ML13112B041

LTR-13-0346 - Vinod Arora Email San Onofre Nuclear Generating Station

ML13112B041
Person / Time
Site:	San Onofre
Issue date:	04/13/2013
From:	Arora V - No Known Affiliation
To:	Macfarlane A, Borchardt R, Leeds E NRC/Chairman, NRC/EDO, Office of Nuclear Reactor Regulation
References
LTR-13-0346
Download: ML13112B041 (91)

Text

OFFICE OF THE SECRETARY Mon, Apr 22, 2013 17:14 Page No: 1 CORRESPONDENCE CONTROL TICKET PAPER NUMBER: LTR-13-0346 LOGGING DATE: 04/13/2013 ACTION OFFICE: EDO/OPA AUTHOR: Arora V AFFILIATION:

ADDRESSEE: Macfarlane A M

SUBJECT:

LTR-13-0346 - E-mails Vinod Arora re: San Onofre Nuclear Generating Station ACTION: Appropriate DISTRIBUTION:

LETTER DATE: 04/13/2013 ACKNOWLEDGED: Yes SPECIAL HANDLING:

NOTES: Multiple e-mails (8) from 4/9 - 4/13 FILE LOCATION: ADAMS DUE DATE: DATE SIGNED:

Joosten, Sandy From: Vinod Arora [vinnie48in@gmail.com]

Sent: Tuesday, April 09, 20131:00 AM To: CHAIRMAN Resource; Borchardt, Bill; Leeds, Eric; Howell, Art; Dorman, Dan; Benney, Brian; Hall, Randy

Subject:

San Onofre NRC/SCE/MHl/Public Awareness Series - by Hahn Baba HAHN Baba April 9. 2013 at 12:45 am Your comment is awaiting moderation.

Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC posting this blog. San Onofre NRC/SCE/MHIIPublic Awareness Series by Hahn Baba NRC, INPO, CPUC, NEI and Scientists expect SCE to supply safe and reliable power at a reasonable cost and not conduct unsafe experiments at the expense of public safety and charge ratepayers for its mistakes.

San Onofre Restart quoting Albert Einstein, "Insanity: doing the same thing over and over again and expecting different results."

To find the true root cause, first analyze the entire Cycle 16 operational differences between Units 2 & 3 and then look at tube-to-tube clearances, tall tubes and anti- FEI AVB design.

Root Causes are defined as the basic reasons (e.g., hardware, process, or human performance) for a problem, which if corrected, will prevent recurrence of that problem.

MHI Root Cause: Insufficient programmatic requirement to assure effective A VB contact force to prevent in plane fluid elastic instability and random vibration and subsequent wear under high localized thermal-hydraulic conditions (steam quality (void fraction), flow velocity and hydro-dynamic pressure).

HAHN Baba Rebuttal: MHI Answer is Incorrect. Contact force is the force in which an object comes in contact with another object. Examples are pushing a car up a hill or kicking a ball or pushing a desk across a room are some of the everyday examples where contact forces are at work. In the first case the force is continuously applied by the person on the car, while in the second case the force is delivered in a short impulse. The most common instances of this include friction, normal force, and tension. According to forces, contact force may also be described as the push experienced when two objects are pressed together. The MHI designed A VBs had zero contact forces in Unit 3 to prevent in-plane fluid elastic instability and subsequent wear under high localized thermal-hydraulic conditions (steam quality (void fraction) and flow velocity). Large u-bends were moving with large amplitudes in the in-plane direction without any contact forces imposed by out-of the plane restraints. The in-plane vibration associated with the wear observed in the Unit 3 RSGs occurred because all of the out-of-plane A VB supports were inactive by design in the in-plane direction. The Unit 3 tube-to-AVB contact forces on the TTW tubes were Zero, that is why did not restrain the tubes in the in-plane direction (Like a Sports Car moving with very high speed in freeway express toll lanes passing by a Stalled Police Car). In plane fluid elastic instability did not happen in Unit 2 because of operational differences, so therefore double contact forces and better supports is conjecture in Unit 2 and a pre-planned and ill-conceived SCE reason to justify restart of Unsafe Unit 2. The baseless contact force theory based on hideous statistical data and manufacturing simulations capable of stopping super express velocity induced in-plane vibration is refuted based on in-depth review of Speculative and Incomplete SCE Root Cause Evaluation, Dr. Pettigrew's 2006 Research Paper, Westinghouse, AREVA, John Large and earlier version ofM~I Reports.

1

NRC AIT Team SCE Root Cause: The combination of unpredicted, adverse thermal hydraulic conditions and insufficient contact forces in the upper tube bundle caused a phenomenon called "fluid~elastic instability" which was a significant contributor to the tube to tube wear resulting in the tube leak. The team concluded that the differences in severity of the tube~to~tube wear between Unit 2 and Unit 3 may be related to the changes to the manufacturing/fabrication of the tubes and other components which may have resulted in increased clearance between the anti~vibration bars and the tubes; (3) Due to modeling errors, the SONGS replacement generators were not designed with adequate thermal hydraulic margin to preclude the onset offluid~elastic instability.

HAHN Baba Rebuttal: Except incorrect answer on contact forces described above, NRC/SCE cause evaluation on SCE created adverse thermal-hydraulic conditions and Mitsubishi faulty computer modeling in Unit 3 is correct. SCE Engineers were running Unit 3 in Test Mode with higher steam flows than Unit 2 to check the improvements in Tube~to-AVB support contact forces. But like Earthquake and Tsunami in Fukushima and Fire in Chernobyl, the misunderstood, well intentioned experiment destroyed Unit 3. First rule of thumb, like Dr.

Pettigrew says, "Avoid in-plane and random vibrations by keeping velocities below 20 feet/sec (SONGS in plane velocities> 56 feet/sec, out~of~plane velocities> 28 feet/sec). Second rule, "Operate SG at higher pressures (> 900 psi) and circulation ratios> 4 to keep the void fractions less than 98.5%. Third Rule," Ask Westinghouse/Combustion Engineering how to design anti-vibration bars, which are capable of preventing fluid elastic instability." Fourth Rule, "ATHOS Models cannot calculate in-plane velocities. Therefore, do not operate SGs in unchartered Territory. "

HAHN Baba Contributing Causes: Too many adverse Design Changes to produce more megawatts, Adverse Operational Parameters, and Human Performance Errors (Lack of Critical Questioning & Investigative Attitude, Lack of Solid Teamwork & Alignment between SCE/MHI Team, Lack of Adequate research and Industry Benchmarking (e.g., NUREG-1841, Palo Verde, etc.) complacence, time pressure)

HAHN Baba Root Cause: Nuclear Safety was compromised Lack of Critical Questioning & Investigative Attitude and undermined by Profits Motivations Dr. Pettigrew's Advise*: To prevent the adverse effects of fluid elastic instability and flow-induced random vibrations, need Solid Teamwork & Alignment between Designer & Manufacturer.

• World's Foremost Renowned and Canadian Research Scientist Professeur Titulaire, Michel J. Pettigrew advise for the last 40 years in a 1976 address to the Canadian Atomic Energy Commission, "Most flow-induced vibration problems, which can be avoided provided that nuclear components are properly analysed at the design stage and that the analyses are supported by adequate testing and development work when required." SCEIMHI A VB Design Team in 2005 rejected recommendations to reduce high void fractions, which caused fluid elastic instability in Unit 3. The recommendations were rejected by MHI/SCE Team, because it would have cut down the profits due to less electricity production, cost more money to implement changes discussed in MHI Root Cause Evaluation Report, delayed the fabrication and installation process and Triggered a Lengthy NRC 10CFR 50.90 License Amendment and Public Hearing Process. SCE/MHI subverted intentionally the regulatory process. That is what Barbara Boxer was saying.

Here are more quotes from SONGS Insiders:

I. To the best of my recollection, the Root Cause Member told me that Unit 2 was running at higher pressures than Unit 3, that is why Unit 2 did not experience FE!. He had a 2006 paper with him published in 2006 by Dr.

Pettigrew in his hand, which warned about the ineffectiveness of the flat bars to prevent fluid elastic instability.

He was researching on curved bars, bars with springs, which could be attached to the tubes to prevent in-plane vibrations and repair the RSGs. What the Root Cause Member said matches with SONGS Procedures, Plant Briefing Sheets and NRC AIT Report Data.

2

2. The Root cause Team leader told to a friend of mine, "I wish that SCE engineers would have made these design changes one at a time and tested them instead of making all the changes at one time."

3. One of the very highly placed SCE Manager and Corporate Emergency Director (Now retired) told me that all these changes were made without much thought and analysis, which consisted of the substitution of Inconel 690 for Inconel 600 as the tube material. Inconel 690 is more resistant to corrosion than Inconel 600. However, Inconel 690 has a thermal conductivity approximately 10% less than that of Inconel 600. The requirement that the SG's thermal performance be maintained, in conjunction with maintaining a specified tube plugging margin, SCE told MHI for increasing the tube bundle heat transfer surface area from 105,000 ft2 to 116,100 ft2 (an 11 %

increase).

4. One of the very highly placed SCE Manager was shaking his head, when he told me, "I wish that SCE Engineers would have duplicated the Palo Verde Replacement Steam Generators and we would not be experiencing this embarrassing day. Combustion Engineering not only designed and replaced six Palo Verde steam generators with considerable improvements and higher thermal megawatt, but solved the problem with the original steam generators." Please note that San Onofre and Palo Verde Original Steam Generators were designed and fabricated by Combustion Engineering, but Palo Verde steam generators are largest in the world.

The Palo Verde Replacement Steam Generators are running fine for the last 10 years without any plugged tubes and San Onofre, everybody knows the story. Now the question is that SCE owns 20% ofthe share ofthe Palo Verde and how come SCE Engineers did not contact their counterparts Answers in the Next Update 3

Joosten, Sandy From: Vinod Arora [vinnie48in@gmaiLcom]

Sent: Tuesday, April 09, 20137:58 PM To: CHAIRMAN Resource; Borchardt, Bill; Leeds, Eric; Howell, Art; Dorman, Dan; Benney, Brian; Hall, Randy; R4ALLEGATION Resource; Lantz, Ryan

Subject:

San Onofre NRC/SCE/MHIIPublic Awareness Series (1st of 2 emails)

Attachments: Attachment 2Mdocx.docx; Attachment 3.docx; Attachment 4.docx; Attachment 5 - FATIGUE DAMAGE TO SONGS STEAM GENERATORS.docx Honorable Dr. MacfarlanelMr. BorchardtlMr. Leeds and Brilliant NRC Staff - FYI - Valuable Information for San Onofre, Sincerely - Vinnie Arora, PE, CEO/President, AVP Arora International I~c:(No~-Profit EngineeringC0I"poration)("ple~e see separate eIIlails for Attachment l(lIld Dr. Jorarp Hcipenfeld's Testimony >andSilmmaryReports.fi:leo. with* California Public Utilities Commission due to NRC Server Size Limitations Part 1)

Please excuse me for any computer or human performance grammatical or spelling errors Life is a unique opportunity to serve the society. Society needs Energy, which is safe, economical and reliable.

Every form of Energy has drawbacks and risks. SCE is responsible for safety, economics and reliability of Unit 2.

NRC, INPO, CPUC, NEI and Scientists expect SCE to supply safe and reliable power at a reasonable cost and not conduct unsafe experiments at the expense of public safety and charge ratepayers for its mistakes.

San Onofre Restart quoting Albert Einstein, "Insanity: doing the same thing over and over again and expecting different results."

NRC Top SG Expert states (February 2013), "With multiple tube ruptures, you'd have an earlier plant transient and you'd be able to identifY the tube or the generator quicker, safety systems would react; so multiple tube ruptures would challenge the operators in a different way. But we have studied that from a risk perspective and we chose not to take regulatory action or regulatory action wasn't necessary. So in some aspects the operators were benefitted by automatic systems and easier diagnosis, but the timing would create another challenge for them, so .."

Circumferentially cracked tubes can rupture without notice at any time during Unit 2 reduced steady state 70%

power operations, anticipated operational transients and main steam line breaks. The additional stresses and jet impingement loads may cause other tubes to rupture and cut into two pieces in a matter of minutes. NRC Rules governing reactor operations simply do not contemplate cascading tube ruptures. Therefore, San Onofre emergency core cooling systems are not designed to prevent a core meltdown if a number of tubes rupture at the same time. Therefore, In accordance with NRC Fukushima Task Force Lessons Learnt, Dr. Joram Hopenfeld's Analysis and observation of SONGS Operators Poor Performance/Equipment Maintenance/Reliability for the last six years, SONGS Operators and emergency core cooling systems are not capable of preventing a core meltdown caused by multiple cracked tube ruptures in defectively designed and degraded unit 2 SG caused by fluid elastic tube-to-tube wear and undetected high cycle thermal fatigue cracks.

Circumferential cracks are more serious than axial cracks because of concerns for double-ended rupture of steam generator tubes and consequent large leaks. In addition circumferential cracks are considered more difficult to detect and accurately size by nondestructive evaluation techniques like X-ray, (local) ultrasonic, and eddy current based bobbin coil probes, mechanically rotating pancake coil (RPC), etc., which have been used in SONGS Tube Inspections for Unit 2. Circular TE and TM, transmit-receive eddy current array probe C-3 and other specialized radiographic probes capable of detecting sub-surface cracks have not been used in the 170,000 1

SONGS Tube Inspections for Unit 2 due to cost and time considerations. None of the SCE Global Experts have addressed the combined synergic effects of tube-to-tube wear and high cycle thermal fatigue cracks in their voluminous 2000 page documents. There are basic errors and blunders in Westinghouse, MHI, SCE, Intertek, AREVA and NRC AIT Reports. The combined effects of tube-to-tube wear and high cycle thermal fatigue

<:;racks have been witnessed by sudden tube ruptures in North Ana in 1987, MHI SG in Mihama, Japan in 1991, three tube leakages in French SGs between 2004 through 2006, 20 tube ruptures/leakages in SGs between 1980 2000 in USA, and SONGs 3 in 2012. Because of basic SCEIMHI mistakes, continued cover-ups and subverting the regulatory process since 2004, ratepayers have lost several Billion Dollars and Public Safety has been compromised. If SCE is so confident and conservative about safety and prudent actions of their Global Experts, provide our experts with the Units 2 & 3 Cycle 16 notarized operational, contact force, tube fatigue analysis and tube Inspection data and we will certify beyond NRC analysis whether SCE and their Global Experts are right or wrong about safe restart of Unit 27 True' San Onofre Root Cause: ~uclear Safety was compromised by lack of Critic~lquesti()ning

& .Investigative AttiUldebySCElMHland undermined* by Profits Moth/ati9ns bySCr::

Attacnment1 -SONGS Unit 2&:39peration~IData! TubeWear~tal Dr.e~ttigreW'sCurVe,Heat trt:19sf~.Ql,I. ~~'~llq MHI SG lnfo~l]l~tipn (afi9*lllr~s & 2m~bles -Separate** Emaitdueto NRC. Server SiZe Limitations)

Attacllment2 . . *Unif2 Anti-Vibration Bar Contact .Force FE I Curve J\tta!ChTTH~nt3 - Flow-inducedvihration of nuclear steam generator U-tubes in two-phase flow (2011 P~er)

