of nearly all of the original steam generators, in most cases well before the end of their design lifetime.  For SONGS, the design of the replacement steam generators included a number of design changes to correct life limiting problems with the original steam generators, based in part on consideration of SONGS-specific and industry-wide operating experience. This included use of more corrosion resistant materials for the tubing and tube support plates to mitigate corrosion. The tubes in the new Replacement Steam Generators (RSGs) were fabricated with thermally treated Alloy 690, which has superior corrosion resistance.  The only drawback the new Alloy 690 has, is that it has a 10% less thermal conductivity/heat transfer rate compared with old tube alloy 600. 

Therefore, to achieve the thermal output of 1729 MWt from the new RSGs as approved by the NRC in 2001 SONGS Power Uprate Application, SCE engineers needed to install 11% more tubes in the new RSGs.  By removal of the stay cylinder from the OSGs, there was enough space to add only 4% (377) more tubes. Since there was no room to add the additional 7% tubes, SCE increased the length of each tube in the U-Tube Bundle (Outside the Industry NORM and is partially responsible for Mitsubishi Flowering effect) by more than 7 inches to obtain 1729 MWt out of the RSGs.  The Unit 2 RSGs were built with bigger tube-to-AVB Gaps and no in-plane protection from vibrations caused by potential fluid elastic instability conditions, because the OSGs did not experience that phenomena.

Based on information from anonymous SONGS operations personal, the root cause team of concerned insiders, a cursory review of plant records (plant procedures, system descriptions and plant daily briefing sheets), engineering calculations performed using SONGS procedures and review of NRC AIT Report, SCE Operators were running Unit 2 RSGs at higher steam pressures (~863-942 psi) and lower reactor thermal power (HEAT 1715-1725 MWt) for almost 22 months with no reported and detected abnormality (6 tubes with 28 to 90 percent wear of the tube wall thickness due to retainer bar vibrations).  Because of the short measured lengths of these flaws, only the 90 percent indication was in-situ pressure tested as part of condition monitoring. The affected tube was successfully pressurized to 5300 psi with no leakage (Southern Californians were saved from another nuclear accident). 

The Unit 3 SG manufacturing process used more accurate and tighter tolerances, which improved tube-to-AVB alignment such that tubes had more contact forces with AVB's and provided an effective “zero” tube-to-AVB gap under operating (hot) conditions. According to trusted SONGS operation insiders, SCE Engineers were convinced that Unit 3 RSGs were built better than Unit 2 RSGs in terms of providing in-plane protection due to better control of tube-to-AVB gap. Therefore, SCE Engineers decide to operate Unit 3 RSGs at lower steam pressures (~833 psi) to get more Megawatts out of the RSGs by supplying more reactor thermal power (HEAT ~1729 -1739 MWt) to the RSG’s tubes from the Unit 3 reactor.  This was unfortunately their biggest miscalculation.  What happened next is well known, because of the low steam pressures (~833 psi), combined with low tube clearances (SONGS design:  0.250 inches; some found as low as 0.050 inches) and no effective in-plane tube support protection, the high reactor power (HEAT) resulted in fluid elastic instability, flow-induced random vibrations, excessive hydrodynamic pressures and localized steam dry-out regions (vapor fraction >99.6%) which resulted in the destruction of SONGS Unit 3 RSGs (1 tube leak, 8 tubes failures at MSLB conditions, 1800 tubes with tube-to-tube /anti-vibrations bars/support plates wear – this amount of damage is unprecedented in the history of the US Operating Nuclear “fleet”).   Fluid elastic instability and Mitsubishi Flowering Effects increased the tube-to-AVB gaps, decreased tube-to-tube clearances and created lower and insufficient contact forces in Unit 3, which were described inaccurately by the NRC and SCE as a MHI manufacturing defect , presumably due to political and/or financial reasons. 

After 22 months of operation the severity of wear in Unit 2 was determined to be similar to that experienced by Unit 3 after 11 months of operation.  The steam generator in Unit 2 had about 2600 wear indications at AVB’s compared to about 3400 wear indications in Unit 3. Without an effective in-plane support system, these high fluid velocities and localized steam dry-out regions resulted in very large or uncontrolled vibrations (amplitudes) of tubes in the in-plane direction and caused fluid elastic instability, flow-induced random vibrations, excessive hydrodynamic pressures and increased tube-to-AVB gaps and created insufficient contact forces in Unit 3.  

Analysis Solves The Mystery: High steam pressures (~863-933 psi) and lower reactor thermal power (HEAT 1715-1725 MWt), coupled with low tube clearances and no in-plane tube support design accompanied with high steam velocities caused flow-induced random vibrations and excessive hydrodynamic pressures, which resulted in SONGS Unit 2 RSGs damage (high tube-to-anti-vibrations bars/support plates wear).  Since the steam pressures were higher, the void fractions were less than 98.5% and no fluid elastic instability occurred in Unit 2, which is consistent with the Westinghouse finding.  The NRC, AREVA, Westinghouse, MHI and SCE all missed this key operational observation in the NRC AIT Report, SCE Unit 3 Root Cause Evaluation and SCE Unit 2 restart Plan.  MHI indirectly alluded to this fact in their technical report but did not say it publicly for reasons unknown to the DAB Safety Team -- perhaps they were afraid of backlash from SCE and NRC.  At least one person working at SONGS discussed this fact about operational differences with other personnel between Units 2 and 3, but nobody listened to him.  His findings were intentionally or unintentionally ignored because everybody inside SCE was focused on blaming MHI in order to recover the insurance money and/or absolving themselves of all blame.  The DAB Safety Team is sure that MHI will pursue this fact during arbitration proceeding to absolve them of this blame and protect their reputation as a builder of high quality RSG’s.

