Yesterday the Commissioner and Administrative Law Judge in the California Public Utilities Commission's investigation ("OII") into San Onofre's steam generator problems released a "Scoping Memo" designed to define (and RESTRICT) the extent of the investigation into the problem. If the memo stands as the definition of scope, it will be strictly limited to little more than figuring out who will pay to rebuild and restart San Onofre. There is no provision for looking at the "big picture" of whether there is any sense in replacing or repairing the broken reactors.
This is a travesty and people need to demand the CPUC OII include a much broader look at the issues. I have taken the first section of the scoping memo ("background") and responded below.
(1) Comments on CPUC's OII "scoping memo"
(2) Quotes from the CPUC OII January 28, 2013 scoping memo
(3) A brief history of the current steam generator problem at San Onofre
(4) Press release regarding the scoping memo from Woman's Energy Matters
(5) Statement regarding the scoping memo from Michael J. Aguirre, Esq.
(6) One-year commemoration of the near-destruction of Southern California planned
(7) Contact information for the author of this newsletter
(1) Comments on CPUC's OII "scoping memo":
The scope of the California Public Utilities Commission's investigation into San Onofre should not be limited in any way. The fact is, that as of January 31, 2012, California has experienced far safer and more reliable electricity service than ever before. The mood of the citizens has also improved as they live with less FUD (Fear, Uncertainty, and Doubt). These advantages should be reflected in more just and reasonable rates for the ratepayers. The increase in safety and satisfaction is in precise response and proportion to the San Onofre Nuclear [Waste] Generating Station (SONWGS) being offline.
Specifically, the operations there over the past few decades, that have led to sudden and/or prolonged power outages, and a constant threat of meltdown, have ceased, and this is a good thing.
Specifically, the dangerous safety violations that have led to whistleblower revelations that have terrified the citizenry because of their safety significance (including retaliatory practices by management, cover-up of fraudulent work practices (such as skipping fire watch rounds), etc.) have been reduced in significance due to the involuntary shut-down of the reactor site.
Specifically, the underhanded practices of the operator, such as misleading local city officials in secret meetings after the citizens have spoken to those cities, and such as targeting any cities the citizens speak to, in a game of whack-a-mole rather than open dialog, honest debate, in-depth discussions, and any reasonable attempt to reach a mutual understand among all the parties.
And specifically, the immoral conduct of Southern California Edison's spokespeople to claim that the nuclear spent fuel waste problem is solvable by government and is not their problem. For approximately 30 years the public were assured by SCE's spokespersons that Yucca Mountain was the solution, so there was no need to discuss the "waste problem" at all and it was dismissed as a non-issue. Then it was the Private Fuel Storage facility in Utah, for a couple of years, that was going to solve the waste problem locally by putting it off on a small but well-paid Goshute Indian tribe. That didn't work out either, and now the public is being told the solution will be the Blue Ribbon Commission's "interim storage facility" whatever, whenever and wherever it is. But even that will be decades in the future if it ever happens at all.
Meanwhile, deadly DRY CASK STORAGE piles up on the coast, now totaling over 40 individual enormous deadly casks, with nearly a hundred more to go, just to get caught up with the current load of highly radioactive used nuclear reactors on site at the present time.
These practices are related part-and-parcel to the extended outages at SONWGS Units #2 and #3: The dangers from SCE's mistakes, cover-ups, lies, falsifications, misrepresentations and other forms of negligence subsides over time as the fuel cools both thermally and radiologically. This outage is GOOD for Southern California: It should not stop. The units should not be restarted.
The root cause of the outage is greed: SCE tried to design new steam generators that would generate the same OR MORE steam within the same size steam generators, using heat transfer tubes with 10% lower heat transfer capabilities. The advantage of the new alloy 690 supposedly was that it was more ductile, stronger, and longer-lasting than the old alloy 600. These all may well be true, when the steam generators are properly designed. Indeed, Edison claimed that a 60-year life was expected, although the CPUC only wanted them to show cost-effectiveness for 18 years (which was only possible by dint of SCE being allowed to exclude numerous cost factors from their calculations, including but not limited to such events as the current outage).
However, even assuming alloy 690 is "better" than alloy 600 in many ways, the disadvantage of a lower heat transfer coefficient apparently was not properly accounted for by SCE in their desire for maintaining or even increasing the steam productivity of the replacement units. SCE apparently forgot basic thermodynamic and hydraulics principles. Then they apparently forgot basic engineering principles of having independent eyes check their work, of benchmarking their assumptions against industry standards, and of reviewing the historic record for what is known about the subject (the subject being "Fluid Elastic Instability" which was first recognized as a potential problem in nuclear reactor steam generators in the 1970s). SCE worked hard, but only at hiding their efforts from the public view: They did this by calling all the dozens of significant changes to the steam generators "like for like" because the outer shells of the steam generators were to be the exact same dimensions. But inside those shells numerous vital changes were being made.
In short, SCE attempted to put a sports car engine in the family sedan, and ran into a wall when they first stepped on the accelerator. Luckily, damage was mainly to their pride, and to the ratepayer's wallets. A catastrophic cascade of tube ruptures was avoided, but only by the narrowest of margins, as proven by the 8 tube failures during subsequent pressure testing.
The proper scope of this proceeding is by its very nature massive: The lives of 8.5 millions Southern Californians are at stake. Billions of dollars in what could be green energy investments might be poured into the rat-hole that is San Onofre Nuclear Waste Generating Station if a poor decision is made in this proceeding.
(2) Quotes from the CPUC OII January 28, 2013 scoping memo:
Quotes from CPUC's OII Scoping Memo, 1/28/3013:
"On November 1, 2012, the Commission issued this Order Instituting Investigation (OII). The Commission will investigate the ongoing shutdown of nuclear generation at the San Onofre Nuclear Generating Station (SONGS), and the resulting effects on the provision of safe and reliable electric service at just and reasonable rates. Specifically, this investigation will consolidate and consider issues raised by the operations, practices, and conduct of Southern California Edison Company (SCE) and San Diego Gas & Electric Company (SDG&E) related to and following the extended outages of SONGS Units #2 and #3. The Commission will examine the causes of the outages, the utilities' responses, and the future operation of the SONGS units as part of a review of SCE's actions, and to assess what costs, if any, are appropriate for recovery from ratepayers."
From: "Scope of the Proceeding":
"This is an evolving OII and some key facts (e.g., third party cost recovery, future actions by Nuclear Regulatory Commission (NRC), and repair/replacement options for the steam generators) have yet to be determined."
Note that "repair/replace" should read "repair/replace/retire" but instead the "retire" option has been ignored.
"Nature and effects of the steam generator failures in order to assess the reasonableness of SCE's consequential actions and expenditures (e.g., was it reasonable to remove fuel from unit #3)."
Note that the example of questioning the logic of removing the fuel from Unit #3 implies, once again, a desire to restart even that most-broken of reactors!
"Issues related to the future operation of SONGS as a reliability source shall be considered in the Long-Term Procurement Planning (LTPP) proceeding, Rulemaking (R.) 12-03-014."
This is yet another way of removing as many issues as possible from the OII even though the OII is supposed to figure out if the SGRP was "cost-effective" or not. "Cost-effective," by definition, should consider ALL the costs!
(3) A brief history of the current steam generator problem at San Onofre:
The first reference I recall to a "cascade" of tube failures was by David Lochbaum, director of the nuclear safety project for Union of Concerned Scientists, possibly as early as February 2012.
Lochbaum formerly worked in the industry and is a trained nuclear engineer. His immediate response to the event was that such leaks are not unusual in newly-installed steam generators (Feb 1, 2012, voiceofoc.org) but after it was established that Fluid Elastic Instability was the likely culprit in unit three (a type of vibration which gets worse and worse once it starts, and tube failure becomes virtually inevitable), he was noting that "Serious leaks also can drain cooling water from a reactor." (KQED, March 16, 2012)
Within a few weeks or at most a couple of months, Arnie Gundersen also was making the claim of a potential cascade of tube failures. Gundersen was also a nuclear industry expert, in fact, an expert on steam generators specifically. In a video available on YouTube (and which I was cameraman for part of), Gundersen described the effect, saying the tubes would "pop like popcorn... pop, pop, pop."