Atl~~llmeJ'1t4:'!G()RlpariSon ()f~it~ubishi$tet:1rneenerato.r~(SanQnpfrEr -- Ext~Rlely Talltube~f<)r

~~~siv~l-Ieat~rrflnsf~r(Equwa:lenttpAct<<ltionof 7()Q* New:1"ubes ~*See NRRRAI#13SCE .

Respon~e)aridHigh* Steam*' Flows caused FElinSQNGS3)

AttacHmentS 'C"MHI*versus* Dr:* Jora'tn.*Hopenfeld's Calculations CurVe Dr. Jorarn Rbpemfeld'sTestimoflY and Summary Reports filed with California PublicUtUities Commission- SepClrate.EmailduetoNRCServer Size Limitations)

Root Causes are defined as the basic reasons (e.g., hardware, process, or human performance) for a problem, which if corrected, will prevent recurrence of that problem.

MHI Root Cause: Insufficient programmatiC requirement to assure effective AVB contact force to prevent in-plane fluid elastic instability and random vibration and subsequent wear under high localized thermal-hydraulic conditions (steam quality (void fraction), flow velocity and hydro-dynamic pressure).

Rebuttal: MHI Answer is Incorrect. Contact force is the force in which an object comes in contact with another object. Examples are pushing a car up a hill or kicking a ball or pushing a desk across a room are some of the everyday examples where contact forces are at work. In the first case the force is continuously applied by the person on the car, while in the second case the force is delivered in a short impulse. The most common instances of this include friction, normal force, and tension.

According to forces, contact force may also be described as the push experienced when two objects are pressed together. The MHI designed AVBs had zero contact forces in Unit 3 to prevent in-plane fluid elastic instability and subsequent wear under high localized thermal-hydraulic conditions (steam quality (void fraction) and flow velocity). Large u-bends were moving with large amplitudes in the in plane direction without any contact forces imposed by out-of the plane restraints. The in-plane vibration associated with the wear observed in the Unit 3 RSGs occurred because all of the out-of plane AVB supports were inactive by design in the in-plane direction. The Unit 3 tube-to-AVB contact 2

forces on the TTW tubes were Zero, that is why did not restrain the tubes in the in-plane direction (Like a Sports Car moving with very high speed in freeway express toll lanes passing by a Stalled Police Car). In-plane fluid elastic instability did not happen in Unit 2 because of operational differences, so therefore double contact forces and better supports is conjecture in Unit 2 and a pre planned and ill-conceived SCE reason to justify restart of Unsafe Unit 2. The baseless contact force theory based on hideous statistical data and manufacturing simulations capable of stopping super express velocity induced in-plane vibration is refuted based on in-depth review of Speculative and Incomplete SCE Root Cause Evaluation, Dr. Pettigrew's 2006 Research Paper, Westinghouse, AREVA, John Large and earlier version of MHI Reports.

NRC AIT Team SCE Root Cause: The combination of unpredicted, adverse thermal hydraulic conditions and insufficient contact forces in the upper tube bundle caused a phenomenon called "fluid-elastic instability" which was a significant contributor to the tube to tube wear resulting in the tube leak. The team concluded that the differences in severity of the tube-to-tube wear between Unit 2 and Unit 3 may be related to the changes to the manufacturing/fabrication of the tubes and other components which may have resulted in increased clearance between the anti-vibration bars and the tubes; (3) Due to modeling errors, the SONGS replacement generators were not designed with adequate thermal hydraulic margin to preclude the onset of fluid-elastic instability.

Rebuttal: Except contact forces described above, NRC/SCE cause evaluation on SCE created adverse thermal-hydraulic conditions and Mitsubishi faulty computer modeling in Unit 3 is correct.

SCE Engineers were running Unit 3 in Test Mode with higher steam flows to check the improvements in Tube-to-AVB support contact forces. But like earthquake and Tsunami in Fukushima and Fire in Chernobyl, the misunderstood experiment destroyed Unit 3. First rule of thumb, like Dr. Pettigrew says, "Avoid in-plane and random vibrations by keeping velocities below 20 feet/sec (SONGS in plane velocities> 56 feet/sec, out-of-plane velocities> 28 feet/sec). Second rule, "Operate SG at higher pressures (> 900 psi) and circulation ratios> 4 to keep the void fractions less than 98.5%.

Third Rule," Ask Westinghouse/Combustion Engineering how to design anti-vibration bars, which are capable of preventing fluid elastic instability." Fourth Rule, "ATHOS Models cannot calculate in-plane velocities."

Contributing Causes: Too many adverse Design Changes to produce more megawatts, Adverse Operational Parameters, and Human Performance Errors (Lack of Critical Questioning & Investigative Attitude, Lack of Solid Teamwork & Alignment between SCE/MHI Team, Lack of Adequate research and Industry Benchmarking (e.g., NUREG-1841, Palo Verde, etc.) complacence, time pressure)

Dr. Pettigrew's Advise*: To prevent the adverse effects of fluid elastic instability and flow-induced random vibrations, need Solid Teamwork & Alignment between Designer & Manufacturer.

• World's Foremost Renowned and Canadian Research Scientist Professeur Titulaire, Michel J.

Pettigrew advise for the last 40 years in a 1976 address to the Canadian Atomic Energy Commission, "Most flow-induced vibration problems. which can be avoided provided that nuclear components are properly analysed at the design stage and that the analyses are supported by adequate testing and development work when required." SCE/MHI AVB Design Team in 2005 rejected recommendations to reduce high void fractions, which caused fluid elastic instability in Unit 3. The recommendations were rejected by MHI/SCE Team, because it would have cut down the profits due to less electricity production, cost more money to implement changes discussed in MHI Root Cause Evaluation Report.

delayed the fabrication and installation process and Triggered a Lengthy NRC 10CFR 50.90 License Amendment and Public Hearing Process. SCE/MHI subverted intentionally the regulatory process.

That is what Barbara Boxer was saying.

3

Here are more quotes from SONGS Insiders:

1. To the best of my recollection, the Root Cause Member told me that Unit 2 was running at higher pressures than Unit 3, that is why Unit 2 did not experience FEI. He had a 2006 paper with him published in 2006 by Dr. Pettigrew in his hand, which warned about the ineffectiveness of the flat bars to prevent fluid elastic instability. He was researching on curved bars, bars with springs, which could be attached to the tubes to prevent in-plane vibrations and repair the RSGs. What the Root Cause Member said matches with SONGS Procedures, Plant Briefing Sheets and NRC AIT Report Data.

2. The Root cause Team leader told to a friend of mine, "I wish that SCE engineers would have made these design changes one at a time and tested them instead of making all the changes at one time."

3. One of the very highly placed SCE Manager and Corporate Emergency Director (Now retired) told me that all these changes were made without much thought and analysis, which consisted of the substitution of Inconel 690 for Inconel 600 as the tube material. Inconel 690 is more resistant to corrosion than Inconel 600. However, Inconel 690 has a thermal conductivity approximately 10% less than that of Inconel600. The requirement that the SG's thermal performance be maintained, in conjunction with maintaining a specified tube plugging margin, SCE told MHI for increasing the tube bundle heat transfer surface area from 105,000 ft2 to 116,100 ft2 (an 11 % increase).

4. One of the very highly placed SCE Manager was shaking his head, when he told me, "I wish that SCE Engineers would have duplicated the Palo Verde Replacement Steam Generators and we would

!Jot be experiencing this embarrassing day. Combustion Engineering not only designed and replaced six Palo Verde steam generators with considerable improvements and higher thermal megawatt, but solved the problem with the original steam generators." Please note that San Onofre and Palo Verde Original Steam Generators were designed and fabricated by Combustion Engineering, but Palo Verde steam generators are largest in the world. The Palo Verde Replacement Steam Generators are running fine for the last 10 years without any plugged tubes and San Onofre, everybody knows the story. Now the question is that SCE owns 20% of the share of the Palo Verde and how come SCE Engineers did not contact their counterparts - Answers in the Next Update No Significant Hazards Consideration -Analysis of New Licen~eAmendment Dr. Joram Hopenfeld states, "A steam generator, in addition to providing a barrier to radioactivity and producing steam, has many other important functions. It is the major component in the plant that contributes to safety during transients and accidents. It provides the driving force for natural circulation and it facilitates heat removal from the reactor core during a wide range of loss of coolant accidents. Proper steam generator operation is of major safety significance and therefore any changes to its design may have potential safety consequences. Southern California Edison("SCE")

has not identified the root cause for the unusually excessive tube wears in the four steam generators

("SGs") of units 2 and 3 at San Onofre Nuclear Generating Station, SONGS. Based on my evaluation of the tube wear data and the in situ leak test results it is my opinion that restarting Units*

.' 2 and 3 would compromise public safety. The new components in the replacement steam generat()rs

("RSGs"), constituted a major change to the original SGs, this lead to vibrations and the unusual rapid tube wear. The components causing the wear would have to be replaced and the SONGS license amended before the units can be restarted. SCE and MHI did not provide any data to support their contention that the various design change options that were discussed in 2006 by the AVB Design Team would have had no significant effect on flow velocities and steam quality. SCE consultants AREVA, Westinghouse and MHI differed on the root cause of tube vibrations and none of their opinions were based on sound scientific principles. The safety consequences of operating with degraded tubes are more seriolJs than envisioned by the consultants. It is apparent that SCE focused its attention on explaining away errors in the design, fabrication and management of the 4

RSGs. There is no indication that any consideration was given to the long-term safety risk of operating Units 2 and 3, with each containing more than 1500 defects (3401 in all)."

Let us examine, why Dr. Hopenfeld, Arnie Gundersen, John Large, Professor Daniel Hirsh and many other experts are saying that it is not safe to operate at Unit 2 at 70% power, while SCE, AREVA, Westinghouse, MHI, Intertek and the NRC Augmented Inspection Team say that Unit 2 reactor operation at no more than 70 percent power will limit unusual tube wear and is safe to operate by reviewing the following questions:

Let us assess the condition of the defectively designed and degraded U2 RSGS, before we answer the following questions:

'. A steam generator, in addition to providing a barrier to radioactivity and producing steam, has many other important functions. It is the major component in the plant that contributes to safety during transients and accidents. It provides the driving force for natural circulation and it facilitates heat removal from the reactor core during a wide range of loss of coolant accidents. Proper steam generator operation is of major safety significance and therefore unanalyzed and untested design changes to SONGS RSGs have created major safety consequences as observed by damages in Units 2 & 3.

• The Root Cause determined by SCE and MHI for both Units 2 & 3 RSGs does not address the exact reason for RSG design and operational flaws. Root Cause is defined as the exact reason (e.g., hardware, process, or human performance) for a problem, which if corrected, will prevent recurrence of that problem. Therefore, SCE/MHI have not determined the exact root cause of the tube-to- tube wear in Unit 3 per CAL ACTION 1, and have not implemented actions to prevent the loss of tube-to-tube wear and demonstrated via a deterministic safety analysis that the AVB structural integrity in the Unit 2 steam generator will be maintained (e.g., collapse of AVB structure and retainer bars failure due to fluid elatic instability) due to Main Steam Line Break (e.g., Mihama, Turkey Point, Robinson), Station Blackout (Fukushima), SG Tube Ruptures ( Mihama, SONGS 3 &

20 other Incidents in US/Europe in the last 20 years) and other anticipated operational transients at power operation between 70 to 100% power .

.* Based on the evaluation of the tube wear data and the in situ leak test results, restarting Units 2 and 3 would compromise public safety. The new components in the replacement steam generators constituted a major change to the original SGs, which lead to the unusual and rapid tube wear. The components causing the wear would have to be replaced and the SONGS license amended before the units can be restarted. The above assessment also applies to SCE proposed five-month test of Unit 2 at 70% of licensed power. After correcting an error in the SCE stress calculations, the present analysis shows that because of the wear damage previously sustained by Unit 2 some tubes will be susceptible to rapid fatigue failure. The tubes will exceed their allowable fatigue life by 22 to 29%

during the next operating cycle. The risk of tube rupture increases with operating time but the MHI and Intertek analysis is not capable of quantifying it in terms of operating time. Unit 2 should not be permitted to operate until SCE provides a thorough assessment of the fretting fatigue discussed in Dr. Joram Hopenfeld's Reports.

• To meet the performance criteria as specified in the plant's technical specifications (UTS") it is necessary to relate tube defect geometry to primary/secondary side leakage. When tube degradation is caused only by thinning due to tube-to tube wear, the degree of tube damage can be assessed by modeling, pulling selective tubes for testing or by in-situ pressure testing. In addition to losing strength due to wall thinning, some of the tubes at SONGS have used up a significant fraction of their allowed fatigue life. Such damage cannot be detected by even the NRC Special Tube 5

Inspections due to time, cost, unavailability of high technology probes, contactors, and/or impossible access within the tube bundle or radiation dose limitations. These tubes are significantly susceptible to sudden ruptures without notice and/or early warnings during steady state normal operations at 70% power, Operational Transients (opening or closing of valves, scrams, loss of offsite power, moderate earthquakes, etc.) and under steam line break accidents at other reactors. The stress on the tube [SONGS Unit 3 Tube, which leaked, Row 106 Column 78, 100 percent through wall wear, length of wear - 29 inches] due to in-plane vibration calculated by MSI was 4.2 ksi and shown to be under fatigue limit (13.6 ksi)." In contrast, Dr. Hopenfeld's calculations in the attached report demonstrate that the stress concentration factor is much higher than calculated by MHI and therefore the fatigue limit of 13.6 ksi will be exceeded by at least 22% (16.7 to 17.5 ksi). Now, the ruptured tube in Mihama experienced out-of-plane FEI and according to the latest research between 2006 -2011 by Dr. Pettigrew and others, the in-plane velocities are double the out-of plane velocities, therefore, during FEI conditions, tubes can realistically experience fatigue of 16.8 ksi. This demonstrates that MHI calculations are under-conservative by a factor of 4, just like the fluid velocities calculated earlier by MHI, which led to the SONGS Unit 3 tube leak, failure of 8 tubes under MSLB testing conditions and loss of more than 35% wall thickness in almost 400 tubes.

Fatigue damage has resulted in several tube failures in US, French and Japanese steam generators, yet in spite of its importance, none of the SCE consultants included fatigue in their evaluation of restarting Unit 2.

• Between 2004 and 2006, three primary-to-secondary leaks occurred at the Cruas NPP: unit 1 in February 2004 and unit 4 in November 2005 and February 2006. The three leaks were all the result of a circumferential crack in the tube at the location where the tube passes through the uppermost tube support plate (TSP #8). Analyses carried out for the last two events, resulted in them being attributed to high cycle fatigue of steam generator tubes due to flow-induced vibration.

• On February 9th, 1991, a heat transfer tube (SG tube) in a Mitsubishi steam generator of the No.2 pressurized water reactor at the Mihama nuclear power station of the Kansai Electric Power Company, broke off during a rated output operation. As a result, about 55 tons of radioactive primary cooling water leaked out from the SG tube into the secondary cooling loop, and the reactor was scrammed by operation of the ECCS (Emergency Core Cooling System). The failure of the SG tube was caused by fretting fatigue resulting from contact of the SG tube with the supporting plate for the SG tubes, because the AVB, which functions to prevent flow-induced vibration, was not inserted deep enough onto the SG tubes in the steam generator. One main steam isolation valve and one pressurizer relief valve could not be operated by remote control. Therefore, the valve operation was carried out manually. The amount of steam released from the main steam relief valve to atmosphere was about 1.3 tons. The amounts of radioactive rare gas and iodine discharged to the atmosphere were about 2.3E10 and 3.4E8 becquerels, respectively. The scale of the accident was ranked "level 3" on the international nuclear events scale (lNES). Stress amplitude of the failed tube estimated based on the striation spacing was found to be in the range of around 51 to 60 MPa (8.4 ksi).

• Unit 2 RSGs have 505 plugged and/or staked tubes. Inspections reveal that there are numerous U bends in both RSGs with tube-to-tube clearances as small as 0.05 inches (SONGS RSGs Design 0.25 inches, Industry NORM 0.35-0.55 inches). The design distance between tubes on the sides at the intersection with the top TSP should be 0.250 inch plus or minus the small broached hole tolerance. The Nominal Gap between tube and AVB in SONGS RSGs built by Mitsubishi is 0.002",

while in Fort Calhoun RSGs (another US plant built by Mitsubishi) the Nominal Gap is 0.0031". The tube diameter (d) and pitch (P) Tube Index (P/d) in SONGS RSGs built by Mitsubishi is 1.33-1.433, whereas tube diameter (d) and pitch (P) Tube Index (P/d) in Arkansas Nuclear One Unit 2 RSGs built by Westinghouse is 1.518-1.672. The in-plane tube spacing at Apex in SONGS RSG is (0.298, 0.344, 0.400 inch) whereas at another plant, it is (0.442, 0.502, 0.562 inches). According to 6

Westinghouse, the actual distance may be between 0.040 and 0.120 inch (1-3 mm). According to AREVA, "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 instances where the closest approach distance is less than this value, based on field measurements using bobbin coil ECT. The bobbin probe on the 140 kHz absolute channel can detect neighboring U-bends if the separation distance is less than approximately 0.15 inches. Using a proximity signal calibration curve, the separation distance between U-bends was measured for all steam generators.

The smallest detected U-bend separation distance is close to contact. 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 separation of the U-bends in Unit 2 with TTW is 0.190 inches as measured by UT. The U-bends with the smaller separation distances are much better candidates for wear by rubbing yet do not exhibit TTW. Intertek states, "There were 4348 indications detected at AVB contact points with a maximum NDE depth of 35%TW during the end-of-cycle (EOC) 16 tube examinations. Wear at tube support plates (TSP) was also detected (364 indications) with a maximum NDE depth of 20%TW." MHI Tube Plugging Criteria states, "Tubes, which eX~libit a potential for losing their integrity during the next operating period due to progressive through-wall wear and/or susceptibility to FEI should be plugged. All tubes with ECT tube-tube wear indications in the free span section should be plugged regardless of the wear depth. Furthermore, tubes with wear indications at the AVB and TSP locations, which are similar to those on the tubes with the wear indication in the free span section, should be preventatively plugged. Tubes with Tube-AVBITSP wear equal to, or greater than, 35%