Public Expectations: The public expects that the NRC complies with President Barack Obama, Senator Barbara Boxer and the NRC Chairman’s Open Government Initiative by using the Reactor Oversight Process, when it audits SCE’s Licensing Basis Documents, facility procedures/records, 10 CFR 50.59 Safety Evaluations, Unit 2 Restart Documents and issues/approves Safety Evaluation, License Amendment Applications and Inspection Reports, Responses to Confirmatory Action Letters and other enforcement violations, as appropriate.   The NRC must complete its mission of ensuring public safety with transparency and public involvement by issuing all documents, emails, telephone records along with holding open and trial-like public hearings without any time limitations from SCE, its vendors and/or contractors.

Questions which must be answered by NRC and SCE prior to any Unit 2 restart: The questions which NRC in its 95 page AIT Report the SCE and its vendors in their 1200 page San Onofre Nuclear Generating Station Unit 2 Return to Service Report have not answered completely, convincingly and unanimously are as follows:

(1)  Whether fluid elastic instability occurred, occurred for a limited time or did not occur at all in the SONGS Unit 2 Replacement Steam Generators (RSG’s)?

(2)  What was the contribution of fluid elastic instability and “Mitsubishi flowering effect” in increasing the Unit 3 Tube-to-AVB gaps, which were built better than Unit 2 and designed to provide an effective “zero” tube-to-AVB gap under operating (hot) conditions?

(3)  How were the SONGS RSG’s specified, designed and fabricated as comparable replacements on a like-for-like basis for the original steam generators in terms of fit, form and function with such numerous untested and unanalyzed design changes under the 10CFR50.59 rule?

(4)  Were these numerous untested and unanalyzed design changes responsible for damage to SONGS RSG’s?

(5)  Were SONGS RSG’s operating beyond their design basis or Industry NORMs and if so, how did that impact the degradation of RSGs?

 

(6)  Was this degradation the fault of SCE’s in-house design team, their Performance Specifications coupled with their numerous design changes and/or the MHI Fabrication/Testing Technology combined with Faulty Thermal-Hydraulic Computer Codes, which caused the unprecedented damage to SONGS RSG’s? 

(7)  Are the Unit 2 RSG’s qualified in the “As-designed and Degraded Condition” for a MSLB or other Anticipated Operational transients?

The DAB Safety Team and the Public expects that SCE and their three NEI Qualified, “US Nuclear Power Plant Designers”, Westinghouse, AREVA and MHI will revise their reports and arrive at concise and clear answers (meeting the NRC Quality Assurance requirements as stated by NRC Chairman Allison Macfarlane) to the puzzling public safety questions in the Unit 2 Return to Service Report.

SONGS Poor RSG’s Design: Due to unsubstantiated claims, complacency, challenges and rewards of innovative steam generators, time and financial pressures, both inexperienced Southern California Edison (SCE) and Mitsubishi Heavy Industries (MHI) Engineers did a very poor job of Industry Benchmarking and keeping up with the Academic Research Papers on the basic lessons of steam generator design (how to prevent the adverse effects of fluid elastic instability, flow-induced random vibrations, excessive hydrodynamic pressures and preventing localized steam dry-out regions observed in SONGS Units 2 and 3 RSG’s with high steam flows to produce more megawatts than the original steam generators (1729 MWt vs. 1705 Mwt)).  In addition, SCE made numerous untested and unanalyzed design changes under the pretense of 10 CFR 50.59 process (like for like change) and intentionally (Public Perception) avoided the NRC 50.90 License Amendment Process.  These factors contributed to the catastrophic failure of a 570 million dollars piece of equipment vital to the safety of Southern California.

SCE Restart Report for SONGS Unit 2: To address the tube leak and its causes, SCE assembled a technical team from MPR Associates, AREVA, Babcock & Wilcox Canada, PVNGS, EPRI, INPO, Westinghouse, Intertek APTECH and MHI, as well as experienced consultants including former NRC executives and a research scientist.  On October 3, 2012, SCE submitted a 1200 page report prepared by these experts (Under the closed Scrutiny, Guidance and Leadership of SCE Managers) to NRC Region IV. The SCE Root Cause Evaluation Report, Operational Assessments reports prepared by SCE, AREVA and Westinghouse, and MHI Technical reports conflict and contradict with each other on the causes and extent of degradation pertaining to the fluid elastic instability in SONGS Unit 2 Replacement Steam Generators (RSGs) and Tube-to-AVB gaps in both Unit 3 and Unit 2 RSG’s.  MHI truly states that specific causes that resulted in tubes being susceptible to fluid-elastic excitation are not yet completely known.  

PRESS RELEASE 

The DAB Safety Team: November 28, 2012

Media Contact: Don Leichtling (619) 296-9928 or Ace Hoffman (760) 720-7261

Copyright November 28, 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.