More recently, John Large, distinguished nuclear engineer in England, and also, like Gundersen, contracted by FOE to look into the San Onofre issues, similarly postulated multiple-tube failures were possible if Unit 2 is restarted, due to the degraded condition of so many of the tubes, the inexactness of the tube wall measuring devices, the inexactness of the predictions of future tube failure (in part due to the poor quality of the measurements), the randomness of some of the vibration problems, and the like.
Additionally, the DAB Safety Team, which also has experts on it or associated with it (some of whom were formerly or are currently employed by SCE (and also includes myself, who has never worked there)), agrees with these assessments.
Several problems are all working against SanO.
One is that while the experts all agree that Fluid Elastic Instability occurred in Unit 3, there is disagreement as to whether it occurred in Unit 2 (for what it's worth, I think not). All agree that the design of SanO's steam generators has no "in-plane" support to prevent FEI in the in-plane direction (that means along the length of the u-bends).
Some of SCE's hired outside industry experts believe that friction forces from the out-of-plane supports has been and will be sufficient to prevent FEI in Unit 2. Other experts disagree with this assessment, including some of SCE's own hired outside industry experts.
In any event, the eight tubes that failed pressure testing point to a second problem: Let's say Unit 2 won't suffer FEI at 70% power -- for 5 months or at all. But what if there is a main steam line break, with a main steam line isolation valve failure? That is known as a "design basis accident" which means the utility is expected to prove it has a workable plan to handle such an event. Experts I've talked to do not believe SCE can possibly make that claim -- there would likely be a large radiation release resulting from loss of coolant, with possible core damage. The wear rates in Unit 2 were too high, even without FEI -- just from flow induced vibration. (FIV is a chaotic (random) vibration, FEI is rhythmic and coordinated -- the tubes all sway together in unison.)
Lastly, San Onofre exposed a generic problem for Pressurized Water Reactors (that is, reactors that use steam generators) everywhere: Multiple tube failures are possible. They can result from FEI or from any leak if unchecked, because one pinhole can expand over a period of an hour or so (I forget the average time frame this can happen but it's pretty short) and then it can snap off and hit another tube, causing multiple tube complete ruptures. Of the eight tubes in Unit 3 that failed pressure testing, three of them failed below main steam line break pressure differences expected between the inside of the tubes and the outside. So that's a second route to multiple tube failures, though perhaps not exactly a "cascade" effect of one tube ripping apart and slamming into another -- though it could initiate that effect, too.
NRC regulations do not currently appear to accept the concept of multiple tube failures. Once considered impossible because the reactor would be shut down first, it's now clear that there was at least one instance where a cascade of tube failures was possible -- San Onofre. San Onofre has proven conclusively to the industry that it can happen, although fortunately it did not occur here last January.
No one can say exactly how close to such an event we actually came. No one can truthfully say for certainly that can't happen here. It almost did.
(4) Press release regarding the scoping memo from Woman's Energy Matters:
For immediate release January 29, 2013
Contact: Barbara George, Exec. Dir. Women's Energy Matters 415-755-3147
CPUC'S SAN ONOFRE INVESTIGATION: PARTIES CRY FOUL
Parties to the California Public Utilities Commission's (CPUC)
investigation of the San Onofre nuclear generating station outage are
crying foul over ongoing procedural delays and a narrow Scoping Memo
issued Tues. Jan. 28th. Women's Energy Matters, the Coalition to
Decommission San Onofre, United Public Workers For Action and Michael
Aguirre charge that both seem designed to force southern California
customers to pay even higher rates in the next couple of years to fund
Edison's reckless plan to restart one of its severely damaged reactors —
instead of getting immediate refunds for the year the nuclear plant has
been offline. Parties ask CPUC to stop paying for these lemons now, and
plan for permanent replacement resources instead.
The Memo would put off refunds until 2014 or even 2015, although
SoCalEdison and SDG&E collected $57m a month in 2012 as if SONGS were
operating. It seems to have deleted any review of 2012 costs for
replacement power (promised in the original Order), or most importantly,
a review of SCE's mismanagement of the Steam Generator Replacement
Project, which appears to effectively deny refunds for any of the $700
million spent on it since 2006.
Edison's radical redesign of the steam generators caused structural
failures after less than two years an industry record. SCE falsely
told the Nuclear Regulatory Commission (NRC) and CPUC that the
replacement generators were "like for like," avoiding a "License
Amendment Review" which might have caught problems such as an
error-filled computer simulation.
Friends of the Earth are suing NRC to force a License Amendment Review
for any restart. Edison plans to restart the slightly less damaged Unit
2 as soon as NRC gives the go-ahead (expected in March/April), run it at
70% power for 5 months and then shut down for inspections.
Local residents warn that SCE's plan puts California's economy at risk,
along with more than 8 million people who live within 50 miles of SONGS.
That's the distance from Fukushima Daiichi that the NRC recommended for
Americans to evacuate. SCE's testimony on possible rate reductions
proposed nearly full funding for most operations but hardly anything for
Quality Assurance and whistleblower protection.
The Memo offers a booby prize, agreeing to consider "community outreach
and emergency preparedness." Parties say they intend to fully explore
California's emergency plans for nuclear accidents, which dates back to
the 1980s and involves multiple agencies. They call for expanding the
evacuation zone from the current 10 miles to 50, extensive real-time
radiation monitoring with public reports, and including earthquake
impacts on emergency planning, which NRC rejected back in the 1980s,
claiming "earthquakes are no worse than fog." Prior to any restart,
parties will insist on a real-world emergency drill, to test emergency
plans that are currently only paper exercises conducted every 3 years by
Women's Energy Matters (WEM) has been a public interest party
("intervenor") in CPUC cases since 2001. See more information at
For in-depth information about SONGS, see http://sanonofresafety.org
CPUC's Scoping Memo is posted at
Broadly stated, the scope of the future Phases of this OII are
envisioned as follows:
• Phase 2 whether any reductions to SCE's rate base and SCE's 2012
revenue requirement are warranted or required due to the extended SONGS
outages… Memo, p. 4. [Note: a decision on Phase 1 would come in August
at the earliest; no date was set for Phase 2 to begin.]
…SCE and SDG&E argue that the Commission may not order refunds of any
expenses related to the SONGS outages … because it would constitute
"impermissible retroactive ratemaking." Even if the Commission were
authorized to make such an order, the utilities contend no refunds could
occur prior to the 2015 GRCs. [GRC = General Rate Case]. Memo, pp. 5-6.
[Note: The Memo asks parties to submit legal briefs on these issues,
indicating that Commission lawyers lack confidence whether refunds are
legal. The question arises: why did the Commission's recent decision on
SCE's 2012 GRC allow SCE & SDG&E to keep collecting SONGS revenues, when
the plant was already out of service and the long-delayed Investigation
was finally about to begin?]
[T]o ensure that review of community outreach is considered in
conjunction with local emergency preparedness activities, this Scoping
Memorandum and Ruling explicitly authorizes review of SCE's actions and
expenditures for community outreach related to the SONGS. p. 10.
Edison's Testimony and other public documents are posted at
The following is an excerpt from SCE's December 15, 2012 Testimony, pp.
The Engineering functional group consists of the Design Engineering,
Nuclear Safety Concerns, and Nuclear Oversight/Assessment divisions. …
Plant Engineers are familiar with the design basis of their assigned
plant systems… [Ed. note - In other words these people might
understand what went wrong with the new steam generator design — so
let's get rid of them]
Nuclear Safety Concerns provides an alternate, confidential mechanism
for SONGS workers to identify conditions related to their personal
safety, the health and safety of the public, or compliance with NRC
Nuclear Oversight and Assessment develops, maintains, and oversees the
Quality Assurance and Performance Improvement programs. It provides
managerial and administrative controls to assure safe operation and
maintenance of SONGS systems required by the NRC… SCE thus
conservatively estimates that it must expend at least 15% of the
GRC-authorized amount for the Engineering group regardless of whether
SONGS is operational, and this 15% of the GRC-authorized amount should
not be subject to refund.