should be plugged in accordance with Technical Specifications." Approximately - 1600 tubes in Unit 2 steam generators were found with wear indications. Only -500 tubes out of these were plugged due to wear (Tube-to-Tube/AVBITSPs). That means - 1080 tubes in Unit 2 steam generators with wear indications have not been plugged. These plugged tubes will continue to wear and requires a safety analysis to demonstrate that these tubes will not cause damage to adjacent tubes during accidents and meet the present licensing requirements.

-The following very important issues have not been addressed in Unit 2 Return to Service reports.

1. Long-term Implications of restarting any of the SONGS units.

Because of the large number of defects in the tubes, there is a risk that additional tubes will have to be removed from service even if the FEI problem has been solved. Defects such as now exist (more than 3000) are known to form nucleation sites for stress corrosion (SCC) and fatigue cracks. It is important to understand that even though alloy 690 is not as prone to SCC as alloy 600 it is not completely immune to SCC. This problem will become more important as the units age because of crud build up in the tube support plates.

2. Uncertainties in the analyses of FEI Long Term Effects.

The consultants did not address uncertainties in sufficient details. Because of the complexity of the technical issues a reviewer, therefore, cannot asses of the robustness of the analysis. John Large states, "A difficulty that I have with the AREVA and, generally, with the other OAs is that whereas the results of analyses, particularly relating probability and confidence, are often resolutely stated, very little of the analytical procedures arriving at the results are open to inspection. Because of the uncertainties I very much doubt that in the present circumstances tube structural integrity could be guaranteed to satisfy the 95% probability at 50% confidence criterion but, that said, AREVA presents no substantial data that enables me to explore and possibly resolve these doubts." Because the uncertainties in predicting how fatigued tubes can propagate failures, it is impossible to assess quantitatively safety risks. It is believed that a main steam line break (MSLB) represents a bounding case. A conservative estimate of the probability of a large early release of radiation with containment 7

bypass would be 10E-4 per year for any operating cycle. Such risk exceeds NRC's safety goals.

Units 2 and 3 at SONGS have the highest risk of tube rupture related core damage than any other nuclear power plant in the USA.

John Large states, "SCE's assertion that reducing power to 70% will at the best alleviate, but not eliminate, the TTW and other modes of tube and component wear is little more than hypothesis the supporting Operational Assessments and analyses have not proven it to be otherwise. I am of the opinion that trialling this hypothesis by putting the SONGS Unit 2 back into service will, because of the uncertainties and unresolved issues involved, embrace a great deal of change, test and experiment. The terms of the Confirmatory Action Letter of March 11 2012, are versed such that to meet compliance the response of SCE via its Return to Service Report,11 must include considerable changes of conditions and procedures that are outside the reference bounds of the present FSAR this is because the physical condition of the RSGs, and the means by which this is evaluated and projected into future in-service operation, have substantially and irrevocably changed since the current FSAR was approved. The fact that SCE fails to satisfy the requirements of the CAL is neither here nor there, although it illustrates the scope and complexity of the response required. At the time of preparing the CAL, the NRC being well-versed in the failures at the San Onofre nuclear plant, surely must have known that the only satisfactory response to the CAL would indeed require considerable changes, tests and experiments to be implemented."

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

SCE's Answer: No, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Basis - Please see SCE License Amendment (www.songscommunity.com)

My Response: SCE, its Independent Experts and NRC Augmented Inspection Team investigations and their answers are incorrect and misleading like all the past actions pertaining to: (1) Inability to determine the true reasons for faulty replacement steam generator design, (2) Inability to determine the true root cause of tube-to-tube wear in Unit 3 and take corrective actions to prevent recurrence of tube-tube wear in Unit 2. Because the true causes have not been determined, that is why both San Onofre Units are shutdown, and NRC/SCE/MHI are facing so much opposition from public and Independent Safety Experts. Southern Californians were lucky, that a nuclear meltdown was averted with the Unit 3 accident, and ratepayers have already lost more than a Billion Dollars. The proposed Unit 2 reactor operation at no more than 70 percent power involves more than significant increase in the probability or consequences of a potential Steam Line Break Accident resulting in multiple steam generator tube leakages and/or ruptures than previously evaluated by SCE, their nuclear industry experts or some at the NRC. This is also a new and different type of accident applicable only to the San Onofre Unit 2 & 3 replacement steam generators because of their faulty design, which has not been previously evaluated. The proposed Unit 2 reactor operation at no more than 70 percent power involves a significant reduction in safety because of the reasons stated below.

Basis - The Unit 2 reactor operation at no more than 70 percent power affects the probability or consequences of steam generator tube rupture significantly because many of the steam generators tubes can leak/rupture under the following conditions:

Three different accident scenarios should be considered in attempting to determine accident severity when degraded tubes rupture.

(i) 70% normal Steady State power operations: Due to random variations in local vibration intensity, tube failure will be initiated in a single tube with relatively slow through the wall crack propagation.

Such failures likely would be confined to a single tube leakage lending to a may be timely detection 8

and action by operator, or after the 5 month inspection period and removal from service, if the wall thickness exceeds 35%.

(ii) Operational Transients @ 70% power operations. Next in severity are tube ruptures from operational transient (opening or closing of valves, scrams, etc). To ensure that the tubes withstand such transients, the ASME code requires that their Cumulative Usage Factor ("CUF") be less than one. When the RSGs were installed the predicted CUF was a low number because it was calculated for pristine tubes. The fact that the tubes were stressed to above their endurance limit and they vibrated with frequencies on the order of 10 to 50 HZ, means that their CUF exceeded unity. A tube with CUF >1 would be prone to a rapid crack growth and sudden rupture under operational and DBA transients.

Notes: MHI states, "The stress on the tube [SONGS Unit 3 Tube, which leaked, Row 106 Column 78, 100 percent through wall wear, length of wear - 29 inches] due to in-plane vibration was 4.2 ksi and was under fatigue limit (13.6 ksi)." In contrast, Dr. Hopenfeld's calculations demonstrate that the stress concentration factor is much higher than calculated by MHI and therefore the fatigue limit of 13.6 ksi will be exceeded by at least 22% (16.7 to 17.5 ksi). Now, the ruptured tube in Mihama experienced out-of-plane FEI and according to the latest research between 2006 -2011 by Dr.

Pettigrew and others, the in-plane velocities are double the out-of plane velocities, therefore, during FEI conditions, tubes can realistically experience fatigue of 16.8 ksi. This demonstrates that MHI calculations are under-conservative by a factor of 4, just like the fluid velocities calculated earlier by MHI, which led to the SONGS Unit 3 tube leak, failure of 8 tubes under MSLB testing conditions and loss of more than 35% wall thickness in almost 400 tubes. Now, SONGS Unit 2 RSGs have hundreds of fatigue damaged tubes, which have not been plugged and a number of U-bends with tube-to-tube clearances as low as 0.05 inches, which is almost one fifth of the design clearance of 0.25 inches. In addition, there are some active and pressurized tubes, some of which have lost wall thickness up to 28%, therefore, these tubes are just below the safety margin of 35% of the tube plugging limit. SCE has chosen not to plug these tubes as a conservative measure in their belief that these tubes will not rupture dlJring the next 5 months. In addition to losing strength due to wall thinning, some of the worn tubes at SONGS Unit 2 have used up a significant fraction of their allowed fatigue life. Such damage cannot be detected by even the NRC Special Tube Inspections due to time, cost, unavailability of high technology probes and contactors, and/or impossible access within the tube bundle or radiation dose limitations. These tubes will be significantly susceptible to sudden ruptures without notice or early warnings during steady state normal operations at 70%

power due to Operational Transients (opening or closing of valves, scrams, loss of offsite power, lifting of steam safety valves due to SG Over-pressurization, etc.) and can result in accidents like Mihama Unit 2 (Japan, 1991, INES Level 3 nuclear incident) or even leading to a nuclear meltdown.

(iii) Main Steam Line Break ("MSLB") @ 70% power operations. An MSLB could lead to the most severe consequence for operations with degraded tubes. SCE's proposed revision to Technical Specification or current licensing basis ("CLB") at 70% power would require that the plant accommodate such accidents. The MSLB accident is of particular concern, because of what happened at Unit 3. Its unprecedented tube-to-tube wear, never experienced in the 60 years of nuclear steam generator operational history, was caused in a very small section of the tube bundle due to the unanticipated creation of high dry steam (fluid elastic instability). The question is why was this condition unanticipated and how could high dry steam cause so much devastating damage in less than 11 months of operation to these tubes, which were designed to last between 40 to 60 years. Because, SCE, AREVA, Westinghouse, MHI, Intertek and NRC Augmented Inspection Team have not actually answered this important question, Dr. Hopenfeld, Arnie Gundersen, John Large, Professor Daniel Hirsh and other experts are saying that it is not safe to operate Unit 2 at 70%

power.

9

Now, I will try my best to answer the question, why it is unsafe to restart Unit 2 at 70% power?

The most severe design basis accident to meet the SONGS Unit 2 TS 5.5.2.11.b.1 steam generator structural integrity is a MSLB at the first weld outside containment. The outside containment scenario 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). 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 or 5 minutes.

The depressurization of the non-isolable steam generator would result in 100% void fractions in the degraded Unit 2 U-Tube bundle due to flashing of the feedwater into steam. This condition of ZERO Water in the steam generators would cause fluid elastic instability (FE I) flow-induced random vibrations, excessive hydrodynamic pressures and Mitsubishi Flowering Effect. The force of the flashing steam would create high-energy jets, lift loose parts and debris present inside the steam generator. These adverse effects would cut holes into already degraded tubes and create additional loading (See Note A below) on tube support plates (TSPs) due to heavy build-up of deposits on trefoil/quadrifoil-shaped holes from SG blowdown and cracked high cycle fatigue U-bend tubes not supported by an Anti-Vibration Bars (AVB). More than 500 tubes were plugged at Unit 2. Even though these tubes would not be not in service, they will continue to wear at the same rate as before. At a certain time, these tubes will break without being detected because no radioactivity will be released. The broken tubes will have relatively low natural frequency and therefore it would be prone for resonance excitation by external forces. The jets, formed at the ends of a broken tube, would cause it to whip and impact adjacent tubes thereby propagating further ruptures. These cumulative adverse conditions in all likelihood would result in a massive cascading of RSGs tube failures (tubes would excessively rattle or vibrate, hitting other tubes with violent impacts) due to extremely low tube-to-tube clearances and no in-plane effective anti-vibration bar support protection system. This Titanic and adverse effect would involve hundreds of degraded and active SG tubes along with all the inactive (plugged /unstabilized) tubes causing an undetermined amount of simultaneous tube leaks/ruptures. Under this adverse scenario, approximately 60 tons of very hot high-pressure radioactive reactor coolant would leak into the secondary system. The release of this amount of radioactive primary coolant, along with an additional approximately 200 tons of steam in the first five to fifteen minutes from a broken steam line would EXCEED the SONGS NRC approved offsite radiological release doses safety margins. So, in essence, the RSG's are loaded guns, or a nuclear accident like Fukushima, waiting to happen. Any failure under these conditions would allow .

significant amounts of radiation to escape to the atmosphere and a major Loss of Coolant Accident*

(LOCA) could easily result causing much wider radiological consequences and even a potential nuclear meltdown of the reactor.

SCE States, "A MSLB alone does not generate sufficient differential pressure to cause tube rupture.

(See Notes below). The differential pressure across the SG tubes necessary to cause a rupture will not occur if operators prevent RCS re-pressurization in accordance with Emergency Operating Instructions."

In Summary: seE DID Actions and unreliable operator actions to detect a leak and to re-pressurize the steam generators as claimed by Edison are not practical to stop a major nuclear accident in Unit 2 in progress in the first 5-15 minutes of a MSLB during the 5-month trial period.

10

Notes: This additional loading would exceed: (2) the safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1

.4 against burst applied to the design basis accident primary-to-secondary pressure differentials, and (3) significantly affect burst or collapse pressures determined and assessed in combination with the loads due to a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

SCE's planned "defense-in-depth" actions are insufficient to stop multiple tube ruptures due to a main steam line beak event.

2. Does the proposed change create the possibility of a new or different kind of an accident from an accident previously evaluated?

SCE's Answer: No, the proposed changes do not create the possibility of a new or different kind of an accident 'from an accident previously evaluated.

Basis - Please see SCE License Amendment (www.songscommunity.com)

My Response: The proposed changes create the distinct possibility of a new or different kind of an accident from an accident previously evaluated. Please see Item 1 above for details.

3. Does the proposed change involve a significant reduction in margin of safety?

SCE's Answer: No, the proposed changes do not involve a significant reduction in margin of safety.

Basis - Please see SCE License Amendment (www.soogscommunity.com)

My Response: The proposed change involves more than a significant reduction in margin of safety.

Please see Items 1 and 2 above for details.

Based on the above, I conclude that the proposed amendment involves more than a significant hazards consideration under the standards setforth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" claimed by SCE is another attempt to subvert and erode the NRC Regulatory Process. NRC Brilliant Engineers know better what to do this time. Let us wait for the NRC Intelligent Safety Answers.

11

High Void Fraction> 99.6% (Fluid Elastic Instability, Random Vibrations & Mitsubishi Flowering Effect - Anticipated Operational Transient And Main Stea ~eak Conditions)

__......_ _ _Jube-to-Tube Plus AVB Wear Depth Rate Low Void Fraction ~ 98.5% (Random Vibrations) 7-tO-AVB ontact Force o 1 2 3 4 5 00 Unit 2 Anti-Vibration Bar Contact Force FEI Curve

Flow-induced vibration of nuclear steam generator U-tubes in two-phase flow In-Cheol Chu i, 1'iII, Heung June Chung, Seungtae Lee Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong-gu, Daejeon 305-353, Republic of Korea Abstract Flow-induced vibration characteristics of a U-tube bundle were experimentally investigated in air-water two-phase flow. The test section was equipped with 39 U-tubes, simulating the innermost region of an actual steam generator. The U-tubes were made of Inconel 690 with a diameter of 19.05 mm. The horizontal region of the U-tubes had a rotated square array with a pitch of 31.11 mm and a p/d of 1.633. The U-tubes and supporting structures have almost the same prototypical geometries.

Vibration responses of six U-tubes were measured with ten 3-axis accelerometers. Two sets of experiments were performed to investigate an onset of fluid-elastic instability, damping ratio, and hydrodynamic mass of the U-tubes. The experiments were performed for a void fraction of 70 95%. The instability constant (K) of the Connors' equation for the present U-tube bundle was evaluated to be in the range of 6.5-10.5.

Research highlights

.... We examined flow-induced vibration of a U-tube bundle in two-phase flow ..... Test section had 39 U-tubes supported by full and partial egg-crates ..... Damping ratio of the U-tubes was much higher than that of cantilever tubes ..... Fluid-elastic instability constant was in the range of 6.5 10.5.

l Injection Port (Ait*w_ Mixtu,,;)

HoI leg Side h

Fig. 1. U-tube bundle test section: a front view (a) and a plane view (b).

ColUITlfl 1 Column 7 Row 1*4 Fig. 2. Sectional diagram of U-tube arrangement in the horizontal part and the location of the instrumented U-tubes. Instrumented U-tubes are marked solid.

r

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OP first mode (a), IP first mode (b), OP second mode (c), IP second mode (d). OP: out of-plane, IP: in-plane.

a 1.8 1.6 E 1.4 E

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(} 2 4 8 14 Gap Velocity, m/s Fig. 4. RMS displacements of row number 5 U-tube (a) and row number 3 U-tube (b).

Rectangular, circular, and triangular symbols represent the void fraction of 80, 85, and 90%, respectively. Solid, hollow with '+', and hollow symbols represent the OP first, IP first, and IP second mode vibrations, respectively. The critical gap velocity was determined at the intersection of two slopes corresponding to the turbulent excitation region and the fluid-elastic instability region, respectively.

a 0.4 Row 4 U-tube OP modes E 0.3 E

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~ 0.1 95 b 0.4 Row 5 U-tube OP 1st mode Fig. 5. Power spectral density functions of row number 4, column number 4 U-tube (a) and row number 5, column number 3 U-tube (b) in the OP mode vibration at the void fraction of 70-95%.

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0,03 0.00 1.B 2.1 2.4 2.7 3.0 3.3 3.6 Gap Velocity, m/s Fig. 6. Total damping ratio versus the gap velocity in the OP first mode vibration at the void fraction of 80% (a) and 90% (b). The damping ratios were measured with all the U tubes free to vibrate.

OP 1 st mode vibration Present 010 Row no.

o 1

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a NS (Pettigrew) 8 0.02 ~_ RT (Pettigrew}.._

• NS (Chung & Chll)

<> AS (Chung & Chul 65 70 75 80 65 95 100 Void Fraction, %

Fig. 7. Total damping ratios of the U-tubes in the OP first mode vibration. The damping ratios were measured with all the U-tubes free to vibrate. The total damping ratios of Pettigrew et at (1989a) and Chung and Chu (2006) are compared together.

Row 5, Colul'l'ln 3 U-Tl.Ibe 0.09

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{2 0.03 0.00 ...................t..___"- '--t..-"'--"--'--'- ""--'--..........t............

H a ro n 00 a 00 ~ 100 Void Fraction, %

Fig. 8. Effect of the supporting condition of neighboring U-tubes on the total damping ratio of the row number 5, column number 3 U-tube in the OP first mode vibration. The damping ratios were separately evaluated for two conditions that all the U-tubes were free to vibrate (all flexible) and that only the row number 5, column number 3 U-tube was free to vibrate and the others were rigidly supported (single flexible).