(5) Statement regarding the scoping memo from Michael J. Aguirre, Esq.:
Contrary to what the PUC news release led the public to believe the PUC issued a "scoping" memorandum today limiting the review of San Onofre issues to those helpful to SCE and hurtful to the public. The scoping memo makes a mockery of the PUC "investigation" because it allows only a very limited review of the issues: (1) assessing the reasonableness of SCE's actions and expenditures after the outage; (2) whether SCE's 2012 expenditures for SONGs was reasonable; (3) the reasonableness of SCE's expenditures for community outreach; and (4) whether SCE should refund any money they were allowed to keep under the General Rate Case issued in December 2012.
Here is what will not be allowed: (1) whether SCE was imprudent and unreasonable in spending $800 million for the 4 new generators to replace the previous generators which tube problems, when the new generators had tube problems worse than those replaced; (2) whether the 4 generators should be taken out of the rate base. The Scoping Order does not address the first question and pushes off the second to some undetermined time in the future. The PUC has mislead the People of California by issuing a news release announcing an investigation while issuing an order that does not permit a reasonable investigation.
It is clear that the PUC has decided to get San Onofre back in operation as soon as possible. The PUC "investigation" is nothing more than a cynical public relations stunt.
Michael J. Aguirre, Esq.
AGUIRRE, MORRIS & SEVERSON LLP
(6) One-year commemoration of the near-destruction of Southern California planned:
NoSanO: 1st Anniversary of the San Onofre Nuclear Shutdown
The 1st Annual NoSanO Anniversary: 1 Year Without San Onofre. Featuring Ed Begley, Jr at 6:30p, Live Music by the Kalama Brothers, and the U.S. Green Chamber of Commerce tells us about The Green Job Market.
One year ago, a radiation leak nearly became a major nuclear disaster. A year without any blackouts proves we don't need to live with the danger. Safe, reliable and sustainable alternatives will provide energy and jobs.
Join us for an enjoyable and informative evening.
$10 Prepaid Admission gets you in the door, plus a Slice of Pizza and a beverage of your choice (wine, beer, soft drink). $15 at the Door. Additional servings: $2.50 per slice / $5.00 beer or wine (other drinks free)
We also have a bus set up from San Diego's Balboa Park (SW corner of Park Blvd & President's Way) with stop at the Oceanside Transit Ctr for only $10 Roundtrip!
Select the SD-O'side Bus Ticket on the Ticket Menu here and also your Pick-Up location (San Diego or Oceanside) from the Drop-down Menu in answer to the Question at Checkout.
Ed Begley, Jr, is an American actor and environmentalist who has appeared in hundreds of films, television shows, and stage performances. He is best known for his role as Dr. Victor Ehrlich, on the television series St. Elsewhere, for which he received six consecutive Emmy Award nominations, and his most recent reality show about green living called Living With Ed on Planet Green with his wife, actress Rachelle Carson-Begley. For more on Ed, see http://www.edbegley.com
The Kalama Brothers: Diverse in their music from Traditional and Contemporary Hawaiian, to Soul and Classic Rock. The magic of their harmonies is something special! They'll fill your hearts with their love and touch your souls with their warmth. http://http://www.kalamabrothers.com
The U.S. Green Chamber of Commerce's goal is to facilitate and support sustainable business practices that spur innovation, job creation, energy efficiency and an overall brighter economic future. The Chamber was established in San Diego in 2009, and has since expanded across the nation with branches in Texas and Florida. http://http://www.usgreenchamber.com
Presented by San Clemente Green, with more info on www. SanOnofreSafety.org
Martha Sullivan posted in North County San Diego Anti-Nuke Working Group
2:25pm Jan 29
Folks, we really need you to COMMIT and Buy your Ticket(s) online -- you'll save $5 and time at the Door, too! And the RT bus from San Diego at $10 is a STEAL -- plus, you'll be wristbanded On the Bus and ready to blow through the Door when you arrive.
Just DO IT: http://www.brownpapertickets.com/event/319537 !
NoSanO: 1st Anniversary of the San Onofre Nuclear Shutdown
(7) Contact information for the author of this newsletter:
** Ace Hoffman,
** POB 1936, Carlsbad CA 92018
** home page: www.animatedsoftware.com
Tuesday, January 29, 2013
Media Alert 13-01-28
Allegation – NRC AIT Report Incomplete, Inconclusive, Inconsistent and Unacceptable
Media Contact: Don Leichtling (619) 296-9928 or Ace Hoffman (760) 720-7261
SONGS UNIT 3 RSG REAL ROOT CAUSE: Lack of “Critical Questioning & Investigative Attitude” by SCE, MHI and NRC Region IV and AIT Team.
The DAB Safety Team has transmitted the following Allegation to “Office of the Chairman of the NRC” and “The United States Senate Committee on Environment and Public Work.”
It is the DAB Safety Team’s goal to help educate both the NRC and the Public by providing unbiased, logical and factual information in order to help assess the real dangers of any San Onofre Unit 2 restart. According to Press Reports and San Onofre Insiders, Unit 2 permission for restart by the NRC is imminent yet the REAL Root Cause for the $1 Billion destruction of Units 2 and 3 RSGs (Including equipment cost and expenses) has not yet even been determined. The Public does not know the status of SCE’s ongoing cause evaluations, SCE’s response to 32 NRR’s RAI’s and NRC’s Special San Onofre Inspections. We like to remind NRC San Onofre Special Panel, what NRC Chairman Macfarlane said during her recent Fukushima Trip, “Regulators may need to be ‘buffered’ from political winds, but they need to be fully subjected to the pressure of scientific and engineering truth and cannot be allowed to make decisions or order actions that are ‘independent’ of facts.” The NRC rush to a faulty judgment cannot be allowed to compromise Public Safety just to please SCE, as this conflicts with President Obama's Policy, the new NRC Chairman’s Standards and the advice of NRC retired Branch Chiefs who have also spoken out.
A NRC Branch Chief gifted with MIT Intelligence, Intuition and a Sixth Sense, told an anonymous participant at an Industry Conference, “Sir, to resolve any complex technical problem and understand unclear regulations, you have to, ‘Read and reread in between the lines’, use, ‘Critical questioning and an investigative attitude’ and ‘Solid teamwork & alignment.”
SONGS UNIT 3 RSG ROOT CAUSE: It appears that Complacent SCE and Inexperienced MHI Engineers did not perform proper academic research and industry bench marking about the potential adverse consequences of the reduction of original CE steam generator pressures from 900 psi to say, 800 psi on fluid elastic instability and flow-induced vibrations. These lower secondary steam operating pressure (800-833 psia) are the primary cause for shortening the life of SONGS Original Combustion Engineering Generators due to increased tube wear and plugging caused by flow-induced random vibrations and destruction of SONGS Unit 3 Replacement Steam Generators due to flow-induced random vibrations, Mitsubishi flowering effects and steam voids or steam dry-outs (AKA fluid elastic instability). In addition, SCE Engineers prepared defective 10 CFR 50.59 Evaluation and design specifications, which were not challenged by MHI, and adequately reviewed by NRC Region IV. MHI at the direction of SCE Engineers made numerous untested and unanalyzed design changes to the steam generators under the pretense of “like for like”, and Elmo Collins said, “The guts of the machinery look …. Different.”
Therefore the SONGS UNIT 3 RSG REAL ROOT CAUSE: Lack of “Critical Questioning & Investigative Attitude” by SCE, MHI and NRC Region IV and AIT Team.