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¢ RS (Olung & Chu) 0.0 60 65 70 75 SO 85 90 95 100 VOid Fraction, %

Fig. 9. Hydrodynamic mass ratio of the present U-tubes in the OP first mode vibration.

The hydrodynamic masses were evaluated for the condition that all the U-tubes were free to vibrate. The hydrodynamic mass ratios of Pettigrew et al. (1989a) and Chung and Chu (2006) are compared together.

Fig. 11. Photographic observation of two-phase flow direction along the horizontal and round parts of the right half side of the U-tubes. Void fraction was 70% and gap velocity was 4.0 m/s.

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0.5 1 5 10 50 Mass Damping Parameter, 2n:~m/pd2 Fig. 12. Results of the fluid-elastic instability of the present U-tubes. Damping ratio, hydrodynamic mass, tube frequency of OP first mode vibration and effective velocity with uniform velocity distribution were used.

Fig. 10. Assumptions on the flow velocity distribution for the evaluation of the effective velocity. Illustrated is non-uniform velocity distribution that the round part of the U-bend region had two times higher velocity than the horizontal part.

san UI'IOfI'e.:lfJ Sf,j, U:iA. I:UJr :sr.;~ t1'anC8,. 1.J08I.1 :sr.;, 1:SeIgI.im, 583 tons. I 318ton,J 268 tons I (triangular pitCh tubing) (square pitch tubing) (riarOIar pitch tubing)

Heat

• Tuba support plabt .*

Please see the tall tube bundle in San Onofre Steam compared with other Mitsubishi Steam Generators. The San Onofre tall tube bundle due to high steam flows caused fluid elastic instability or high dry steam and destroyed SONGs Unit 3. By making the tube bundle tall without a 10 CFR 50.90, SCE increased the average length of 9727 tubes by 7 inches each to gain additional 7% heat transfer area equivalent to 700 new tubes to generate 120 more thermal megawatts per Steam Generator. Everybody is under the impression that SCE added only 377 tubes, but in reality, SCE added 1077 tubes including 700 tubes by making the tube bundle tall. SCE response to NRC RAI #13, states, "The RSGs have more tubes (9,727 versus 9,350) than the OSGs and a smaller value for the maximum number of plugged tubes (779 versus 2,000).

RSG tubes have a larger average heated length (729.56 in. versus 680.64 in.) than the 05G tubes. These features result in larger values for the RSG for heat transfer area, tube bundle flow area, and tube bundle water volume. This is beneficial in the short and long term for SBLOCAs, which rely upon the steam generators for RCS heat removal. "

This adverse change is 100% opposite is of What SCE is telling the NRC. Here is why,

1. Involved a more than significant increase in the probability or consequences of a steam generator tube rupture leak due to 100% tube-to-tube wear in one tube in SONGS Unit 3, failure of 8 tubes at MSLB testing pressures and loss of> 35% wall thickness in more than 300 tubes. This event was caused due to the misrepresented SCE design changes reported in RAI #13. The probability or consequences of this change were more significant than previously evaluated in the NRC Approved FSAR?

2. The proposed License Amendment change running Unit 2 at 70% power creates the possibility of a concurrent steam line break and consequential cascading tube ruptures due to fluid elasticity or high dry steam and jet impingement forces., This is a a new or different kind of an accident from an accident previously evaluated?

3. The proposed License Amendment change running Unit 2 at 70% power change involves more than a significant reduction in margin of safety because of the high potential of a nuclear meltdown beyond operator control and offsite releases exceeding the limits previously analyzed in the FSAR due to the possibility of a concurrent steam line break and consequential cascading tube ruptures?

FATIGUE DAMAGE TO SONGS STEAM GENERATORS J. Hopenfeld SCE/MHI made a mistake in their stress analysis, which directly impacts the safety of restarting Unit 2. When the error is corrected, the result clearly shows that Unit 2 has already used up its allowed fatigue life and is not fit for service any longer. This means that if Unit 2 is restarted at any power level an abrupt pressure change such as inadvertent closing or opening of a valve or a steam line break could lead to a sudden tube ruptures. The ASME code and t\IRC regulations do not permit safety components to operate when their fatigue life has been exhausted.

The source of MHI's error resulted from how they calculated the increase in the local stress at geometrical discontinuities (notches), which are formed when two metal surfaces come in contact during vibration. Since the worn surfaces of the tubes inside the steam generators cannot be seen, MHI made two key assumptions, which are inconsistent with the observation that both the tube and the supporting bar are worn into each other. First, MHI assumed that the ASME endurance limit could be applied directly to the notched tube surfaces. Since it is commonly known that surface roughness significantly reduces fatigue life and since the ASME data is for smooth polished surfaces, this assumption would underestimate the amount of fatigue damage.

Second, when using the Peterson chart, MHI assumed unrealistically large fillet radius and consequently derived a low concentration stress factor. Large radii would decrease the local stress and cause the tube to fail at a higher stress thereby increasing its fatigue life. Only by using these two, arbitrary non-conservative, assumptions was MHI able to conclude that Unit 2 did not suffer any fatigue damage.

As depicted in the MHI drawings the support bar and the tube form a sharp discontinuity at the contacting surface, therefore the appropriate geometry for calculating the stress concentration is an abrupt geometry change (very small radii), not a large radius shoulder fillet that was assumed by MHI. When a correction is made to account for the sharp notch, the corrected stress indicates (see Figure 1 below) that the tubes have used up their fatigue life during the first cycle of operation. Structures with sharp notches can fail catastrophically when subjected to high cycle vibrations. (MHI redacted their assumption so the exact value of the radius they used is unknown.)

The loss of fatigue life is a major defect in the tube material; NRC regulations 10CFR50, Appendix B, Criterion 16 specify that for a licensee to maintain his operating license, such non-conformance must be promptly identified and corrected. The licensee must assure that "corrective action Os) taken to preclude repetitiOn. NRC's General Design Criteria 4 and 10CFR50 Appendix A also specify that steam generator tubes must be able to" accommodate the effects of loss of coolant accidents" The fact that the NRC has not already raised these issues in any of their "Requests for Additional Information, RAls" indicates that the NRC would be ignoring its own regulations if it allows SCE to restart Unit 2.

In Summary: The SeE request for approval to operate Unit 2 at 70 % power for 150 days provided no explanation for the selection of this inspection interval. The absence of such explanation and the absence of an indication of the actions that would follow demonstrate the unreliability of SeE entire assessment of restarting Unit 2. Edison did not specify pass/fail criteria for the tubes during the outage inspection. Given the fact that fatigue damage does not lend itself to detection, SeE request is unacceptable and should be rejected.

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Joosten, Sandy From: Vinod Arora [vinnie48in@gmail.com]

Sent: Tuesday, April 09, 2013 8:01 PM To: CHAIRMAN Resource; Borchardt, Bill; leeds, Eric; Howell, Art; Dorman, Dan; Benney, Brian; Hall, Randy; R4AllEGATION Resource; lantz, Ryan

Subject:

San Onofre NRC/SCE/MHl/Public Awareness Series (2nd of 2 emails)

Attachments: Attachment 1.docx; ATTACHA_FINALREPORT_HopenfeldTestimony (1 ).pdf; TESTIMONYFINALWBA_HOPEN_032913 (1).pdf Honorable Dr. MacfarlanelMr. Borchardt/Mr. Leeds and Brilliant NRC Staff - FYI - Valuable Information for San Onofre, Sincerely - Vinnie Arora, PE, CEO/President, AVP Arora International Inc. (Non-Profit Engineering<:?rp?ration) (Attachment 1. and Dr. Joram Hopenfeld's Testimony and Summary RepQrts fil~d with Cal;rf~m1a Public Utilities Commission) 1

FlGlIRE 8B: (!J2) STFAM PRESSL'RE STEADY STATE OPERATICN eL"RVE!

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Figure 1 SONGS Unit 2 Steam Generators Steady Steady Operation Curve (,June 2012)

(Received From Anonymous SONGS Insiders) 1

Figure lA SONGS Unit 3 Steam Generators Steady Steady Operation Curve (June 2012)

(Received From Anonymous SONGS Insiders)

Root Cause EvaluatIOn, Urut ') ~ Generator lube Leak and Tube~ Tube We",

Condition Report 201836127, Re\fi$>on{J, 5:7J2(112 Si'IO Onofre N~Generatiog Statkm (SONGS) 2500 2500 650 650 2250 2250 6112 5980 553.0 541,3 198,000 2G9,880 1895 2003 2$4125 172.375 174.6:>

1.242,366 1,286,200 1,97HAO 2.041.300 1,S()S,437 1.546,700 Figure 2 - SONGS Steam Generator Design/Operations Features (Public Domain)

Reference Root Cause Evaluation: Unit 3 Steam Generator Tube Leak and Tube-to-Tube Wear Condition Report:

201836127, Revision 0, 5/7/2012, San Onofre Nuclear Generating Station (SONGS), Page 38 2

Figure 3 - SONGS Unit 3 SG E-088 Tube Damage - Figure Redrawn (Public Domain)

Solid Red Dots in circles show unprecedented tube-to-tube wear due to fluid elastic instability in the in-plane direction due to double the out-of plane velocities, high dry steam and high dynamic pressures caused by high reactor thermal power in the hot side heat flux, narrow tube to pitch diameter, excessive number of tall tubes, anti-FEI out-of-plane vibration bar structure, higher reactor coolant flows and operation at low steam pressures of 833 psi to generate more thermal heat out of the SG to generate more megawatts and more profits for EIX/SCE. This excessive heat in 4% region of the tube bundle with higher than normal wear(See Figures 10 through 18 below) was beyond the thermal design performance of the SG due to inadequate design and operational parameters discussed above. This adverse phenomena, studied only in experimental reactors, but not observed in operating steam generators prior to SONGS Unit 3, caused one tube leak in SONGS Unit 3 SGS, failure of 8 tubes at main steam line break testing pressure (3X MSlB pressure), loss of wall thickness in hundreds oftubes ~ 35% NRC & SONGS Technical Plugging limit, and additional 2 tubes lost wall thickness of 99%. If these 2 tubes would have leaked in conjunction with the one discovered leaking tube, SONGS Operators would not have been able to diagnose the accident in progress quickly enough, and could have resulted in a Meltdown of Unit 3. What you do not see in the above Red Dots, is that the hundreds of damaged tubes in Unit 2, in addition to losing strength due to wall thinning. have used up a significant fraction of their allowed fatigue life. Such damage cannot be detected by even the NRC Special Tube Inspections due to time, cost, unavailability of high technology probes and contactors, and/or impossible access within the tube bundle or radiation dose limitations. These tubes will be Significantly susceptible to sudden ruptures without notice or early warnings during steady state normal operations at 70% power, Operational Transients (opening or closing of valves, scrams, loss of offsite power, moderate (anticipated) earthquakes, etc.) and under steam line break accidents at other reactors such as Crauss, Turkey Point, Robinson &

Mihama (1991 and 2004 3

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Surlaco holt flux q" at afunction of txCO$$ ttmporaturo 4T** T.... Tut Figure 4 - Heat Flux as a function of difference in temperature between tube surface and steam saturation http://en.wikipedia.orglwikiIFilm_boiling Two-phase Steam water Fluid Flow: With reference to upflow in vertical channel (two-phase steam water flow in between the tubes in a nuclear steam generator), one can identify several flow regimes, or patterns, whose occurrence, for a given fluid, pressure and channel geometry, depends on the flow quality and flow rate. The situation of interest in nuclear reactors is nucleate boiling. Consider a vertical channel. At axial locations below the onset of nucleate boiling, the flow regime is single-phase liquid. As the fluid marches up the channel, more and more steam is generated because of the heat addition. As a result, the flow regime goes from bubbly flow (for relatively low values ofthe flow quality) to plug (intermediate quality) and annular (high quality). Eventually, the liquid film in contact with the wall dries out (this was the cause of SONGS Unit 3 Fluid Elastic Instability). In the region beyond the point of dry-out, the flow regime is mist flow and finally, when all droplets have evaporated, single-phase vapor flow (as occurred in SONGS Unit 3 Destruction and as would occur in Unit 2 under MSlB conditions).

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Mitsubishi Root Cause, Document No.L5-04GA588, Rev 0, Figure 3.3.3 MHI states in its Root Cause, "Causes of Type 1 Tube Wear (Tube-to-Tube wear): Most of the Type 1 wear (nw) indications suggest that the wear is due to tube in-plane motion (vibration) with a displacement (amplitude) greater than the distance between the tubes in the adjacent rows, resulting in tube-to-tube contact. Tube in-plane motion can be caused by turbulence and fluid elastic instability (FEI). However, turbulence induced (random) vibration by itself is insufficient to produce displacements of this magnitude. Displacements as large as those associated with in-plane tUbe-to-tube contact can only be produced by fluid elastic vibration. Further, the contiguous grouping of the nw tubes is another characteristic of fluid elastic instability. in order for large in-plane displacements to occur two conditions are necessary. First, the tube needs to be unrestrained in the in-plane direction and second the environment must be conducive to FEI (velocity, density, damping, etc.). MHI has analyzed whether random vibration was a precursor to the in-plane FEI that was observed in Unit 3. Two possible scenarios were considered. Scenario #1: In-plane FEI in Unit 3 had no precursor, Scenario #2: Wear from random vibration progresses to the point of loss of in plane support, followed by the onset of in-plane FE!. The first scenario is more likely supported based on the investigation: (1) While the number oftubes with tube-to-AVB wear without in-plane nw is greatest at the top of the tube bundle, the number of nw tubes with tube-to AVS wear is almost uniformly distributed along the different AVB intersections. If random vibration wear were a precursor for in-plane FEI nw, then the pattern of AVB wear for nw tubes should resemble the tube-to-AVB wear pattern (I.e. be concentrated at the top of the tube bundle). However, this is not observed for tubes with nw, (2) While the tube-to-AVB wear depth for tubes without in-plane nw is greatest at the top of the tube bundle, the tube-to-AVB wear depths for tubes with in-plane nw is almost uniformly distributed along the AVB intersections. (See Fig. 3.5-2.) If random vibration wear were a precursor for in-plane FEI wear, then the AVB wear for the tubes with in-plane FEI would be greatest at the top of the U-bends. But for nw tubes, the average wear depth is almost the same in all AVB support locations and there is no tendency to concentrate at the top of the tube bundle, and (3) The average 10% of AVB wear depth in Unit 2 and Unit 3 excluding nw tubes is almost the same. (See Fig. 3.5-2.) Therefore, if random vibration were a precursor to in-plane FEI one would expect to see a Similar number of tubes with tube-to-tube wear in the two RSG units. However, Unit 2 only has 2 tubes with nw."

5

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, 260 tonsJ ' 583 tOns I 268 tons I (triangular'pitch tubing) (triangular pitch tubing) (triangular pitch tubing)

Moisture separator Heat ,

transfer tube E,

Figure 6 - Comparison of Mitisubishi Steam Generators for Export (Public Domain)

A MHI 2006 Brochure states, "Designs differ between individual customers because the specifications of replacement components are determined for each individual power plant. There is no standard design for a replacement SG because the specifications and plant requirements vary among customers. Steam generators (SG) have been replaced in PWR plants worldwide for more than 20 years. Since power utilities replacing their SGs are apt to want to increase their electric output by improving efficiency and equipment reliability, an increase of the heat transfer area is often required for the replacement SGs. It is technically challenging to enhance the performance and improve reliability even using the latest technology due to the strict restrictions on the interface dimensions of the replacement equipment. Despite this tough situation, MHI has been increasing its exports steadily, and has already supplied four units since 2003 (two to Belgium and two to the USA), and is in the process of designing or manufacturing 12 more units (two to Belgium, four to the USA, and six to France). These achievements show that MHl's advanced technology, quality, and process control capability are acquiring the reputation for high reliability among European and American utilities."

Based on an Anonymous SONGS insider reports, "SCE installed brand new replacement engineers designed to last for 40 years with the purpose ofRSGs thermal performance to be as large as possible within the dimensional and other limitations." Westinghouse states, "For most of the straight leg section of the tube, the gap velocities at lower power levels and at 100% power are similar. The recirculating fluid flow rate is relatively constant at all power levels. However, in the U-bend region, the gap velocities are a strong function of power level. The steam flow in the bundle is cumulative and increases as a function of the power level and the bundle height which causes high fluid quality, void fraction, and secondary fluid velocities in the upper bundle.' MHI states, "The SCE/MHI AVB Design Team recognized that the design for the SONGS RSGs resulted in higher steam quality (void fraction) than previous designs and had considered making changes to the design to reduce the void fraction (e.g., using a larger down-comer, using larger flow slot design for the tube support plates, and even removing a TSP). But each of the considered changes had unacceptable consequences and the AVB Design Team agreed not to implement them. Among the difficulties associated with the potential changes was the possibility that making them could impede the ability to justify the RSG design under the provisions of 10 C.F.R. §50.59. Thus, one cannot say that use of a different code than FIT-III would have prevented the occurrence of the in-plane FEI observed in the SONGs RSGs or that any feasible design changes arising from the use of a different code would have reduced the void fraction sufficiently to avoid tube-to-tube wear. For the same reason, an analysis of the cumulative effects of the design changes including the departures from the OSG's design and MHI's previously successful designs would not have resulted in a design change that directly addressed in-plane FE!." I conclude that due to high steam flows and a normal than tall tube bundle (Average length of heated tubes increased from 680 inches to 730 inches along with the addition of 377 tubes specified by SCE), narrow tube pitch to tube diameter ratio, low tube-to-tube clearances, and steam generator operation at 833 psi caused void fractions of2: 99.6% in the high region of wear, which caused FEI in Unit 3. Had SCE/MHI not avoided the NRC 50.90 License amendment process, reduced the void fractions < 98.5 %, read Dr. Pettigrew's research 2006 papers and NUREG-1841, performed careful and thorough analysis of cumulative effects of the design changes including the departures from the OSG's design, FEI in Unit 3 could have been averted and Ratepayers would not have been subject to a potential nuclear meltdown and could have saved Billions of Dollars.

LESSONS LEARNT: Haste makes waste, inadequate use of human performance tools and improperly conducted and poorly designed research experiments can cause (PotentialjActual) nuclear meltdowns and accidents as we saw in Chernobyl, Fukushima, Three Mile Island, Fukushima, Mihama and SONGS 3. Southern Californians want to avert a Potential meltdown of "defectively designed and degraded Unit 2 by not allowing NRC/SCE/MHI to restart it at any power level as an unapproved experiment without adequate repairs/replacement.

6

Fig. 1 Comparison of steam generators for export Designs differ between individual customers because the specifications of replacement components are determined for each individual power plant.

In-plarie direction ' - . / .. Qut-of:-prarie direction

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Ann.,nll"",,\n bar RoW iliiectlon 'Nater-steam n1xed flOw Fig. 2 Tube support structure at U-bend Fig.3 High-performance small-size separator Included in improvement designs, this is avibration prevention mechanism for (single structure) heat transfer tubes. Another improvement design, this is ahigh*performance small*size separator.

Figure 7 - Comparison of Mitisubishi Steam Generators for Export (Public Domain)

A MHI 2006 Brochure states, "Major items of improvements in Replacement Steam Generators: There is no standard design for a replacement SG because the specifications and plant requirements vary among customers. Figure 15 compares the dimensions of recently exported SGs, in which widely varied specifications were applied. However, by applying the following latest advanced technologies to all SGs, improvements were made which cope with all past problems such as tube corrosion, vibration and wear, fatigue, and water hammer, and products which satisfy customers' advanced demands for heat transfer capability and moisture content are being supplied. (1) Tube material of high nickel alloy TT690 with excellent corrosion resistance. (2) Outstanding tube support plate design, tube expansion technology in tube sheets. (3) Tube support structure at U*bends with high support function. (4) High*

performance moisture separators. Of these advanced demands for recent replacement SGs, items (3) and (4) deserve special attention.

2.2.1 Tube support structure at U*bends (3) The tube support structure at a U*bend is shown in Fig. 2. This is a unique design with reduced flow resistance while assuring a high support function by increasing the number of support points. Together with excellent assembly technology duling manufacturing, high reliability against vibration and wear of heat transfer tubes is achieved. 2.2.2 High*

performance moisture separators (4) MHI has developed a small, high*performance moisture separator by optimizing the geometry of the parts based on extensive field pressure tests (Fig. 3). As a result, replacement SGs corresponding to power up-rating and/or advanced moisture requirements can be designed."

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Figure 8 - Vibrations amplitude as a function of flow pitch velocity for a flexible cylinder in a rigid cluster Violette R., Dr. Petticrew M. J. & Dr. Mureithi N. W. state in a paper published in 2006 "In nuclear power plant steam generators, U tubes are very susceptible to undergo fluid elastic instability (See Figure 3 below) because of the high velocity of the two-phase mixture flow in the U-tube region and also because of their low natural frequencies in their out of plane modes. In nuclear power plant steam generator deSign, flat bar supports have been introduced in order to restrain vibrations of the U-tubes in the out of plane direction. Since those supports are not as effective in restraining the in-plane vibrations of the tubes, there is a clear need to verify if fluid elastic instability can occur for a cluster of cylinders preferentially flexible in the flow direction. Almost all the available data about fluid elastic instability of heat exchanger tube bundles concerns tubes that are asyrnrnetrically flexible. In those cases, the instability is found to be mostly in the direction transverse to the flow. Thus, the direction parallel to the flow has raised less concern in terms of bundle stability."

8

Table 1 - San Onofre RSGS Design and Operational Data U2/30SGs Other U2/30SGs before Plants after 2001 2001 Design and Operational Parameters PVNGS Power Power which caused FEI, FIRV and MFE U2 RSGs U3 RSGs ANOU2 uprate uprate (1) Reactor Thermal Power, MWt 1724 1729 1995 1729 1705

[lA) Unit Electrical Generation, MWe 1183 1186 N/A N/A N/A (2) Number ofTubes 9727 9727 12,580 9350 9350 (3) Average Length of Heated Tubes, 729.56 729.56 766.8 680.64 680.64 inchesm (4) Heat Transfer Area, ft2 116,100 116,100 105,000 105,000 (5) Tube Wall Thickness, inches 0.043 0.043 0.043 0.048 0.048

[SA) Tube Diameter, inches 0.75 0.75 0.75 0.75 0.75 (5B) Tube Pitch, inches 1.0 1.0 0.87 1.0 1.0 (5C) Tube Array Triangular Triangular Triangula Triangular Triangular r/Square (5D) Tube Index 1.33-1.43 1.33 - 1.43 1.52-1.67 1.33-1.43 1.33-1.43 (5E) Tube to tube clearance, inches 0.25 0.25 0.25 0.25 (SF) Nominal Gap between tube and 0.002 0.002 0.003 N/A N/A AVB", cold, inches (5G) Nominal Gap between tube and 0 0 N/A N/A N/A AVB", Hot, inches (5H) Nominal Gap, Manufacturing N/A N/A N/A N/A N/A Dispersion, inches (51) Tube Wall ThicknessfTube 0.057 0.057 N/A 0.064 0.064 Diameter Ratio (5)) Average Heated Tube Length/Tube 973 973 N/A 908 908 Diameter Ratio (6) Reactor Coolant Flow (at cold leg 75.76 79.79 5.6' 198,000 198,000 temperature)" Million lbs.lhour (6A) Reactor Coolant Operating 598 598 618.9 599 611 Temperature (Thot), OF (6B) Reactor Coolant Operating 541.3 545.3 560.9 542 553 Temperature (Tcold), OF (7) RSG Operating Pressure (@100% 892-942 833 1039 816 900 power), psia (7A) Steam Saturation Temperature,oF 531 523 549 520.5 532 (8) Feed-water Inlet Temperature, OF 442 442 N/A 445 445 (9) Feedwater/Steam How, Million 7.7/7.6 7.7/7.6 8.95 7.18-7.63 7.41 lbs.lhour (10) Steam Moisture Content, % <0.10 <0.10 <0.20 <0.20 (11) Steam Quality, %  ? 87.6 - 89.7 73.4 N/A N/A (llA) Void Fraction, % < 98.5% 99.6% 98.5 96.1 96.1 (l1B) Interstitial or Gap Fluid Velocity, -v22.00 28.30 17.91 22.90 22 feet/second (l1C)Maximum Dynamic Pressure < 4000 4140 4220 N/A N/A (N/m2)

(12) Reactor Coolant Volume, ft3 2003 2003 N/A 1895 1895 (13) Circulation Ratio 3.3 3.3 4.3-5.7 N/A N/A (14) Delta Te (Ts, 6A) (TsAT, 7A) ,oF 67 75 68 78.5 79 Overall Heat Transfer Coefficient 1235 1280 N/A N/A N/A (Estimated) Btu/hr-ft2-oF Fluid Elastic Instability NO YES NO NO NO Flow-induced Random Vibration YES YES  ? YES YES Mitsubishi Flowering Effect YES YES N/A N/A N/A 9

Table 2 - San Onofre Unit 2 RSGS Design and Operational Data U2 RSGs Design and Operational Parameters which caused FEI, @100% U3 RSGs U2 RSGs FIRVand MFE Power @70%Power @MSLB (11 Reactor Thermal Power, MWt 1729 1")1'"

0 (lA) Unit Electrical Generation, MWe 1183  ? N/A

~"OfTUb" 9727 9727 12,580 of Heated Tubes, inches m 729.56 729.56 729.56

4) Heat Transfer Area, ft2 116,100 116,100 116,100 (5) Tube Wall Thickness, inches 0.043 0.043 0.043 (SA) Tube Diameter, inches 0.75 0.75 0.75 (5B) Tube Pitch, inches 1.0 1.0 0.87 (5C) Tube Array Triangular Triangular Triangular/Square (5D) Tube Index 1.33-1.43 1.33 1.43 1.52-1.67 (5E) Tube to tube clearance, inches 0.25 0.25 Not Determined (SF) Nominal Gap between tube and AVB", cold, 0.002 0.002 0.003 inches

[5G) Nominal Gap between tube and AVB", Hot, inches 0 0 0 (5H) Nominal Gap, Manufacturing Dispersion, inches N/A N/A N/A (51) Tube Wall Thickness/Tube Diameter Ratio 0.057 0.057 0.057 (5)) Average Heated Tube Length/Tube Diameter 973 973 973 Ratio (6) Reactor Coolant Flow (at cold leg temperature)., 79.8 78.2 78.2 Million lbs.lhour (6A) Reactor Coolant Operating Temperature (Thot), 598 591 591 of (6B) Reactor Coolant Operating Temperature (Tcold), 541 551 551 of (7) RSG Operating Pressure (@100% power), psia 892-942 946 ATM (8) Steam Operating Temperature (@ 0% power), up 531 538 212 (8A) Stearn Flow, Million lbs.lhour 7.6 5.1 549 (8B) Feed-water Inlet Temperature, Op 442 407 N/A *

(9) Feedwater Flow, Million lbs.lhour 7.6 5.1 33.8* to Environment

• in 3-5 Minutes Quality, %

raction, % 90%~

98.5%

>90%

100%

Gap Fluid Velocity, feet/second 25.1 > 50

• (11B )Secondary fluid density, Ibm/cubic feet 7 12 <7 (12) Reactor Coolant Volume, ft3 2003 2003 2003 to Environment in 3-5 Minutes (13) Circulation Ratio 3.3 4.9 0 (13A) Down-comer Feed-water Flow, Ml 24.8 24.8 0 (14) Delta Te= (Ts, 6Al- (TSAT, 7A), OF 67 53 -400 Fluid Elastic Instability NO NO YES (Film Boiling)

Significant Radiation Flow-induced Random Vibration YES YES YES Mitsubishi Flowering Effect YES YES YES Flashing Feedwater let Impingement Forces on Tubes NO NO YES 10

BEFORE THE PUBLIC UTILITIES COMMISSION OF THE STATE OF CALIFORNIA Order Instituting Investigation on the ) Investigation 12-10-013 Commission's Own Motion into the Rates, )

Operations, Practices, Services and )

) (Filed October 25,2012)

Facilities of Southern California Edison Company and San Diego Gas and Electric )

Company Associated with the San Onofre )

Nuclear Generating Station Units 2 and 3. )

)

)

DR. JORAM HOPENFELD'S DIRECT TESTIMONY ON BEHALF OF THE WORLD BUSINESS ACADEMY Sabrina Venskus Venskus & Associates, P.C.

21 South California Street, Suite 204 Ventura, CA 93001 Phone: 805.641.0247 Fax: 213.482.4246 venskus@lawsv.com Counsel to World Business Academy

1 BEFORE THE PUBLIC UTILITIES COMMISSION 2 OF THE STATE OF CALIFORNIA 3

Order Instituting Investigation on the ) Investigation 12-10-013 4 Commission's Own Motion into the Rates, )

Operations, Practices, Services and )

5 ) (Filed October 25,2012)

Facilities of Southern California Edison Company and San Diego Gas and Electric )

6 Company Associated with the San Onofre ) Exhibit: WBA-Ol 7 Nuclear Generating Station Units 2 and 3. )

)

8 )

9 10 PREPARED TESTIMONY OF DR. JORAM HOPENFELD 11 Q: Please state your name and business address for the record.

12 A: Joram Hopenfeld. 1724 Yale PI. Rockville, MD 20850.

13 14 Q: What is your academic background?

15 A: Engineering, University of California at Los Angeles: BS 1960, MS 1962, Ph.D 16 1967.

17 18 Q: What are your professional qualifications?

19 A: I have 50 years of experience in industry and government primarily in the areas of 20 steam generator performance, thermal-hydraulic, and material testing in nuclear and 21 fossil power plants. My major activities were focused on corrosion/erosion, thermal 22 hydraulics in Sodium Cooled Nuclear Power Plants, Pressurized Water Reactors, 23 ("PWRs") and in Coal Fired Plants. I have managed a major international program on 24 steam generator performance during accidents involving various thermal transients in 25 PWRs. Attachment B to this testimony contains my Curriculum Vitae.

26 27 Q: Please describe any additional qualifications in the areas of steam generator 28 operations.

Page 2 of13

A: As a result of my work at the Nuclear Regulatory Commission, ("NRC") my 2 opinion regarding the safety implication of steam generator tube degradation was 3 adopted. Consequently, in 2001 the NRC launched a STEAM GENERATOR ACTION 4 PLAN, SGAP (NRC web site) to address the various safety issues that I have raised in a 5 series of documents starting 1992, known as the DPO. In September 2007 the NRC 6 issued a new performance technical requirement specification, TS, to reduce the risk from 7 accident induced tube ruptures and tube ruptures during normal operations, 8 I made many presentations to the Atomics Safety Licensing Board (ASLB) and the 9 Advisory Committee on Reactor Safety (ACRS) on the following issues:

10

• Erosion/Corrosion ( FAC), Steam Generator Degradation Monitoring.

11

• Safety Consequences of Steam Generator Tube Failures, Iodine transport and 12 spiking, POD of crack detection by Eddy Current, Metal Fatigue - PWRs and l3 BWRS, and Vibrations in BWR dryers.

14

• Managed sensitivity studies with the RELAP computer code on the ability of the 15 operator to keep the SG ("Steam Generator") secondary water at mid level as a 16 function of the number ruptured tubes.

17

• Designed, fabricated and field tested instrumentation for a very harsh vibration 18 environment.

19

• Modeled jet erosion as a potential for leakage increase during SG accidents.

20 21 Q: Have you testified at this Commission previously?

22 A: Yes. In 2005 I reviewed SCE's cost-benefit analysis to justifY the replacements of 23 the four SGs at SONGS (Proceeding A.04-02-026.) I had identified aging, terrorist acts 24 and regulatory actions as the major uncertainties which were not included in the analysis.

25 In addressing aging I discussed the "bathtub' concept where early failures occur due to 26 design, fabrication, and installation errors while later failures occur due to wear. My 27 discussion was focused on the aging of components other than the SG tubes because SCE 28 did not indicate that the RSGs would incorporate many new design features. Had I Page 3 of13

1 known that SCE contemplated such changes I would have focused my attention more on 2 the initial portion of the 'bathtub' curve, which I have discussed in my testimony.

3 4 Q: What is the purpose of your testimony today?

5 A: The purpose of my testimony is to provide expert opinion on the nature and effects 6 of the steam generator failures. In connection with this purpose, I have also provided my 7 expert opinion about the implications and risks of restarting SONGS Unit 2 at 70%

8 license power.

9 10 Q: What documents did you review for purposes of preparation of your written 11 testimony?

12 A: I have reviewed the following documents:

13 a. Augmentation Inspection Team Report July 18,2012 14 b. NEI96-07 15 c. 10CFR50.59 16 d. SONGS Return to Service Report Oct. 2012 17 e. SCE Root Cause Evaluation May 7, 2012 18 f. SCE Root Cause UES-20120254, March 2013 19 g. Many technical papers used in the preparations of my report, which is attached to 20 this testimony. (see list of references on page 25 of the report) 21 h. Selected excerpts of SCE testimony 22 23 Q: Do you know why SeE shut down SONGS Units 2 and 3 in January 2012?

24 A: Yes. It was shut down due to excessive SG tube wear and a tube leak, which 25 caused the release of radioactive materials into the environment.

26 27 Q: What is the function of a steam generator in a pressurized water reactor 28 (PWR)?

Page 4 of13

1 A: The SG functions as a heat exchanger, by means of which the high temperature 2 pressurized heated radioactive primary water on the inside of the tubes heats up the 3 secondary water on the outside of the tubes, in order to generate the steam that turns the 4 turbine which in tum generates electricity. In addition to providing a barrier to 5 radioactivity and producing steam, a steam generator has many other important functions.

6 It is the major component in the plant that contributes to safety during transients and 7 accidents. A steam generator provides the driving force for natural circulation and it 8 facilitates heat removal from the reactor core during a wide range of loss of coolant 9 accidents. Proper steam generator operation is of major safety significance and therefore 10 any changes to its design may have potential safety consequences.

11 12 Q: Describe how many tubes there are in each SG at SONGS?

13 A: Approximately 9,700.

14 15 Q: Describe how many tubes there are in total in SONGS Units 2 &3?

16 A: Approximately 19,400.

17 18 Q: Do you know how many tubes are currently defective at SONGS according to 19 the NRC?

20 A: Approximately 3400 21' 22 Q: Are any tubes experiencing thinning to a significant degree?

23 A: Yes. 387 tubes are experiencing thinning beyond the "safe threshold" as defined 24 by the plugging limit. To prevent a tube, or any other component, from bursting under 25 pressure the ASME code as well as the NRC require that the tube maintain a certain 26 minimum wall thickness. Tubes must be plugged when their wall tickenness is less than 27 40% of the nominal wall thickness.

28 Page 5 of13

1 Q: Do you believe it is possible for SCE or the NRC to know the total number of 2 tubes that are defective at SONGS?

3 A: No 4

5 Q: Please explain why.

6 A: To meet the performance criteria as specified in the plant's technical specifications 7 ("TS") it is necessary to relate tube defect geometry to primary/secondary side leakage.

8 When tube degradation is caused only by thinning, the degree of tube damage can be 9 assessed by modeling, pulling selective tubes for testing or by in-situ pressure testing. As 10 discussed extensively in my report entitled "An Assesment Of San-Onofre Steam 11 Generator Tube Failures" ("Report")( seeAttachment A), in addition to losing strength 12 due to wall thinning some of the tubes at SONGS used up a significant fraction of their 13 allowed fatigue life. Such damage cannot be detected during service inspections.

14 15 Q: How many tubes have been plugged?

16 A: 1,322 were plugged. To view this number in perspective, it should be noted that 17 as of 1998 a total of only 188 tubes were plugged in 62 US plants.

18 19 Q: Do you believe that SCEIMID vibration fatigue analysis requires a more 20 detailed evaluation and if so, why?

21 A: Yes. Because the analysis is fundamentally flawed.

22 23 Q: Please explain the basis for your answer.

24 A: MHI used an incorrect value for the stress concentration factor and consequently 25 their calculated peak stress intensity for the tubes is incorrect. My analysis in the 26 attached Report at Attachment A shows that the stress concentration factor is much 27 higher and therefore the endurance limit will be exceed by at least 22%. This means that 28 some tubes have already used up a significant fraction of their fatigue life. In this case, a Page 60f13

1 sudden increase in tube stress due to operational or an accident transients would cause an 2 immediate tube rupture.

3 The flow-induced mechanical vibrations, such as those observed at SONGS, are a 4 main source to high cycle fatigue damage. Such fatigue damage cannot be measured with 5 existing plant instrumentation. Fatigue damage resulted in several tube failures in US 6 steam generators, yet in spite of its importance, none of the SCE consultant included 7 fatigue in their evaluation of restarting Unit 2.

8 As I discuss in detail in my Report, the Cumulative Usage Factor ("CUF")

9 determines the fatigue life of components in nuclear plants. This is a calculated number 10 that the ASME code requires of operators of pressure systems to ensure that the CUF be 11 maintained below unity. The NRC adopted the ASME code methodology. The fact that 12 some tubes would exceed the endurance limit indicates that their CUF had already 13 exceeded unity.

14 The severity of cyclic stresses must be properly estimated and not be based on 15 arbitrary non-conservative assumptions. As discussed in my report, excessive redaction 16 by MHI makes the task of finding errors in the analysis difficult. To restart Unit 2 SCE 17 must defend its position with a technically conservative methodology, that fatigue 18 damage can be ignored.

19 20 Q: Have you formed an opinion as to (1) why the unit 2 and 3 SGs are so flawed 21 and (2) why the tubes started leaking within 1.0 years of operation, in Unit 3.

22 A: The most probable mechanism in unit 3 was high fluid forces acting on the tubes 23 due to Fluid Elastic Instability ("FEI"). It is less clear what the mechanism was that 24 caused the vibrations in Unit 2. The specific technical root cause that triggered FEI and 25 resulted in the rapid tube wear is not known. In my opinion the lack of independent 26 competent review of the significant changes that were made to the original steam 27 generators was a major reason for the rapid tube wear.

28 Page 7 of13

1 Q: Why in your opinion did SCE not make the suggested design changes in order 2 to reduce the risk of SG degradation and tube leaks in the replacement steam 3 generators?

4 A: SCE did not provide any data to support their contention that the various design 5 change options, which were discussed in 2005 by the AVB Design Team, would have 6 had no significant effect on flow velocities and steam quality. Without knowing how each 7 of those design changes would have reduced the risk of extensive vibrations I cannot 8 opine why SCE did not make the suggested changes. To support their contention SCE 9 should have included in their root cause report, a sensitivity analysis over the full 10 spectrum of the proposed design changes to demonstrate the lack of benefits from such 11 changes.

12 Q: Why in your opinion did SCE not inform the NRC that the proposed MRI 13 steam generators represented significant changes from the original steam generator 14 design approved by NRC?

15 A: SCE either wanted to avoid a more lengthy public and NRC review through a 16 license amendment as required by CFR §50.59 or they did not understand the degree to 17 which the design changes could affect vibrations.

18 19 Q: In your opinion, given these new design changes and associated tube 20 degradation, was SCE operating SONGS outside of its current licensing base?

21 A: Yes. The fact that the in situ pressure tests in unit 3 caused eight tubes to leak and 22 fail their performance criteria is a clear indication that SONGS operated beyond its 23 current licensing base ("CLB") during some months in 2011.