· Lessons Learnt: MHI, AIT, DAB Safety Team together with other World Nuclear and SG Experts / Manufacturers agree that, “Lower secondary steam operating pressures (800-833 psi) are severe and can easily cause SG flow-induced random vibrations and fluid elastic instability.” At lower secondary steam operating pressures (800-833 psi), the utility can generate more thermal megawatts out of the SG and hence add more power to the grid and thereby make more money for SCE/EIX Officers and shareholders. These lower secondary steam operating pressures (800-833 psi) were the primary cause for shortening the life of SONGS Original Combustion Engineering Steam Generators. The real lesson learnt is that ALL parties must follow the NRC Chairman’s advice, pay attention to their work and always ensure that public safety is THE overriding obligation for the regulators, licensee, its vendors and contractors. NRC “Reasonable Assurance” for the protection of adequate health and safety of the public from postulated radiological accidents cannot be compromised at any time during the design, operation, fabrication, testing, surveillance, maintenance and inspection of a nuclear power plant. SCE and its offsite response organizations need to demonstrate; (1) Feasibility of an Operator Action during a postulated main steam line break with multiple Unit 2 steam generator tube ruptures, (2) They can effectively implement emergency plans and procedures with zero Drills/Exercise Performance indicator failures during a Fully Staffed NRC/FEMA evaluated exercise prior to any Unit 2 Restart.
· Comments about the NRC Augmented Inspection Team San Onofre Report
NOTE: We highly recommend that NRC Augmented Inspection Team and NRC San Onofre Special Review Panel thoroughly review SONGS Unit 2 Return to Service MHI, AREVA, Westinghouse, DAB Safety Team and John Large Reports, then carefully examine the operational differences between Unit 2 and 3 and then update the NRC AIT report with a FACTUAL Root cause for FEI in Unit 3 and NO FEI in Unit 2. NRC San Onofre Special Review Panel also needs to review the SONGS Unit 2 Restart Reports (done by SCE, Westinghouse, AREVA and MHI), SCE Unit 3 Root Cause Evaluation, NRC AIT Report, ATHOS Modeling Results and Unit 2 Operational Data and then arrive: (1) At a unanimous, clear and concise conclusion whether FEI occurred in Unit 2 or not, and (2) Provide a GAP ANALYSIS (The scientific, technical and engineering reasons why all these reports are so different) prior the February 12, 2013 NRC Public Meeting.
The AIT inspection concluded that: (1) SCE was adequately pursuing the causes of the unexpected steam generator tube-to-tube degradation. In an effort to identify the causes, SCE retained a significant number of outside industry experts, consultants, and steam generator manufacturers, including Westinghouse and AREVA to perform thermal-hydraulic and flow induced vibration modeling and analysis; (2) The combination of unpredicted, adverse thermal hydraulic conditions and insufficient contact forces in the upper tube bundle caused a phenomenon called “fluid-elastic instability” which was a significant contributor to the tube to tube wear resulting in the tube leak. The team concluded that the differences in severity of the tube-to-tube wear between Unit 2 and Unit 3 may be related to changes to the manufacturing/fabrication of the tubes and other components which may have resulted in increased clearances between the anti-vibration bars and the tubes; (3) Due to modeling errors, the replacement generators were not designed with adequate thermal hydraulic margin to preclude the onset of fluid-elastic instability. Unless changes are made to the operation or configuration of the steam generators, high fluid velocities and high void fractions in localized regions in the u-bend will continue to cause excessive and accelerated tube wear that could result in tube leakage and/or tube rupture; (4) The thermal hydraulic phenomena contributing to the fluid-elastic instability is present in both Unit 2 and 3 steam generators; (5) Based on the updated final safety analysis report description of the original steam generators, the steam generators’ major design changes were appropriately reviewed in accordance with the 10 CFR 50.59 requirements.
So based on a review of the AIT Report and World’s Experts, the three potential causes, which were significant contributors to the “fluid-elastic instability” in SONGS Unit 3 and the tube-to-tube wear resulting in the tube leak are as follows:
A. Insufficient contact tube-to-AVB forces and differences in manufacturing or fabrication of the tubes and other components between Units 2 & 3.
B. Due to modeling errors, the SONGS replacement generators were not designed with adequate thermal hydraulic margin to preclude the onset of fluid-elastic instability.
C. Differences between Unit 2 and Unit 3’s Operational Factors.
A. Let us now examine that whether insufficient contact tube-to-AVB forces in the Unit 3 upper tube bundle caused “fluid-elastic instability” which was a significant contributor to the tube-to-tube wear resulting in the tube leak.
A.1- MHI states, “By design, U-bend support in the in-plane direction was not provided for the SONGS SG’s”. In the design stage, MHI considered that the tube U-bend support in the out-of-plane direction designed for “zero” tube-to-AVB gap in hot condition was sufficient to prevent the tube from becoming fluid-elastic unstable during operation based on the MHI experiences and contemporary practice. MHI postulated that a “zero” gap in the hot condition does not necessarily ensure that the support is active and that contact force between the tube and the AVB is required for the support to be considered active. The most likely cause of the observed tube-to-tube wear is multiple consecutive AVB supports becoming inactive during operation. This is attributed to redistribution of the tube-to-AVB-gaps under the fluid hydrodynamic pressure exerted on the tubes during operation. This phenomenon is called by MHI, “tube bundle flowering” and is postulated to result in a spreading of the tube U-bends in the out-of-plane direction to varying degrees based on their location in the tube bundle (the hydrodynamic pressure varies within the U bend). This tube U-bend spreading causes an increase of the tube-to-AVB gap sizes and decrease of tube-to-AVB contact forces rendering the AVB supports inactive and potentially significantly contributing to tube FEI. Observations Common to BOTH Unit-2 and Unit-3: The AVBs, end caps, and retainer bars were manufactured according to the design. It was confirmed that there were no significant gaps between the AVBs and tubes, which might have contributed to excessive tube vibration because the AVBs appear to be virtually in contact with tubes. MHI states, “The higher than typical void fraction is a result of a very large and tightly packed tube bundle, particularly in the U-bend, with high heat flux in the hot leg side. Because this high void fraction is a potentially major cause of the tube FEI, and consequently unexpected tube wear (as it affects both the flow velocity and the damping factors) the correlation between the void fraction (steam quality) and the number of
tubes with wear in a given void fraction region was investigated. From this investigation, a
strong correlation between the void fraction (steam quality) and the percentage of tubes with
the Type 1 (tube-to-tube) and Type 2 (tube-to-AVB) wear was identified. The correlation between flow velocity and the number of tubes with wear was also investigated. The results show that when the flow velocity is high, the percentage of tubes with wear increases, even though this correlation is not as strong as that between the void fraction (steam quality) and the percentage of tubes with wear.”
A.2 – AREVA states, “At 100% power, the thermal-hydraulic conditions in the U-bend region of the SONGS replacement steam generators exceeded the past successful operational envelope for U-bend nuclear steam generators based on presently available data. The primary source of tube-to-AVB contact forces is the restraint provided by the retaining bars and bridges, reacting against the component dimensional dispersion of the tubes and AVBs. Contact forces are available for both cold and hot conditions. Contact forces significantly increase at normal operating temperature and pressure due to diametric expansion of the tubes and thermal growth of the AVBs. After fluid elastic instability develops, the amplitude of in-plane motion continuously increases and the forces needed to prevent in-plane motion at any given AVB location become relatively large. Hence shortly after instability occurs, U-bends begin to swing in Mode 1 and overcome hindrance at any AVB location.”