24 25 Q: In your opinion, given these new design changes and associated tube 26 degradation, if SCE restarts SONGS, will it be operating outside of its current 27 licensing base?

28 Page 8 of 13

1 A: As discussed in my Report, the fact that a large number of tubes, more than 500 in 2 each unit, have been plugged and will continue to wear requires a demonstration that 3 these tubes will not cause damage to adjacent tubes during accidents and meet the 4 present licensing requirements.

5 6 Q: Returning to the MHI designed and manufactured replacement steam 7 generators, why have there been so many tube failures within such a relative short 8 period of their operation?

9 A: The tube degradation can be traced to the major design changes made to the 10 original steam generators:

11

• Tube support plate (TSP) geometry from (broached to eggcrate) 12

• U Tube geometry (bend vs. square) 13

• Length and number of tubes, tube material and tube wall thickness.

14

• Original stay cylinders were replaced by tubes 15

• Nominal tube/AVB gap 16

• Operating pressure 17

• Void fraction 18 19 I discuss the impact of these changes on tube failures in my Report. I conclude that 20 any person familiar with the state of the art of steam generator design and operation 21 should have recognized that these changes can have a significant effect on tube vibration 22 and safety.

23 24 Q: Are you familiar with the changes that the design team was contemplating to 25 minimize the risk of vibration in the RSG?

26 A: Yes. MHI and seE claim that their studies showed that the contemplated changes 27 would have not reduced the risk significantly because their computer FIVATS predicted a 28 stability ratio less than one. In spite the fact that data showed that in plane FEI can occur Page 9 of13

I SCE that that there was "negligible possibility of fluid elastic vibration". As already 2 mentioned above, a sensitivity analysis could be very helpful in understanding SCE IMHI 3 claims.

4 5 Q: Do you believe that SCE and its consultants MHI, AREVA, and 6 Westinghouse thoroughly assessed the risk of restarting Unit 2?

7 A: No. The consultants did not address the increase in safety risk associated with 8 operating Unit 2, at any power leveL As discussed in my Report, because the tubes in 9 Unit 2 have already sustained certain fatigue damage, that damage must be related to 10 safety risk. A possible methodology in this regard, is to relate the change, L\, of the large 11 early release frequency (LERF) to the LERF when the replacement steam generators 12 started service and the tubes were in pristine condition. LERF is a relevant parameter 13 because it represents a condition where a core has been uncovered and the containment 14 has been bypassed.

15 16 Q. What other important issues were not addressed in SCE and consultants' 17 reports?

18 A. At least two issues: 1. Long term Implications of restarting any of the SONGS 19 units. 2. Uncertainties in the consultants' analyses Fluid Elastic Instability Analysis 20 (FEI) 21 1. Long Term Effects. Because of the large number of defects in the tubes, there is a 22 risk that additional tubes will have to be removed from service even if the FEI 23 problem has been solved. Defects such as now exist (more than 3000) are known 24 to form nucleation sites for stress corrosion (SCC) and fatigue cracks. It is 25 important to understand that even though alloy 690 is not as prone to SCC as alloy 26 600 it is not completely immune to SCC. This problem will become more 27 important as the units age because of crud build up in the tube support plates.

28 Page 10 of13

1 2. Uncertainties. As discussed in the attached Report, the consultants did not 2 address uncertainties in sufficient details. Because of the complexity of the 3 technical issues a reviewer, therefore, cannot asses of the robustness of the 4 analysis.

5 6 Q: In your opinion does the operation of these replacement steam generators 7 represent a potential public health and safety hazard?

8 A: Yes. The operation of the new steam generators represents an unacceptable risk as 9 measured by NRC safety goals of 10-4 per year, as discussed in detail in my Report.

10 11 Q: In your opinion, given the current condition of the replacement SG can the 12 plant in its present condition be safely turned on, without putting the public at 13 additional risk?

14 A: No. As I have explained in my report the tubes in the RSGs have already been 15 damaged and if the plants are restarted sooner or later the tubes will leak.

16 17 Q: Is it true that a leak in even a single tube at SONGS will lead to releases of 18 radiation into the environment?

19 A: Yes.

20 21 Q: Is it true that the more tubes that leak the more radiation will be released into 22 the environment?

23 A: Yes.

24 25 Q: In your opinion, is there a risk of a major accident and radiation release, in 26 the form of a steam line break, if SONGS is restarted?

27 28 Page 11 of 13

1 A: Yes. A steam line break is a design basis accident, DBA. All reactors are required 2 to withstand such accidents because of the finite risk that they will occur. As I have 3 discussed in my report, because of the severity of tube degradation the probability of a 4 core melt with a large early release ("LERF"), is 1 in 10,000 years during any cycle of 5 operation.

6 7 Q: Describe what is meant by a steam line break.

8 A: A steam line is a large pipe that exits the steam generator inside the containment 9 and enters the turbine building outside the containment. A break in the pipe, for 10 whatever reason, will release steam directly to the environment. Steam line breaks do not 11 occur frequently, however, they have occurred both in nuclear and fossil plants and 12 therefore they are included in the DBAs.

13 14 Q: Describe how a steam line break causes tube leaks.

15 A: A steam line break can cause a degraded tube to leak because the rapid 16 depressurization following the break would impose high stresses on the tube.

17 18 Q: Do you have an opinion on the percentage risk of a steam line break if 19 SONGS is restarted and operated through the end of the SONGS operating license 20 in 2022?

21 A: Yes. Risk assessment studies are based on a probability of 10_4 per year that a 22 steam line break will occur.

23 24 Q: Is this prepared testimony and the Report entitled "An Assesment Of San 25 Onofre Steam Generator Tube Failures" dated March 25, 2013 referenced herein 26 and attached as Attachment A hereto, your complete testimony at this time?

27 A: Yes.

28 Page 12 of13

1 Q: Does this conclude your prepared testimony?

2 A: Yes.

3 4

5 Executed on March 25. 2013 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 Page 13 of13

AN ASSESMENT OF SAN-ONOFRE STEAM GENERATOR TUBE FAILURES Dr. Joram Hopenfeld March 25, 2013 1

TABLE OF CONTENTS SUMMARy ........................................................................................................................ 3 DISCUSSION ..................................................................................................................... 5

1. General Comments .................................................................................................... 5 A. Unwarranted Redactions ................................................................................... 5 B. Unspecified Uncertainties .................................................................................. 5 C. Effects of Wear and Fatigue Interaction on the Accuracy of Wear Predictions for Unit 2 Restart .................................................................................................... 6

2. SCE Proposed Restart of Unit 2 at 70% of Licensed Power ..................................... 6 A. Fitness for Service Assessment .......................................................................... 6 B. Accident Progression .......................................................................................... 7 C. Failure ofRSG Tubes to Meet Performance Criteria ....................................... 8

3. Title 10 Code of Federal Regulations Part 50.59 Requirements ............................. 10

4. NRC Assessment ..................................................................................................... 12 APPENDIX A - FATIGUE ANALySIS .......................................................................... 14 A. Stress Concentration ................................................................................................ 14 B. Loss of Wall Thickness (wall thinning) ................................................................... 15 C. Surface Finish .......................................................................................................... 15 D. Corrected MHI Stress Calculation ........................................................................... 16 APPENDIX B - BASIC CONCEPTS ............................................................................... 23 REFERENCES ................................................................................................................. 25 2

SUMMARY

A steam generator, in addition to providing a barrier to radioactivity and producing steam, has many other important functions. It is the major component in the plant that contributes to safety during transients and accidents. It provides the driving force for natural circulation and it facilitates heat removal from the reactor core during a wide range of loss of coolant accidents. Proper steam generator operation is of major safety significance and therefore any changes to its design may have potential safety consequences.

Southern California Edison("SCE") has not identified the root cause for the unusually excessive tube wears in the four steam generators ("SGs") of units 2 and 3 at San Onofre Nuclear Generating Station, SONGS. The two units were shut down in January 2012 and have been idle since. SCE is assessing whether the units can operate safely at their licensed power.

Based on my evaluation of the tube wear data and the in situ leak test results it is my opinion that restarting Units 2 and 3 would compromise public safety. The new components in the replacement steam generators("RSGs"), constituted a major change to the original SGs, this lead to vibrations and the unusual rapid tube wear. The components causing the wear would have to be replaced and the SONGS license amended before the units can be restarted.

The above assessment also applies to SCE proposed five-month test of Unit 2 at 70% of licensed power. After correcting an error in the SCE stress calculations, the present analysis shows that because of the wear damage previously sustained by Unit 2 some tubes will be susceptible to rapid fatigue failure. The tubes will exceed their allowable fatigue life by 22 to 29% during the next operating cycle. The risk of tube rupture increases with operating time but the analysis is not capable of quantifying it in terms of operating time. Unit 2 should not be permitted to operate until SCE provides a thorough assessment of the fretting fatigue issue discussed in this report.

SCE did not define specific pass-fail criteria for the proposed Unit 2 test; it would be difficult to evaluate the validity of such criteria because failures by vibration fatigue cannot be measured with existing plant instrumentation. Benefit-cost analysis for Unit 2 test was also not provided.

Because of the unprecedented severity and scale of tube degradation at SONGS there is no data to determine the safety consequences of operations with failure prone 3

tubes. The existing performance criteria are of little help because they were derived from accident progression of tubes with small stress corrosion cracks--rather than accident progression with fatigued tubes. These criteria are not sufficiently conservative because primary to secondary leakage is expected to develop more suddenly when tube failure is caused by vibration fatigue.

Because the uncertainties in predicting how fatigued tubes can propagate failures, it is impossible to assess quantitatively safety risks. The present assessment was therefore limited to a discussion of discrete accident scenarios without assigning a probability to any specific scenario. It is believed that a main steam line break (MSLB) represents a bounding case. A conservative estimate of the probability of a large early release of radiation with containment bypass would be 10-4 per year for any operating cycle. Such risk exceeds NRC's safety goals. Units 2 and 3 at SONGS have the highest risk of tube rupture related core damage than any other nuclear power plant in the US.

Based on my review of the AIT report, the March 2013 MHI root cause report, NEI 96-07, and 10 CFR 50.59, I conclude that SCE did not adhere to the 50.59 guidelines regarding license amendment requirements. A main cause of the RSG failures can be attributed to the fact that the design changes initially were not reviewed by a financially independent and competent engineering organization.. The documents produced by SCE following the January 2012 incident contain serious technical flaws and reflect lack of impartiality.

SCE andMHI did not provide any data to support their contention that the various design change options that were discussed in 2006 by the AVB Design Team would have had no significant effect on flow velocities and steam quality.

SCE consultants AREVA, Westinghouse and MHI differed on the root cause of tube vibrations and none of their opinions were based on sound scientific principles. The safety consequences of operating with degraded tubes are more serious than envisioned by the consultants.

It is apparent that SCE focused its attention on explaining away errors in the design, fabrication and management of the RSGs. There is no indication that any consideration was given to the long-term safety risk of operating Units 2 and 3, with each containing more than 1500 defects (3401 in all).

4

DISCUSSION

1. General Comments It is well understood that given the complexity of the situation, there will be uncertainties, but SCE and its consultants do not identify or address any uncertainties.

Thus, the reports lack in transparency and omit key critical information preventing the reader from verifying key assumptions and assessing the robustness of the conclusions.

Following are some examples that illustrate this point.

A. Unwarranted Redactions As discussed in Appendix A, MHI redacted the value of the fillet radius that it assumed in the fatigue analysis. This assumption has a significant effect on determining tube fatigue life. Since the visual examination of the tubes at SONGS can not be used to measure this radius, one must conclude that MHI did not use any proprietary information in assigning a value to the fillet radius. In fact, by MHI's own admission, (Figure 3), that "the wear surface cannot be seen" it can be concluded that MHI selected an arbitrary value for the redacted fillet radius. The redacted fillet radius has no commercial value to MHI; the redaction appears to be intended only to mislead readers that do not wish or think to question MHI's assumptions.

B. Unspecified Uncertainties The stability ratio ("SR") is a critical parameter which was calculated by SCE (Attachment 6 - Appendix D) to show that at 70% power the amplitude of vibration will not result in rapid tube rupture by excessive wear. SR<1 is an adequate criteria against high vibrations but it must be understood that the SR is based on experimental constants (the instability constant K for example) which were generated in laboratory tests. When that laboratory test do not properly represent the RSG prototype, the final SR values are subject to uncertainties.

For example, Westinghouse obtained the K values in their test model MB-2 (Ref 1, Appendix p. 14). Since the MB-2 tubes are shorter in length than their model F (Ref 9) the reported values ofK are subject to large uncertainties. It is not at all clear how the MB-2 data is applicable to the SONGS RSGs. To appropriately scale the Westinghouse K values to the RSG, it would fITst be necessary to apply correction factors to allow for the difference in length and tube support geometries. While reducing power level decreases the probability of tube wear it is not at all clear that the 70% power level 5

represents a conservative value when considering the large uncertainties in the Westinghouse analysis.

TH computer codes are based on the porous model and proprietary experimental data and correlations. Sensitivity studies are commonly used to highlight the effects of design changes on code predictions, thereby allowing uncertainty estimates. Since the FIT -Ill code has not been benchmarked and MHI has not provided any sensitivity results, it is impossible to tell whether MHI's predictions were in error due to a faulty computer code or a faulty input.

When SCE assigns a confidence level of 50% to certain predicted SR value it is impossible to assess its validity because the uncertainties are unknown. It is therefore difficult to assess the risk associated with the 70% power operations. An independent reviewer has no choice but to use bounding models to predict risk.

C. Effects of Wear and Fatigue Interaction on the Accuracy ofWear Predictions for Unit 2 Restart In spite of the synergy between wear and fatigue (Ref 10), SCE did not address it.

. None ofSCE's consultants apparently considered the synergetic effects of wear and fatigue in their assessment of tube wear in Unit 2 at the reduced power level. For a given tube the repeated impacts on its surface together with its sliding motion will affect tube wear. The major uncertainty in using the ASME data relates to the incubation period, which depends on the combination of number of impacts and sliding motions. No data is available to clarifY this point; physical reasoning would dictate that the ASME plots used by MHI are not sufficiently conservative for the type of wear that has occurred at SONGS.

2. SeE Proposed Restart of Unit 2 at 70% of Licensed Power A. Fitness for Service Assessment The proposed restart of Unit 2 is essentially a test. SCE did not provide success or fail criteria for the test nor did SCE indicate how the test could impact future SONGS operations. Nevertheless, understanding the risk of restarting and operating unit 2 at any power level for any length of time is important because it provides insight on the reliability of future operations at any power level.

The unusual large vibration experienced by the RSG weakened the tubes by reducing their fatigue life and resistance to burst pressure. Using the fatigue model 6

developed by MHI (Ref. I Attach. 4) but correcting for a mistake in the stress calculations, Appendix A, it can be concluded that some tubes, including those which were plugged, will use up their fatigue allowance soon after reactor restart. These tubes will be susceptible to sudden rupture during normal operations and under steam line break accidents. To ensure safe operations, the ASME code requires that the tubes not be stressed beyond their endurance limit (13.6 ksi). As shown in Appendix A, the endurance limit will be exceeded by 22 to 30% during the next cycle. These numbers may not be sufficiently conservative because of the unknown surface roughness of the worn tubes. Unlike loss of tube strength due to wall thinning alone, which can be detected during in service inspections, vibration fatigue does not lend itself to periodic inspections because a large fraction of the tube fatigue life is spent in the crack incubation period where it cannot be detected. Once the crack has been initiated tube failure occurs quickly.

B. Accident Progression Three different accident scenarios should be considered in attempting to determine accident severity when degraded tubes rupture. (Basic concepts and design basis accidents are defmed in Appendix B)

(i) Steady State Operations. Due to random variations in local vibration intensity tube failure will be initiated in a single tube with relatively slow through the wall crack propagation. Such failures likely would be confined to a single tube lending to a timely detection and removal from service.

(ii) Operational Transients. Next in severity are tube ruptures from operational transient (opening or closing of valves, scrams, etc). To ensure that the tubes withstand such transients, the ASME code requires that their Cumulative Usage Factor eCUF") be less than one. When the RSGs were installed the predicted CUF was a low number because it was calculated for pristine tubes. The fact that the tubes were stressed to above their endurance limit and they vibrated with frequencies on the order of 10 to 50 HZ, means that their CUF exceeded unity. A tube with CUF > 1 would be prone to a rapid crack growth and sudden rupture under operational and DBA transients. Appendix B shows why such transients are important.

(iii) Main Steam Line Break ("MSLB"). An MSLB could lead to the most severe consequence for operations with degraded tubes. SCE's current licensing base

("CLB") requires that the plant accommodate such accidents. The MSLB accident is of particular concern to the type of tube failures that occurred at SONGS because of the 7

high excitation frequency that can generate as accounted by a witness at Turkey Point (Ref. 2),

"1 cannot explain my experience with the magnitude of these resonance vibrations, pressure, pulsation and the W' carbon steel they sheared, because no one can really understand this unless they have been there. 1 cannot relate how much more powerful this was than a full power scram because of the steam chugging vibrations."

More than 500 tubes were plugged at Unit 2. Even though these tubes are not in service they will continue to wear at the same rate as before. At a certain time the tube will break without being detected because no radioactivity will be released. The broken tube will have relatively low natural frequency and therefore it would be prone for resonance excitation by external forces. The jets, formed at the ends of a broken tube, would cause it to whip and impact adjacent tubes thereby propagating further ruptures.

While the operator could cope when only one tube ruptures he or she may not be able to do so if the ruptured tube causes several other tubes to rupture rapidly.

C. Failure ofRSG Tubes to Meet Performance Criteria Even though eight tubes failed to meet the performance criteria in Unit 3 tests, (Refs. 3,4) neither SCE nor the NRC discussed the significance of this event. This is the first time that a failure of so many tubes occurred in a US plant and therefore its significance must be well understood.

The technical specifications ("TS") specifY that steam generators cannot operate unless it can be demonstrated that structural integrity, operational leakage and accident leakage meet the specified requirements. The fatigue analysis in Appendix A and the failure of the eight tubes to pass the pressurization tests demonstrate the risk of not meeting TS requirements if either of the units was placed back in service. As discussed below the present performance criteria are not conservative for the tube type damage that was observed in the RSGs.

Performance criteria for accident leakage are intended to demonstrate that accident leakage will remain safe as determined by the stress limits in the ASME code.

The NRC developed a methodology, Regulatory Guide, RG 1.121, to demonstrate that performance criteria will be met between inspection periods. (Ref. 5) The need to consider tube degradation by fatigue is discussed on page 6, the effects of cyclic loadings forces must be considered in determining the minimum wall thickness including stress 8

concentration effects. To meet this requirement, the error in the MHI calculations as discussed in Appendix A must be corrected.

Prompted by RG 1.121 a large number of studies have been conducted to better understand the sequence of events during steam line breaks that could lead to core melt with containment by-pass also known as large early release, LERF. These studies culminated in a new performance based technical specifications requirements, which were issued in Aug. 2007(Ref. 6). The new TS specified safety factors (SF) of 3, 1.4 and 1.2 against failures at normal operation, DBAs and combined accident and non-accident loads respectively.

The 2007 SFs were derived to prevent radioactivity releases resulting from operations with tubes with undetected stress corrosion cracks. The formation of stress corrosion cracks (See) is not the cause of tube degradation at SONGS. At SONGS, the tubes degraded due to both reductions in wall thickness and fatigue life. There are similarities and differences between see initiated cracks and fatigue initiated cracks; both are characterized by distinct crack initiation and crack propagation periods. In the case of see, however, the incubation time is relatively short and propagation time is long thereby facilitating early detection. In contrast, in vibration fatigue cracks spend most of their life in incubation and rupture quickly giving the operator little time to take a preventive action. In both cases, however, when the applied stress exceeds a critical value the tube will rupture abruptly. The tube rupture at Indian Point in 2000 is an example where an undetected see crack in the u bend region propagated quickly to rupture when the slow movements of the vertical legs caused the stress to increase above the critical value.

The fact that 8 tubes at SONGS failed pressure tests has more serious implications than if those tubes had failed due to see cracks. Because of the low probability of early detection of fatigue crack failures, (due to a combined effect of wall thinning and fatigue), it is more difficult to prevent such failures than those resulting from see cracks.

Performance criteria, which were solely based on leakage from see cracks, are not conservative when the leakage originates from fatigue cracks.

In sum, the SONGS FSAR should be updated to reflect the conclusion of the present analysis (Appendix A) that some tubes have been degraded by vibration fatigue and the existing performance criteria are not sufficiently conservative to assure that that the tubes will retain their integrity during steam line break accidents. The updated SONGS FSAR must include an assessment of the safety implications of not detecting tube fatigue due to fretting and cyclic loads. The FSAR must show that the cumulative 9

fatigue factor, CUF, will be less than one for all tubes,as required by the ASME code and NRC regulations.

Contrary to NRC regulations and requirements, SCE has not demonstrated that at 70% power, Unit 2 tubes will not fail during operational and LOCA transients. .

3. Title 10 Code of Federal Regulations Part 50.59 Requirements Changes or modification of nuclear plants and test facilities are controlled by 10CFR 50.59 requirements. Changes can be made without a license amendment provided that they do not adversely affect the plant's FSAR. NRC and EPRI, Ref. 7, have formulated guidelines (NEI 96-07), which help with the determination of whether a license amendment is needed for a proposed change. These guidelines are not prescriptive and ultimately it is the responsibility of the applicant to defend its claim that the changes will not have an adverse affect on plant safety. NEI 96-07 discusses various aspects of the FSAR and points out how to specifically address design or safety evaluation methods.

The design ofthe RSGs differs from the original SGs by:

• Tube support plate (TSP) geometry from (broached to eggcrate)

• U Tube geometry (bend vs. square)

• Length and number of tubes, tube material and tube wall thickness.

• Original stay cylinders were replace by tubes

• Nominal tube/A VB gap and bar geometry

• Operating pressure

• Void fraction The above design alterations directly affect the thermal hydraulic performance of the steam generator on the secondary side, causingchanges in the axial and cross velocities as well as the void distribution. In addition, these alterations affect the individual natural frequency ofthe tubes and their supports. A person familiar with state of-the art steam generator design and operations should have recognized that these alterations would significantly affect tube vibration.

Since excessive vibration can result in a tube rupture, a cursory examination was sufficient to indicate that a license amendment under 10CFR 50.59 was required. A more formal way to illustrate this is to trace, step by step, the NRCIEPRl NEI96 -07 procedure by comparing the original FSAR prior to the installation of the RSGs with the updated 10

FSAR. Since an understanding of the role of thermal hydraulic ("TH")codes is central for such a comparison, a brief description ofTH codes is appropriate here.

TH codes are major evaluation tools to determine whether a steam generator will produce the required amount and quality of steam to drive the turbines safely. In addition to determining overall performance, TH codes are used to calculate local velocities, pressure, temperatures, and steam quality. The accuracy of such calculations is critical not only because they affect the overall SG performance but also because these local conditions determine whether the tubes will vibrate excessively, or promote corrosion and crud accumulation. TH predictions provide inputs to other codes that determine loads on various components during normal and accident conditions. TH codes rely on experimental inputs such as boiling correlations which are specific to a given steam generator and are proprietary. By the definition ofNEI 96-07, TH codes are classified as evaluation methods: the calculation frame work used for evaluating behavior or response of the facility or an SCC'~.

According to NEI 96-07 the design changes at SONGs must be considered together with changes in the method of evaluating the new design: "Changing from a method described in the FSAR to another method unless that method has been approved by NRC for the intended application" (3.4)

Thus any changes in the TH computer codes as described in the original SONGS FSAR would require that those changes be considered during the 10CFR 50.59 screening process.

The original steam generators were based on a TH computer code CRIB but the updated FSAR did not describe CRIB for analyzing overall performance. The design and performance of the RSG were based on an entirely different MHI code, FIT-lll; FIT-lll is not an NRC approved computer code and therefore its use would require a license amendment under 10CFR 50.59 as discussed above.

According to the NRC (Ref. 2) FIT-lll has not been benched marked and therefore its predictions are based on unknown uncertainties. Given the major design changes in the RSG and given the fact that MHI has no experience in designing and fabricating steam generators of the size of the subject RSGs, the reliance on FIT-lll represented a safety risk that should have been discussed in the FSAR. In conclusion, neither SCE nor MHI complied with the 10CFR 50.59 requirements.

When steam generators fail early in life it can be attributed to design or fabrication errors. It is most likely that the failure of the RSGs here resulted from 11

allowing the units to operate with excessive velocities and void fractions without adequate tube support. The common practice of reducing such errors is to conduct a thorough review by competent and independent experts during the design stage of the project. There is no indication that such a review was applied to the subject RSG units.

The design team was aware of data showing the possibility of in plane FE! but chose to ignore it because SCEIMHI has never seen such phenomena before.

Given the fact that MHI did not have experience in making such changes to SGs of that size, (Ref. 8), and were using a TH code which was not benchmark ed, the need for an independent review was that much more apparent.

The replacement of the stay tubes and the change in the TSP hole geometry affected the flow distribution in the center of the tube bundle. A competent independent expert, experienced with US reactors, would have likely recognized that the MHI changes could expose the tubes to high cycle fatigue from flow-induced vibrations.

The 10CFR 50.59 processes notwithstanding, SCE fails to demonstrate that the changes and modification to the original steam generators were conservative and did not present a safety risk to the public.

4. NRC Assessment The NRC Augmentation Inspection Team, AIT, (Ref. 3) briefly described how SCE addressed the 10CFR 50.59 requirements. The team discussed the NEI 96-07 guidelines but did not determine whether SCE complied with those guidelines and whether SCE was justified in not applying for a license amendment. The team attributed the rapid tube degrading to faulty FIT-lll predictions. The team noted that FIT-Ill was not benchmarked. The NRC has long recognized that benchmarking and peer reviews of analytical codes are necessary because they provide the means to establish the associated biases and uncertainties. The RELAP and TRAC TH codes have been benchmarked and reviewed by several peer review groups over a period of more than 30 years. As outlined below, many questions relating to the future of the subject RSG remain unanswered:

• The applicability of the August 2007 performance-based Technical Specifications to the type of tube degradations that were observed at SONGS. Should the new TS be modified to account for fatigue damage?

• Would the NRC conduct independent studies to specifically identify which of the changed component(s) resulted in the accelerated tube wear?

• What criteria would the NRC use to determine the ~ LREF when SCE submits a license amendment for the restart of Unit 2?

12

• .Which of the inputs to the FIT-lll code led to the predictions of low velocities and void fractions causing the parties to believe that the RSG contained sufficient margins to prevent vibrations? How do these inputs differ from those used in FIT-Ill for other designs where no vibrations were observed?

• The degree of fatigue damage suffered by tubes that were not removed from service (plugged) in the four RSGs.

• The degree of due diligence that SeE conducted prior to ordering the RSGs from MHI.

In conclusion, only an impartial and balanced assessment would maximize the lessons learned from the SONGS experience.

13

APPENDIX A - FATIGUE ANALYSIS The relative motion between the tubes and the anti-vibration bars (AVBs), the tube support plates, and the retainer bars result in tube wear and fatigue. This can produce relatively quick tube failures when the stresses generated during vibrations are sufficiently large.

As described in Attachment 4, MHI Document L5-04GA564, MHI used a fmite element model ("FE"), to calculate that the tubes were subjected to a stress of 4.2 Ksi (page 16-2), which is smaller than the endurance limit stress of 13.6 ksi (page 16-2).

Consequently, MHI concluded that the tubes would not fail from fatigue even if they were subjected to infinite number of stress cycles (page 16-13).

The MHI results are based on two erroneous assumptions. When these assumptions are corrected, as discussed below, the opposite conclusion is reached, which is that the tubes will be susceptible to failure from fatigue.

A. Stress Concentration It is a well-established fact that geometrical discontinuities such sharp comers introduce high local stresses, which act as a site for crack initiation. A common engineering practice is to fillet or chamfer sharp comers to reduce stress concentrations and increase fatigue life. A large database is available to guide designers in selecting the particular fillet for a given application. MHI used a design chart, (Figure 2), for a tube in pure tension to determine the stress concentration factor Kt. Assuming an undisclosed value for the fillet radius and the value of the parameter t ,MHI concluded that Kt was less than 1.5 when tlr .33. These numbers indicate that MHI used a value oftlh that exceed unity. Had MHI assumed smaller values for tIh, Kt would have exceeded 1.5 because Kt is sensitive to the assumed geometry of the fillet. MHI selected an arbitrary geometry, which is not valid, and for this reason only MHI obtained an unrealistically low value for Kt.

Figure 2 at page 17 is intended for applications when one is trying to minimize stress concentration. Visual examination of the contact between the AVB plates and the tubes do not suggest that the relative motion resulted in geometry with minimum stress concentration. On the contrary, as shown in Figures 3 and 4, the method in which the A VB interacted with the tubes allows for a formation of sharp comers at the intersection of the plate comer with the tube. MHI's own discussion is not consistent with their application of Chart 5 in Figure 2. The observation that the "tube and the AVB are worn into each other" and the fact that the AVB plate has sharp comers suggest that Chart 3.5 at page 17 does not apply to observed wear pattern.

The model shown in Figure 5 at page 19, represent more closely the wall-thinned geometry than that ofthe one used by MHI in selecting the stress concentration factor.

Since Figure 2 does not provide data for fillets with very small radius, it is necessary to 14

consider a similar geometry giving Kt values for small radii. In Figure 6 (a special case of Figure 2, when di = 0 ), Kt is plotted for very small radiuses for bending, (Kt values in tension are similar.)

Using a reported wear of35%TW, Kt is calculated as follows:

t = (1 %TW) T = (1-0.35) 0.043 0.028in dID 0/0-2t = 0.750 /0.750 - 0.056 = 0.750/0.694 = 1.08 Kt = 5 when r= 0.0014 for Kt r 0.002 (0.694) = 0.0014 (Theoretically the chamfer radius of a sharp comer is zero, and therefore Kt will tend to be very large For a finite but small radius of 0.0015, which is close to describing a sharp comer, Kt exceeds 5.)

B. Loss of Wall Thickness (wall thinning)

The effective wall thickness Teff. of the geometry in Figure 4 can be expressed as:

Teff t x (2 e )/360 + T x ( 360 -2 e )360 e = 2 Cos -1 ( dl2 +t )/(dl2 +T)

For a 35 and 70% TW wear, e equals 44.6 and 44.1 degrees respectively; the corresponding effective tube thickness equals 0.0360 and 0.0345, respectively.

C. Surface Finish Fatigue life and therefore the endurance limit is strongly affected by surface finish. Studies (Reference 11) show that fatigue life can be reduced by as much as a factor of 2 to 3 when a smooth surface is roughened to about 1.6 microns.

The data (Edison Attachment 6- Appendix 0, pages 130-131) indicates that the fretted tube surfaces do not maintain their original surface finish; instead they are severely scarred. Such scars are sites for the formation of micro-cracks.

Bounding calculations would require that the ASME design stress used by MHI (13.6 ksi) be lowered to account for surface finish. It is not clear, however, that the introduction of both a stress concentration and surface finish correction simultaneously would not be overly conservative. Since no data was found in the literature where both a sharp comer and adjacent rough surface, a surface finish correction was not included in the present assessment.

15

D. Corrected MHI Stress Calculation Corrected stress MHI stress x concentration correction factor C, x thickness correction factor Tc, = 4.2 x C x Tc C actual stress concern. factor / MHI concentration factor = 5/1.5 = 3.33 lITc Decrease in wall thickness/original wall thickness = 0.036/0.043 for beginning of cycle 0.0345/0.043 at the end of cycle assuming the same wear rate.

Tc = 1.19 to 1.25 Increase in stress= 4.2 x 3.33 x 1.19 to 4.2 x 3.33 x 1.25 = 16.7 to 17.5 Actual increase over the endurance limit = 16.7/13.6 to 17.5/13.6 = 1.22 to 1.29.

16

1. Purpose The purpose oIlhis document is to show that the slress 01 the b.be in SONGS RSG die to in-plane viJration is I.I1der the fatiglE lmit

2. Conclusions The stress on !he tU:Je die to n.pIane 'Iribraticn is 4.2ksi and is tnder fatiglE runit (13.6ksI).

The b.be has sbu:lu"aI integrity foc the sb"ess die to il-plane viJration from the view poD: of fati~ evaluation.

3. Assumptions and Open Items The tU:Je defonns in-pIane until contacting wth the Cllier next lube in Row direction dlE to in-plane vhation.

The stress dlE to in-plane vibration is high cycle fatigue

4. Acceptance Criteria The faliglE limt is 13.6k.si aa;oning to the foIawing design faligue Ctne.

~ h.

\, ~ c.-A

~

~~

in 100...

,t_IJ 14 """'r. ~c Figu-e 4-1 Design Fatigue Cuve for Tube Figure 1 Fatigue data used by MHI to determine tube fatigue life. The cycle independent line represents the endurance limit. MHI used an endurance limit of 13.6ksi. Attachment

4. page 16-2. Data for smooth specimen.

17

e><ARTS 157 2.2 (d~ ", ). d;ft >2~ tlhl-OJS ,/'" ~

I V ~

2.0

./

V rOo

~

.....- i---~

~i

/ ./' ."t~.oo Kt 1.8 1.6

~

i Y ,..V

/ ,/'"

/. / ' / "

v:. / ' -----

~ - / L..-- i---

-~

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-=t~_ ~.~ =:- onL -'

1.4 r? ,,"max

~~

r-rt~I K.~-

p p

"""om 1.2 J. ['l'" .. p

/1 t T IM!fi+1>

if i 2 3 4 tlr Chart 3.5 Stress Con<:ellttatioo factors K, for a tube in tension with fillet (Lee and Ades 1956; ESDU 1981).

Figure 2 Stress concentration factors used by MHI for calculating maximum tube stress, (pageI6-2 )Source: W.D. Pilkey, Peterson's Stress Concentrations Factors, John Wiley and Sons 1997.

18

3.2 wear Patlem-2 (local Wear on TUbe Surface)

Characteristics (j) local wear occurs on the tube but the wear surface Is not exposeo (cannot be seen)