A.3 – Westinghouse states, “Test data shows that the onset of in-plane (IP) vibration requires much higher velocities than the onset of out-of-plane (OP) fluid-elastic excitation. Hence, a tube that may vibrate in-plane (IP) would definitely be unstable OP. A small AVB gap that would be considered active in the OP mode would also be active in the IP mode because the small gap will prevent significant in-plane motion due to lack of clearance (gap) for the combined OP and IP motions. Thus, a contact force is not required to prevent significant IP motion. Manufacturing Considerations: There are several potential manufacturing considerations associated with review of the design drawings based on Westinghouse experience. The first two are related to increased proximity potential that is likely associated with the ECT evidence for proximity. Two others are associated with the AVB configuration and the additional orthogonal support structure that can interact with the first two during manufacturing. Another relates to AVB fabrication tolerances. These potential issues include: (1) The smaller nominal in-plane spacing between large radius U-bend tubes than comparable Westinghouse experience, (2) The much larger relative shrinkage of two sides (cold leg and hot leg) of each tube that can occur within the tubesheet drilling tolerances. Differences in axial shrinkage of tube legs can change the shape of the U-bends and reduce in-plane clearances between tubes from what was installed prior to hydraulic expansion, (3) The potential for the ends of the lateral sets of AVBs (designated as side narrow and side wide on the Design Anti-Vibration Bar Assembly Drawing that are attached to the AVB support structure on the sides of the tube bundle to become displaced from their intended positions during lower shell assembly rotation, (4) The potential for the 13 orthogonal bridge structure segments that are welded to the ends of AVB end cap extensions to produce reactions inside the bundle due to weld shrinkage and added weight during bundle rotation, and (5) Control of AVB fabrication tolerances sufficient to avoid undesirable interactions within the bundle. If AVBs are not flat with no twist in the unrestrained state they can tend to spread tube columns and introduce unexpected gaps greater than nominal inside the bundle away from the fixed weld spacing. The weight of the additional support structure after installation could accentuate any of the above potential issues. There is insufficient evidence to conclude that any of the listed potential issues are directly responsible for the unexpected tube wear, but these issues could all lead to unexpected tube/AVB fit-up conditions that would support the amplitude limited fluid-elastic vibration mechanism. None were extensively treated in the SCE root cause evaluation.”
A.4 - John Large, Internationally Known Scientist and Chartered Nuclear Engineer from London says about the SONGS Unit 2 Replacement Steam Generators (RSGs) AVB Structure, “It impossible to reliably predict the effectiveness of the many thousands of AVB contact points for when the tube bundle is in a hot, pressurized operational state. The combination of the omission of the in-plane AVB restraints, the unique in-plane activity levels of the SONGS RSGs, together the very demanding interpretation of the remote probe data from the cold and depressurized tube inspection, render forecasting the wear of the tubes and many thousands of restraint components when in hot and pressurized service very challenging indeed. Phasing of AVB-TSP Wear -v- TTW: I reason that, overall, the tube wear process comprises two distinct phases: First, the AVB (and TSP) -to-tube contact points wear with the result that whatever level of effectiveness is in play declines. Then, with the U-bend free-span sections increased by loss of intermediate AVB restraint(s), the individual tubes in the U-bend region are rendered very susceptible to FEI induced motion and TTW. Whereas the OAs commissioned by SCE broadly agree that the wear mechanics comprises two phases, there are strong differences over the cause of the first phase comprising in-plane AVB wear: AREVA claim this is caused by in-plane FEI whereas, the contrary, Mitsubishi (and Westinghouse) favor random perturbations in the fluid flow regime to be the tube motion excitation cause. Put simply: i) if AREVA is correct then reducing the reactor power to 70% will eliminate FEI, AVB effectiveness will cease to decline further and TTW will be arrested; however, to the contrary, ii) if Mitsubishi is right then, even at the 70% power level, the AVB restraint effectiveness will continue to decline thereby freeing up longer free-span tube sections that are more susceptible to TTW; or that iii) the assertion of neither party is wholly or partly correct. I find that the AVB assembly, which features strongly in the onset of TTW, is clearly designed to cope only with out-of-plane tube motion since there is little designed-in resistance to movement in the in-plane direction - because of this, it is just chance (a virtually random combination of manufacturing variations, expansion andpressurization, etc) that determines the in-plane effectiveness of the AVBs.”
A.5 - SCE claims, “The facts identified in this analysis indicate that even though the Unit 3 tube bundle components (tubes and AVBs) might have been fabricated and assembled better, the tube "to" AVB gaps built gaps might have been in fact larger in the Unit 3 RSGs as suggested by the ECT results. Based on this, it cannot be ruled out that the tube-to-AVB gaps are larger and more uniform in the Unit 3 RSGs than the Unit 2 RSG’s. This might have resulted in reduction of the tube-to-AVB contact force and consequently in multiple consecutive AVB supports being inactive. Inactive tube support might have resulted in "tube-to-tube" wear.
A.6 – DAB Safety Team Conclusions: SONGS Unit 3 RSG’s were operating outside SONGS Technical Specification Limits for Reactor Thermal Power and Current Licensing Basis for Design Basis Accident Conditions. Let us summarize what these experts are saying: (1) AREVA is saying, “After fluid elastic instability develops, the amplitude of in-plane motion continuously increases and the forces needed to prevent in-plane motion at any given AVB location become relatively large. Hence shortly after instability occurs, U-bends begin to swing in Mode 1 and overcome hindrance at any AVB location”, (2) Westinghouse is saying, “A small AVB gap that would be considered active in the OP mode would also be active in the IP mode because the small gap will prevent significant in-plane motion due to lack of clearance (gap) for the combined OP and IP motions. Thus, a contact force is not required to prevent significant IP motion”, (3) MHI is saying, “High steam flows and cross-flow velocities combined with narrow tube pitch-to-diameter ratio caused elastic deformation of the U-tube bundle from the beginning of the Unit 3 cycle, which initiated the process of tube-to-AVB wear and insufficient contact forces between tubes and AVBs. Tube bundle distortion is considered a major contributing cause to the mechanism of tube-to-tube/AVB/TSP wear seen in the Unit 3 SG’s. After 11 months of wear, contact forces were virtually eliminated between the tube and AVBs in the areas of highest wear of Unit 3 tubes as confirmed by ECT and visual inspections. According to MHI Technical Document, the RSG Anti-bar Vibration Structures were only designed for out-of plane vibrations and not in-plane vibrations”, (4) John Large is saying, “Its impossible to reliably predict the effectiveness of the many thousands of AVB contact points for when the tube bundle is in a hot, pressurized operational state.” In essence, all the three NEI qualified, “U.S. Nuclear Plant Designers” are saying: (a) Firstly, these AVB structures are not designed to prevent in-plane vibrations, (b) Secondly, once fluid elastic instability develops, the AVB is not strong enough to prevent large in-plane amplitude of tubes, (c) Thirdly, excessive fluid-induced random vibrations and fluid hydrodynamic pressures (Mitsubishi Flowering Effect) cause the loss of tube-AVB contact forces and increase the tube-to-AVB gaps and the onset of fluid elastic instability develops, and (d) Lastly, tube-to-AVB gaps and contact forces are irrelevant to prevent fluid elastic instability from progressing and causing tube-to-tube and tube-to-AVB wear. Therefore, based on a review of MHI, AREVA and Westinghouse excerpts shown above, it is concluded, that FEI, flow-induced random vibrations and MHI Flowering effect redistributed the tube-to-AVB gaps in Unit 3 RSG’s as measured by SCE. Westinghouse even goes this far to state, “None of the MHI manufacturing issues were extensively treated in the SCE root cause evaluation.” Hence, It is concluded that the NRC and SCE claims that insufficient contact forces in Unit 3 Tube-to-AVB Gaps ALONE caused tube “to” tube wear are misleading, erroneous and designed to put the blame on MHI for purposes of making SCE look good in the public’s eyes and for collecting insurance money from MHI’s so-called “manufacturing defects”.
B. Let us now examine the effects of modeling errors, that the SONGS replacement generators were not designed with adequate thermal hydraulic margin to preclude the onset of fluid-elastic instability.
B.1 – NRC AIT Report states, “The ATHOS thermal-hydraulic model predicts bulk fluid behavior based on first principals and empirical correlations and as a result, it is not able to evaluate mechanical, fabrication, or structural material differences or other phenomena that may be unique to each steam generator. Therefore this analysis cannot account for these mechanical factors and differences which could very likely also be contributing to the tube degradation.”
B.2 – Ivan Cotton states, “Fluid elastic instability is one of the most damaging types of instabilities encountered in heat exchangers and steam generators and can impose a severe economic penalty on the power and chemical industries. At present our understanding of the mechanisms leading to fluid-elastic instability is very limited and more experiments are needed to more fully delineate the conditions for the onset of fluid-elastic instability.” Such experimentation should only be done in a sealed lab, NOT our environment with the lives of eight million local residents at stake in the outcome!
B.3 – Ishihara, Kunihiko and Kitayama state, “Tube vibrations become large as tube thickness/diameter ratio (T/D) increases and tube length/diameter ratio (L/D) decreases, and the tube vibrations strongly depend on the dynamic characteristics of tubes such as the natural frequency and the damping ability.” In the case of SanO’s replacement SGs, the tubes, especially in the U-bend region, were too close together, poorly restrained, poorly damped, along with too much heat flux and an inappropriate pressure-to-flow ratio, along with other causes which resulted in FEI and FIV.
B.4 – Fairewind states, “Realistically, the 3-D steam analysis is not accurate enough to apply to such important safety related determinations. To make such mathematical risk 3-D analysis, a very large margin of error must be applied, and that has not been done. For example, if the 3-D steam analysis determines that plugging 100 tubes is a solution, then plugging ten times that number might be the appropriate solution due to the mathematical errors in the 3-D analysis being applied by Edison and Mitsubishi. Edison has taken many of its mathematical/computer models as “gospel” rather than accepting their wide margins of error, which directly affect safety, which is precisely what got them in trouble in the first place. It should have been obvious to Edison that MHI FIT-III has not been benchmarked, and had not been previously used in licensing procedures showing that the use of FIT-III might have an adverse effect on the FSAR safety analysis thus necessitating the entire license amendment review and public hearing process. As noted by the AIT, Edison approved the use of FIT-III code even though the code was not benchmarked, nor identified as acceptable in the FSAR. Consequently, Edison operated San Onofre without knowing the uncertainties in the Replacement Steam Generators’ performance characteristics. Predicted liquid levels, pressure drops, vibrations, and temperatures at both Units 2 and 3 were all subject to unknown uncertainties during both normal and abnormal operations. In my opinion, by approving the use of an un-benchmarked and untested design tool like FIT-III, Edison did not did not meet the requirements expected from a nuclear licensee.
Use of an un-benchmarked computer code that is not included in the FSAR protocol demands a formal FSAR license amendment process including the requisite public hearings.”
B.5 – Mitra, V.K. Dhir, I. Catton state, “Flow induced vibrations in heat exchanger tubes have led to numerous accidents and economic losses in the past. Efforts have been made to systematically study the cause of these vibrations and develop remedial design criteria for their avoidance. Instability was clearly seen in single phase and two-phase flow and the critical flow velocity was found to be proportional to tube mass. It is also found that nucleate boiling on the tube surface is also found to have a stabilizing effect on fluid-elastic instability.”
B.6 - John Large, says “Factual Issues v) & vi) – SONGS SG Comparison to Other Operating SGs: I identify a number of issues with the representation of Figures 4-3 and 5-1 of the AREVA Tube-to-Tube Report, including: i) it is not exactly clear which properties are being represented on the spider diagram for comparison with the other operational SGs; even so ii) since it is most unlikely that AREVA has undertaken a comprehensive (ATHOS) simulation of each of the five nominated SGs, the comparisons drawn are likely to be between aggregate or bulk flows within the entire tube bundle of each SG; iii) as acknowledged by AREVA, the SONGS RSGs are dominated by in-plane flow regimes whereas all other SGs are characterized by out-of-plane flow regimes; and iv) none of the comparative SGs has been identified. In other words, unless the spider diagrams of Figure 4-3 and 5-1 somehow, and I cannot reason how, are making a direct comparison of the complex two-phase fluid cross-flow situation in the SONGS and other five comparative plant steam generators, then these figures only provide the bases of a somewhat meaningless comparisons.”
B.7 – SCE states that SONGS Unit 3 Damage (FEI) was caused due to outdated MHI Thermal-Hydraulic Computer Models. According to NRC AIT Report, SONGS did not specify the value of FEI in its Design and Performance Specifications SO23-617-1. Academic Researchers have discussed and warned about the adverse effects of fluid elastic instability (tube-to-tube wear) in nuclear steam generators since the 1970’s. Westinghouse and Combustion Engineering (CE) have designed CE replacement steam generators (RSGs) to prevent the adverse effects of fluid elastic instability since 2000’s (e.g., PVNGS).
B.8 – The NRC AIT Report dated November 9, 2012 states, “the FIT-III thermal-hydraulic model was still in-progress at the time of the inspection and no final conclusions were reached for the cause of the non-conservative flow velocities, which were used as inputs in the tube vibration analysis and resulted in non-conservative stability ratios. Since the licensee had not completed the cause evaluation for this unresolved item, the inspectors were not able to make a final determination of whether a performance deficiency or violation of NRC requirements occurred. The inspectors were informed that Mitsubishi was performing an evaluation of the potential factors that contributed to the low flow velocities in FIT-III relative to the velocities calculated by the ATHOS model developed after the tube leak event in Unit 3. This evaluation was included in Document SO23-617-1-M1530, Revision 1, which also intended to demonstrate the validity of FIT-III results for the original tube vibration analysis. This evaluation was still being finalized and not yet approved by Edison. The licensee and Mitsubishi continued to evaluate this unresolved item and no final conclusions were reached at the time of the inspection. The NRC is continuing to perform independent reviews of existing information, and will conduct additional reviews as new information becomes available.” In another related finding, NRC inspectors stated, “SCE Engineers did not meet Procedure SO123-XXIV-37.8.26 requirements to ensure the design of the retainer bar was adequate with respect to the certified design specification. Specifically, the licensee failed to ensure that there was sufficient analytical effort in the design methodology of the anti-vibration bar assembly to support the conclusion that tube wear would not occur, as a result of contact with the retainer bars due to flow-induced vibration. The inspectors determined that the requirements for flow-induced vibration in the certified design specification, along with the expectations in Procedure SO123-XXIV-37.8.26, provided sufficient information to reasonably foresee the inadequate design of the retainer bars during the review and approval of design Calculations SO23-617-1-C749 and SO23-617-1-C157, including the associated design drawings provided by Mitsubishi.”
B.9 – Arnie Gundersen states, “Not only is Mitsubishi unfamiliar with the tightly packed CE design, but Edison’s engineers added so many untested variables to the new fabrication that this new design had a significantly increased risk of failure. As a result of the very tight pitch to diameter ratios used in the original CE steam generators, Mitsubishi fabricated a broached plate design that allows almost no water to reach the top of the steam generator. The maximum quality of the water/steam mixture at the top of the steam generator in the U-Bend region should be approximately 40 to 50 percent, i.e. half water and half steam. With the Mitsubishi design the top of the U-tubes are almost dry in some regions. Without liquid in the mixture, there is no damping against vibration, and therefore a severe fluid-elastic instability developed. Because of the Edison/Mitsubishi steam generator changes, the top of the new steam generator is starved for water therefore making tube vibration inevitable. Furthermore, the problem appears to be exacerbated by Mitsubishi’s three-dimensional thermal-hydraulic analysis determining how the steam and water mix at the top of the tubes that has been benchmarked against the Westinghouse design but not the original CE design. The real problem in the replacement steam generators at San Onofre is that too much steam and too little water is causing the tubes to vibrate violently in the U-bend region. The tubes are quickly wearing themselves thin enough to completely fail pressure testing. Even if the new tubes are actively not leaking or have not ruptured, the tubes in the Mitsubishi fabrication are at risk of bursting in a main steam line accident scenario and spewing radiation into the air.”
B.10 – Comment on Limitations of ATHOS thermal-hydraulic Models: The ATHOS thermal-hydraulic model predicts bulk fluid behavior based on first principals and empirical correlations and as a result, it is not able to evaluate mechanical, fabrication, or structural material differences or other phenomena that may be unique to each steam generator. Furthermore, the combination of the omission of the in-plane AVB restraints, the unique in-plane activity levels of the SONGS RSGs, together with the very demanding interpretation of the remote probe data from the cold and depressurized tube inspection, render forecasting the wear of the tubes and many thousands of restraint components when in hot and pressurized service very challenging indeed. ATHOS thermal-hydraulic Models used for 70% power have not been benchmarked, and tested against SONGS Unit 2 RSG degraded tube bundles performance for several cycles of depressurized/pressurized operation. Hence, ATHOS analyses cannot accurately predict the behavior of pressurized degraded SG tube bundles and their interaction with their anti-vibration bar support structure, which could very likely contribute to unknown amounts of tube-to-tube wear and/or AVB degradation in Unit 2, at 70% or 100% power during a main steam line break accident resulting in a cascade of multiple steam generator tube ruptures.
B.11 – Conclusions on Modeling Errors: The NRC AIT Report dated November 9, 2012 states, “the FIT-III thermal-hydraulic model was still in-progress at the time of the inspection and no final conclusions were reached for the cause of the non-conservative flow velocities, which were used as inputs in the tube vibration analysis and resulted in non-conservative stability ratios. Since the licensee had not completed the cause evaluation for this unresolved item, the inspectors were not able to make a final determination of whether a performance deficiency or violation of NRC requirements occurred. SCE and MHI are both negligent because they did a very poor job of Industry and Academic Research benchmarking regarding the applicability of thermal-hydraulic computer models during the redesign of San Onofre’s original CE SGs. SCE is negligent because they did not check the results of MHI’s outdated Thermal-Hydraulic Computer Models to meet their specification and procedure requirements. This does not meet the NRC Chairman’s Standards. Therefore, it is concluded that SCE claims as stated above are not factual. SCE engineers did not check the work of MHI with a critical and questioning attitude and did not meet the San Onofre Design Procedures, 10CFR50, Appendix B, Quality assurance Standards and or NRC Regulations. NRC AIT Team jumped the boat by putting all the blame on MHI in July 2012 and now they are retreating in November 2012 by stating with a sunken face, “The inspectors were not able to make a final determination of whether a performance deficiency or violation of NRC requirements occurred.”
C. Let us now examine the other differences between Unit 2 and Unit 3’s Operational Factors, which were significant contributors to the “fluid-elastic instability” in San Onofre Unit 3 and the tube-to-tube wear resulting in the tube leak.
C.1 – Adverse Design/Operational Factors responsible for Fluid Elastic Instability: Low steam generator pressures (SONGS RSGs range 800-850 psi, the primary cause of the onset of severe vibrations) allow the onset of FEI, whereby U-tube bundle tubes start vibrating with very large amplitudes in the in-plane directions. Extremely hot and vibrating tubes need a little amount of water (aka damping, 1.5% water, steam-water mixture vapor Fraction 99.5%). Without the water, the extremely hot and vibrating tubes cannot dissipate their energy. In effect, one unstable tube drives its neighbor to instability through repeated violent impact events which causes tube leakage, tube failures at MSLB test conditions and/or unprecedented tube-tube wear, Tube-to-AVB/Tube Support Plates wear, as we saw in San Onofre Unit 3. So in review, due to narrow tube pitch to tube diameter, tube natural frequency, low tube clearances, in certain portions of the RSGs U-tubes bundle, fluid velocities exceed the critical velocities due to extremely high steam flows (100% power conditions). These high fluid velocities cause U-tubes to vibrate with very large amplitudes in the in-plane direction and literally hit other tubes with repeated and violent impacts. Due to lower secondary steam operating pressures (required to generate more heat, electricity and profits) and excessive pressure drops due to high flows and velocities, steam saturation temperature drops. This lowering of steam saturation temperature combined with high heat flux in the hot leg side of the U-tube bundle causes steam dry-outs to form (Vapor fraction >99%), known as “NO Effective Thin Tube Film Damping.” Thin film damping refers to the tendency of the steam inside the generators to create a thin film of water between the RSG tubes and the support structures and each other. That film is enough to help keep the tubes from vibrating with large amplitudes, hitting other tubes violently, and to protect the Anti-Vibration Bar support structures and maintain the tube-to-AVB gaps and contact forces. These adverse conditions in Unit 2 at 70% power operation (RTP) with the present defective design and degraded RSGs, known as fluid elastic instability (Tube-to-Tube Wear, or TTW) can lead to rapid U-tube failure from fatigue or tube-to-tube wear in Unit 2 due to a main steam line break as seen in Unit 3’s RSG’s. In summary, FEI is a phenomenon where due to San Onofre RSGs design intended for high steam flows causes the tubes to vibrate with increasingly larger amplitudes due to the fluid effective flow velocity exceeding its specific limit (critical velocity) for a given tube and its supporting conditions and a given thermal hydraulic environment. This occurs when the amount of energy imparted on the tube by the fluid is greater than the amount of energy that the tube can dissipate back to the fluid and to the supports. The lack of Nucleate boiling on the tube surface or absence of water is found to have a destabilizing effect on fluid-elastic stability.
C.2 – Unit 2 FEI Conflicting Operational Data
· NRC AIT Report SG Secondary U2/3 Pressure Range 833 - 942 psi
· SCE RCE SG Secondary U2/3 Pressure - 833 psi
· RCE Team Anonymous Member - Unit 2 SG Secondary Pressure 863 psi
· SONGS SG System Description Unit 2 SG Pressure Range 892 - 942 psi
· Westinghouse OA SG Secondary U2/3 Pressure ~ 838 psi, Void Fraction 99.55%
· SCE Enclosure 2, MHI ATHOS results - U2/3 Void Fraction 99.6%
· SCE Enclosure 2, Independent Expert results - ATHOS U2/3 Void Fraction 99.4%
· DAB Safety Team SG Secondary U2 Pressure 863 -942 psi, Void Fraction 96-98%
· SONGS Plant Daily Briefing Unit 3 Electrical Generation – 1186 MWe
· SONGS Plant Daily Briefing Unit 2 Electrical Generation – 1183 MWe
C.3 – Unit 2 FEI Conclusions
C.3.1 - NRC AIT Report - Operational Differences between U2/3 - The result of the independent NRC thermal-hydraulic analysis indicated that differences in the actual operation between units and/or individual steam generators had an insignificant impact on the results and in fact, the team did not identify any changes in steam velocities or void fractions that could attribute to the differences in tube wear between the units or steam generators.
C.3.2 - SCE Unit 2 Restart Report Enclosure 2 Conclusions - Because of the similarities in design between the Unit 2 and 3 RSGs, it was concluded that FEI in the in-plane direction was also the cause of the TTW in Unit 2.
C.3.3 - SCE U2 FEI SONGS RCE Team Anonymous Member Conclusions – FEI did not occur in Unit 2
C.3.4 - Westinghouse OA Conclusions: (a) An evaluation of the tube-to-tube wear reported in two tubes in SG 2E089 showed that, most likely, the wear did not result from in-plane vibration of the tubes since all available eddy current data clearly support the analytical results that in-plane vibration could not have occurred in these tubes, and (b) Operational data – Westinghouse ATHOS Model shows no operational differences in Units 2 & 3 (void fraction ~99.6%) and then Westinghouse says in (a) above that FEI did not occur in Unit 2. Westinghouse is contradicting its own statement.
C.3.5 - AREVA OA Conclusions - Based on the extremely comprehensive evaluation of both Units, supplemented by thermal hydraulic and FIV analysis, assuming, a priori, that TTW via in-plane fluid-elastic instability cannot develop in Unit 2 would be inappropriate.
C.3.6 - John Large States, “I note here that there are three clear conflicts of findings between the OAs: From AREVA that AVB-to-tube and TTW result from in-plane FEI, contrasted to Westinghouse that there is no in-plane FEI but most probably it was out-of-plane FEI, and from MHI that certain AVB-to-tube wear results in the absence of in-plane FEI from just turbulent flow. My opinion is that such conflicting disagreement over the cause of TTW reflects poorly on the depth of understanding of the crucially important FEI issue by each of these SCE consultants and the designer/manufacturer of the RSGs.”
C.3.7 - DAB Safety Team Conclusions – Due to higher SG pressure (Range 863 – 942 psi) and lower thermal megawatts as compared to Unit 3, FEI did not occur in Unit 2. This is consistent with the position of RCE Team Anonymous Member. The NRC AIT Report, SCE, Westinghouse, MHI, SCE chosen “Independent” Experts and AREVA conclusions on Unit 2 FEI are inconsistent, confusing and inconclusive. The NRC AIT Report, SCE and AREVA conclusions on Unit 2 FEI are unacceptable.
C.3.8 - The NRC San Onofre Special Review Panel should direct other branches within the NRC (NRC-RES and/or the ACRS) to review the above data without any prior “turf” bias and present their findings to the public for review and comment prior to any restart decision being made by the NRC.
4 – Unit 2 Main Steam Line Break (MSLB) Scenario: The most severe design basis accident to meet the San Onofre Unit 2 TS 184.108.40.206.b.1 steam generator structural integrity is a MSLB at the first weld outside containment. This assumption minimizes the flow resistance between the break and the affected SG and maximizes the mass & energy (M&E) release. The analyses focus on M&E releases at licensed Rated Thermal Power (RTP or 100% Power). The outside containment case includes the assumption that the main steam isolation valve (MSIV) in the steam line with the least flow resistance fails to close following the main steam isolation signal (MSIS). This assumption maximizes the M&E release during a MSLB outside of the containment. Super-heating within the SG initiates upon U-tube uncovery as specified in the NRC Information Notice 84-90. The turbine stop valves are assumed to close instantaneously at the time of the reactor trip. This assumption is conservative for a MSLB event because the entire steam inventory at the time of reactor trip is assumed to be forced out the break in 300 seconds (5 minutes). No Operator action outside Control Room can be assumed, if it takes less than 30 minutes. Westinghouse states, “it should be understood that there is more mass of secondary coolant in the steam generator at no load than at full power. Therefore, no load is the worst case for steam line break analyses.”
Definitions of NRC Response Columns [Source: NRC Inspection Manual Chapter 0305]
Column I – All Assessment Inputs (Performance Indicators (PIs) and Inspection Findings) Green; Cornerstone Objectives Fully Met
Column II – One or Two White Inputs (in different cornerstones) in a Strategic Performance Area; Cornerstone Objectives Fully Met.
Column III – One Degraded Cornerstone (2 White Inputs or 1 Yellow Input) or any 3 White Inputs in a Strategic Performance Area; Cornerstone Objectives Met with Moderate Degradation in Safety Performance
Column IV – Repetitive Degraded Cornerstone, Multiple Degraded Cornerstones, Multiple Yellow Inputs, or 1 Red Input; Cornerstone Objectives Met with Longstanding Issues or Significant Degradation in Safety Performance
Column V. Overall Unacceptable Performance; Plants Not Permitted to Operate Within this Band, Unacceptable Margin to Safety Licensed Activities
Column V. Overall Unacceptable Performance; Plants Not Permitted to Operate Within this Band, Unacceptable Margin to Safety Licensed Activities
Definitions, Abbreviations and Acronyms
· 10 CFR 50.59 Safety Evaluation - Section 50.59 of Title 10 of the Code of Federal Regulations (10 CFR 50.59) defines the conditions under which reactor licensees may make changes to their facilities or make changes to procedures or to conduct tests or experiments without prior NRC approval. In general, such changes, tests, or experiments may be carried out unless they would involve a change to the technical specifications or an unreviewed safety question (as defined in 50.59(a)(2)). [Source: World Wide Web]
· AIT: NRC’s Augmented Inspection Team
· AREVA: Nuclear engineering firm owned by French Atomic Energy Commission
· Arnold "Arnie" Gundersen is chief engineer of energy consulting company Fairewinds Associates and a former nuclear power industry executive, and who has questioned the safety of the Westinghouse AP1000, a proposed third-generation nuclear reactor. Gundersen has also expressed concerns about the operation of the Vermont Yankee Nuclear Power Plant. He served as an expert witness in the investigation of the Three Mile Island accident. Source: World Wide Web]
· ATHOS is a three-dimensional computational fluid dynamics (CFD) code for analyzing steam generator (SG) thermal-hydraulic (TH) performance characteristics [Source: Westinghouse OA]
· AVB: Anti Vibration Bar
· CFR: Code of Federal Regulations
· CPUC: California Public Utilities Commission
· DBA: Design Basis Accident
· ECT: Eddy Current Testing
· EIX – Edison International
· EPRI: Electric Power Research Institute
· FEI: Fluid Elastic Instability is a phenomenon where the tubes vibrate with increasingly larger amplitudes due to the fluid effective flow velocity exceeding its specific limit (critical velocity) for a given tube and its supporting conditions and a given thermal hydraulic environment. This occurs when the amount of energy imparted on the tube by the fluid is greater than the amount of energy that the tube can dissipate back to the fluid and to the supports, because of lack of squeeze film damping, nucleate boiling or presence of localized regions in U-Tube bundles with high vapor fractions (e.g., >99.6%, steam voids, steam dry-outs, etc.) [Source: MHI/DAB]
· FIRV: Flow-Induced Random Vibrations is a phenomenon where the tubes vibrate due to forces created by turbulent flow as a result of fluid velocity and density fluctuations. Vibration amplitudes due to random vibration are generally small (smaller than those due to tube fluid-elastic instability). [Source: MHI]
· FSAR: Final Safety Analysis Report
· FSM: Fluid elastic Stability Margin
· FWLB: Feed-Water Line Break
· John Large is a Chartered Engineer, a Consulting Engineer, a Fellow of the Institution of Mechanical Engineers, Graduate Member of the Institution of Civil Engineers and Fellow of the Royal Society of Arts. [World Wide Web]
· LOCA: Loss Of Coolant Accident
· MHI: Mitsubishi Heavy Industry
· MSIV: Main Steam (line) Isolation Valve
· MSLB: Main Steam Line Break
· MWt: Mega-Watts Thermal
· NOPD: Normal Operating Pressure Differential
· NRC: Nuclear Regulatory Commission
· NRC Reasonable Assurance is the recognition that “adequate protective measures can and will be taken in the event of a radiological emergency.” Reasonable assurance is based on licensees complying with NRC regulations and guidance, as well as licensees and offsite response organizations demonstrating that they can effectively implement emergency plans and procedures during periodic evaluated exercises. [Source: www.nrc.gov]
· NRC San Onofre Special Review Panel - The NRC has established a special panel to coordinate the agency’s evaluation of Southern California Edison Co.’s proposed plan for restarting its Unit 2 reactor and ensuring that the root causes of problems with the plant’s steam generators are identified, and addressed. [Source: US NRC Blog]
· NRR: NRC’s Office of Nuclear Reactor Regulations
· OD: Outer Diameter
· P/D: Pitch to Diameter ratio
· OSG: Original Steam Generator
· RCE: Root Cause Evaluation
· RCPB: Reactor Coolant Pressure Boundary
· RCS: Reactor Coolant System
· RSG: Replacement Steam Generator
· RWST: Refueling Water Storage Tank
· SCE: Southern California Edison
· SG: Steam Generator
· SGTR: Steam Generator Tube Rupture
· SM: Stability Ratio
· SONGS: San Onofre Nuclear (Waste) Generating Station (alternate abbreviation: SONWGS)
· TSP: Tube Support Plate
· TTS: Top-of-Tube Sheet
· TTW: Tube-to-Tube Wear
· TW: Tube Wear
· Westinghouse - Westinghouse Electric Company provides fuel, services, technology, plant design, and equipment for the commercial nuclear electric power industry.
The DAB Safety Team: Don, Ace and a BATTERY of safety-conscious San Onofre insiders plus industry experts from around the world who wish to remain anonymous. These volunteers assist the DAB Safety Team by sharing knowledge, opinions and insight but are not responsible for the contents of the DAB Safety Team's reports. We continue to work together as a Safety Team to prepare additional: DAB Safety Team Documents, which explain in detail why a SONGS restart is unsafe at any power level without a Full/Thorough/Transparent NRC 50.90 License Amendment and Evidentiary Public Hearings. For more information from The DAB Safety Team, please visit the link above.
Our Mission: To prevent a Trillion Dollar Eco-Disaster like Fukushima, from happening in the USA.
Copyright January 28, 2013 by The DAB Safety Team. All rights reserved. This material may not be published, broadcast or redistributed without crediting the DAB Safety Team. The contents cannot be altered without the Written Permission of the DAB Safety Team Leader and/or the DAB Safety Team’s Attorney