(2) Unable to determine if wear occurs on bi>e or AVB or both

@ Unable to determine the direction of motion c:r vibration

@ An extreme intel'JX'etaUon Is that both bi>e and AVB are worn into each other.

(a) case 1 (b) Case2 Fig-2 Wear Pattern 2 Figure 3 Wear due to AVB/tube Interaction - Attachment 4.1t should be noted that both the impact and the sliding motions playa part in the tubelAVB interaction. These factors reduce tube strength because of material loss but also because of loss of fatigue strength.

19

Non-proprtetary Version I

)(P.10-20)

____________________ Doc_ume:+,:lS-04GAS64(9) v Fig. 6-4 Wear shape of tube at the contact point with AVB Figure 4 MHI description of wear shape at tube/AVB contact point. Attachment 4, page 10-20.

20

T = original wall tbidmes d=innerdia Figure 5 Schematic for determining a stress concentration factor Kt and reduced wall thickness of a tube due to double sided wear to a thickness t over an arc defined bye 21

166 sttOUlDER FIU.EfS Dld Cblll't 3.1 t &i:css concentratioo fso:1or5 K.. for beDding qf a stepped 00r of circlliar a{.)s~ scction wilh" ,~OOI,lkll.'T litl". <I:1_u ('n pI1"IOI:!Ii~i"I"""l~ Qr U-V1:'1l <m<;I!l;trlm,1n 19:1<1; Wiln'lI.1.11d While 11)73). This chan SCI"'CS 10 SUPplcffil!'llt Chttrt JJO.

Figure 6 Stress concentration used in the present analysis. W.D. Pilkey, Peterson's Stress Concentrations Factors, John Wiley and Sons 1997. (Similar Kt values in tension Peterson's Chart 3.4, and for internally pressurized vessel Chart 3.6 for a small radius.)

22

APPENDIX B - BASIC CONCEPTS To maintain their current licensing base (CLB) operators must show that steam generators with degraded tubes will maintain their integrity during certain hypothetical accidents (DBAs). The degree of safety is determined by the amount of radioactivity (dose) that would be released to the environment following the accident. Typical DBAs are steam generator tube ruptures, ("SGTRs") and main steam line breaks ("MSLBs").

In addition, severe accidents, which can be initiated by station blackouts (SBOs) or transients without a scram (ATWS) must also be considered because they impose immense stress on steam generator tubes. Weakened tubes such as those found at SONGS would not be able tolerate such stresses. In fact three tubes at Unit 3 failed the steam line break pressure test.

DBAs for PWRs are defined as:

1) SGTR: spontaneous rupture of a single SG tube with a concurrent failure of a single active component; 2)MSLBwith primary to secondary leakage within TS limit and with a concurrent failure of a single active component.

Tube ruptures in steam generators can be also induced by other events that increase the ~p across the tube, i.e.,secondary and primary side depressurization (stuck MSL and Safety Valves, SG overfill, seam dump controls, ATWS initiated events on the primary side).

As a result of operating Units 2 and 3 with defective steam generators the two units were operating beyond their licensing base exposing the public to a significant risk from large radioactivity release due to core melt and containment bypass. The magnitude of the risk depended on the amount of primary to secondary leakage which would have leaked had a steam line broke during plant operations and the inability of the operator to stop the leakage. In my opinion the probability is less than 0.1 that the operator would have been able to control a leakage equivalent to more than five tube ruptures. In comparison, about 170 tubes in each steam generator of Unit 3 were susceptible to failure during a steam line break accident.

Existing SG programs do not require the removal from service of steam generator tubes with shallow defects. Shallow defects may reduce fatigue life when the tube is operating in an FEI environment and therefore possesses a high probability of tube leakage during severe accidents. To comply with 10 CFR 50 Appendix B, Criterion 16, SCE must remove from service all SG tubes that potentially could lead to a large early release (LERF) during a severe accident.

A conservative estimate of the probability of a major accident can be made by assuming that a steam line break will occur and because of the large ensuing primary to secondary leakage, the core will be uncovered. The probability of the occurrence of a steam line break is 10-4 per year. then conservatively, without giving any credit to the 23

operator, this is also the probability of a core melt with a containment by-pass That means that during any a cycle at SONGS, the probability of a core melt with a large early release is 1 in 10,000 years.

r(tl:llOT(1t I NLET NOZZLE TUBE SHEET NlloiARY [NLE:T r~CtoiI THt REACTOR GORE Figure 7 Essential Features of a SONGS Steam Generator, Ref .1 24

REFERENCES

1. SONGS Return to Service Report, October 3,2012

2. Attachment 4- MHI Doc.L5-04GA564; Attachment 2- Appendix B, AREVA; Attachment 5--MHI Screening criteria; Attachment 6 - Appendix D, Westinghouse

3. R. Spencer, "Resonance Vibrations In Steam Generator Tubes During Steam Line Break Depressurization (ML003726262)

4. San Onofre Nuclear Generation Station- NRC Augumentation Inspection Team Report 05000362/2012002010, July 18,2012 and November 9, 2012

5. San Onofre Nuclear Generating Station, Units 2&3 Root Cause Analysis UES 20120254 Rev. 0

6. Bases For Plugging Degraded PWR Steam Generator Tubes. Regulatory Guide 1.121 August 1976

7. NRC Regulatory Issue Summary 2007-20 "Implementation of Primary to Secondary Leakage Performance Criteria" August 23,2007.

8. Guidelines for 10 CFR 50.59 Implementation, NEI 96-07 Rev 1, Nov. 2000

9. Hitoshi Kaguchi, "MHI's Steam Generator Operating Experience with Tube Vibration and Wear." Feb. 7,2013

10. NUREG/CR-366I Prototypical Steam Generator 9 MB-2 Transient Testing Program. Scaling Analysis Report. March -1984

11. PWR Steam Generator Tube Fretting and Fatigue Wear, EPRI- 6341 April,1989

12. NUREG ICR -6909. Effect ofLWR Coolant Environments on the Fatigue Life of Reactor Materials 25

Joosten, Sandy From: Vi nod Arora [vinnie48in@gmail.com]

Sent: Wednesday, April 10, 201312:50 AM To: CHAIRMAN Resource; Borchardt, Bill; Leeds, Eric; Howell, Art; Dorman, Dan; Benney, Brian; Hall, Randy; Lantz, Ryan; R4ALLEGATION Resource

Subject:

San Onofre NRC/SCE/MHIIPublic Awareness Series HAHN Baba April 10,2013 at 12:42 am Your comment is awaiting moderation.

Sincere Thanks to NRC Chainnan, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

San Onofre NRC/SCEIMHIIPublic Awareness Series - by Hahn Baba Steam generator tubes leak and rupture despite inspection and plugging. Question is can the operator detect the leak and shutdown the plant. That is a 95% probability of success and 5 percent chance of failing. There have been more than 20 leaks and tube ruptures in the last 30 years in USA with timely detection and shutdown with no reported offsite releases affecting the public or exceeding federal limits. San Onofre MHI generators, I cannot say .. Even at 70% power, there is a much higher chance of multiple leaks and tube ruptures in Unit 2 due to manufacturing defects, cracked and plugged tubes, operational transients, mother nature's mood, equipment malfunctions and operator errors. SCE has to either satisfy NRC, Dr. Joram Hopenfeld and the Public about safety of Unit 2, repair or replace the defective generators or decommission the plant. So far SCElMHIIIntertek, Westinghouse and AREVA response is totally unsatisfactory and unconvincing. Customer service and safety always prevails over SCE profits, if NRC and CPUC rules are followed to the Letter. .. Let us see what NRC Commission does to satisfy 8.4 Million southern Californians, who pay the bills for San Onofre and SCE Management.

1

Joosten, Sandy From: Vinod Arora [vinnie48in@gmail.com]

Sent: Thursday, April 11 ,2013 11 :08 AM To: CHAIRMAN Resource; Borchardt, Bill; Leeds, Eric; Howell, Art; Dorman, Dan; Benney, Brian; Hall, Randy; Lantz, Ryan; R4ALLEGATION Resource

Subject:

NRC's Paper Thin Transparent Safety Umbrella Crushed by Ted Craver's Political & Financial Hurricane - Regulatory Capture - Shame - 8.4 Million Concerned Southern Californians have no Rights in Democratic America Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

San Onofre NRC/SCE/MHIIPublic Awareness Series - Please excuse me for any computer or human performance grammatical or spelling errors NRC Proposes No Safety Threat Finding With San Onofre. SCE once again finds shelter in NRC Commission's favorable Findings. NRC Commission's Finding is in direct conflict with His Excellency President of the United States, Senator Boxer and Representative Edward Markey Open Doctrine and violates the Trust, Rights and Safety of 8.4 million Southern Californians HOT NEWS - April 11 ,2013 - Edison International (EIX)'s request for a license amendment for a proposed restart of its crippled California nuclear reactor doesn't pose significant safety risks, federal regulators said in a preliminary finding. Edison's request to operate its San Onofre reactor at reduced power does not involve an increased risk of an accident or create the possibility of a new or different accident from those previously evaluated for its license, the U.S. Nuclear Regulatory Commission said in an e-mailed document. The NRC may approve Edison's license amendment before the 60-day waiting period normally required after a notice is published if no hazards are found, the agency said.

The public will have 30 days after the finding is published in the federal register to comment on the NRC's conclusion before the agency makes a final determination. The document will be published next week, said Victor Dricks, a spokesman for the NRC. Without a finding of a significant hazard, the NRC can issue a license amendment before holding a public hearing, Shaun Burnie, director of nuclear campaigns at Friends of the Earth, said in a telephone interview. "It's a get out of jail free card," Burnie said. Edison didn't immediately respond to a request for comment.

An approval of the amendment won't allow the plant to restart, the NRC has said. The agency will make a separate determination of the utility's restart plan and it may be challenged in meeting Edison's June deadline, Doug Broaddus, chief of the San Onofre special-projects branch for the NRC, said at an April 3 meeting.

Senator Barbara Boxer, a California Democrat, faulted the agency's preliminary finding. "The NRC staff proposal, which could pave the way for the restart of the San Onofre nuclear power plant before the investigations of the crippled plant are completed, is dangerous and premature," Boxer, chairman of the Environment and Public Works Committee, said today in a statement. She also said the plant is in an area that's at risk of an earthquake and tsunami. Representative Edward Markey of Massachusetts, the top Democrat on the House Natural Resources Committee, joined Boxer in criticizing the agency. Sen. Barbara Boxer blasted a preliminary finding by Nuclear Regulatory Commission staff that restarting a unit at the idled San Onofre Nuclear Generating Station would not present significant safety hazards. Boxer and Rep. Edward Markey, D-Mass, sent a letter to the NRC this week demanding that a comprehensive investigation of the plant be completed before any units be permitted to operate. They also said full public hearings should be held before a decision is made.

~oxer said in response to the commission staff's preliminary finding. "It makes absolutely no sense to even consider taking any steps to reopen San Onofre until these investigations look at every aspect 1

of reopening the plant, given the failure of the tubes that carry radioactive water. "In addition, the damaged plant is located in an area at risk of earthquake and tsunami," she said. "With 8 million people living within 50 miles of this plant, the staff proposal is beyond irresponsible." On Monday, Southern California Edison announced it had formalized a request to amend its operating license to allow it to operate its Unit 2 reactor at 70 percent beginning June 1.

Summary of

Conclusions:

Based on in-depth review of academic literature, all available San Onofre Reports and Tube Inspection data, and industry operating experience, it is concluded that in SONGS Unit 2 defectively-designed and degraded steam generators, even at 70% power normal power operations with proposed amended SCE License, multiple double-ended rupture of steam generator tubes can occur at any time due to anticipated operational transients and main steam line break accidents. These steam generator tubes ruptures caused by fluid elasticity in-plane tube-to-tube wear and undetected and un-quantified incubating macroscopic high cycle thermal fatigue circumferential cracks rupturing the active, worn and plugged tubes into two pieces at tube-sheets, tube-support plates, mid and free spans not supported by anti-vibration bars can result in large radiation leaks due to 100% tube-bundle uncovery, flashing sub-cooled feedwater into high dry steam and jet impingement. The steam generator with multiple tube ruptures leads to the conclusion that the steam generator could be full of high dry steam for a significant period of time. The amount of iodine released from the ruptured steam generator could be significantly larger than that previously calculated. A potential radiological accident would transport with steam 100% of the iodine contained in the 15,000 gallons of reactor coolant to the environment exceeding the federal regulations within 10 minutes during an extremely fast-paced transient beyond the operator control and failure of defense-in-depth actions. Core Damage Probability (CDP) and Large Early Release Probability (LERP) associated with multiple double-ended rupture of steam generator tubes will significantly increase than previously approved NRC SONGS FSAR Limits. The resulting doses would be significantly higher than the dose consequences analyzed in the SONGS UFSAR for the post-trip SLB event with a concurrent iodine spike. The postulated post-trip SLB with multiple tube ruptures and concurrent iodine spike Exclusion Area Boundary, Low Population Zone, and Control Room doses would be significantly higher than the post-trip SLB Control Room limit of 5 Rem TEDE, and the Exclusion Area Boundary and Low Population Zone limit of 2.5 Rem TEDE.

Incubating macroscopic circumferential cracks caused by fluid elastic instability, flow-induced random vibrations and high cycle thermal fatigue are extremely difficult to detect and be accurately sized by nondestructive evaluation techniques like X-ray, ultrasonic, and eddy current based bobbin coil probes, mechanically rotating pancake coil (RPC), etc., which have been used in 17,000 SONGS Tube inspections. State-of-the-art systems: Zetec MIZ-80 iD system, Tecnatom TEDDY+, Circular TE and TM, transmit-receive eddy current array probe C-3 and other specialized radiographic probes capable of detecting sub-surface cracks caused by high cycle thermal fatigue have not been used in the 170,000 SONGS Tube Partial and Limited Inspections as shown below for Unit 2 due to access problems in the most problematiC inner-most sections of the U-Tube Bundle, the high-cost, lack of availability of highly specialized tools and contractors, radiation doses, and time considerations in a rush to start Unit 2. The inspection scope defectively designed and degraded SONGS Unit 2 RSGs should have covered 100% hot leg and cold leg tube inspections, 100% of dents or dings, 100% of tube inspections in the tight radius U-bends, 100% area of the Top of the Tube Sheet and Tube Support Plates. None of the SCE Global Experts agree amongst themselves with the cause of tube to-tube wear in Unit 2 and have not addressed the combined synergic effects of tube-to-tube wear and high cycle thermal fatigue cracks in their voluminous 2000 page documents. There are basic errors, invalidated assumptions, incorrect benchmarking of Unit 3 and blunders in Westinghouse, MHI, SCE, Intertek, AREVA and NRC AIT Reports. The combined effects of tube-to-tube wear and high cycle thermal fatigue cracks have been witnessed by sudden tube ruptures in North Ana in 1987 (See Item 7 below), MHI SG in Mihama, Japan in 1991 (See Item 8 below), three tube leakages in 2

French SGs between 2004 through 2006 (See Item 9 below), 20 tube ruptures/leakages in SGs between 1980-2000 in USA, and SONGS 3 in 2012. If SCE is so confident and conservative about safety and prudent actions of their Global Experts, why not provide Independent Experts with the Units 2 & 3 Cycle 16 notarized operational, contact force, tube fatigue analysis and details of tube Inspection data? Independent Experts can be persuaded in interest of public safety to certify beyond NRC validations, whether SCE and their Global Experts are right or wrong about safe claims of restart of Unit 2?

1. SCE states, "Rosemead, Calif. (Dec. 18, 2012) - The Mitsubishi Heavy Industry testing under review by the Nuclear Regulatory Commission (NRC) was not consulted or relied upon in developing Southern California Edison's (SCE) proposed restart plan for Unit 2 - a plan which includes preventive tube-plugging and operating the unit at 70 percent power for a five-month period. SCE's international team of experts conducted more than 170,000 inspections to understand the tube wear problem, and confirmed the effectiveness of the corrective actions we have identified to solve the tube wear problem. This work included three independent operational assessments of tube wear issues conducted by Areva Inc, North America, Westinghouse Electric Company LLC and InterteklAptech , none of whom based their review and recommendations on Mitsubishi's testing. This was confirmed Tuesday by an NRC administrator at an NRC public meeting in Rockville, Md. SCE submitted technical information to the NRC on Oct. 3 in support of a proposed restart of Unit 2, which is safely offline. The unit will not be restarted until all plans have been approved by the NRC. The Unit 3 restart was not included in that regulatory filing and remains shut down."

2. SeE SONGS Unit 2 Return to Service Report, Enclosure 2, Section 6.1, "Summary of Inspection Results", page 22 states, "This section provides a summary of the different types of tube wear found in the SONGS Unit 2 and 3 SGs. Wear is characterized as a loss of metal on the surface of one or both metallic objects that are in contact during movement. The following types of wear were identified in the SONGS Units 2 and 3 SG tubes:

• AVB wear - wear of the tubing at the tube-to-AVB intersections

• TSP wear - wear of the tubing at the tube-to-TSP intersections

• TTW - wear in the tube free-span sections between the AVBs located in the U-bend region.

• RB wear - wear of the tubing at a location adjacent to a RB (RBs are not designed as tube supports for normal operation)

• FO wear - wear of the tubing at a location adjacent to a FO.

Most of the tube wear identified in the SGs is adjacent to a tube support. Figure 6-1 is a side view of an SG, showing the relationship of the tubes to the two types of tube supports: TSPs in the straight portions and AVBs in the U-bend portions of the tubes. All tubes are adjacent to many of these two types of tube supports. The RB supports are not shown because a very small number of tubes are adjacent to them. TTW indications occurred in the free span sections of the tubes. The "free span" is that section of the tube between support structures (AVBs and TSPs shown in Figure 6-1). TTW occurred almost exclusively in Unit 3 and is located on both the hot and cold leg side of the U-tube. In many cases, the region of the tube with TTW has two separate indications on the extrados and intrados of the tube. The wear indications on neighboring tubes have similar depth and position (ranging from 1.0 to 41 inches long and 4% to 100% through wall) along the U-bend, confirming the tube-to-tube contact. Table 6-1 provides the Wear Depth Summary for each of the four SGs based on eddy current examination results. Detailed results of the examinations performed are provided in the Units 2 and 3 CM reports included as Attachments 2 and 3. Figures 6-2 through 6-5 provide distributions of wear at AVB and TSP supports for all four SGs."

3

3. Attachment 2 by AREVA states, 'The SONGS Unit 2, 2C17 inspection scope occurred in three distinct phases. The first phase followed the planned shutdown for the 2C17 refueling outage and first SG lSI. The next two inspection phases, performed in April and July 2012, were a direct result of a SG tube leak in Unit 3. The tube leak resulted from tube-to-tube wear (TTW) that was caused by fluid-elastic instability. These subsequent inspections are referred to as 2C 17 RTS (Return-to Service) inspections. The second-phase inspection (April 2012) was a full-length U-bend inspection of

~ubes deemed most susceptible to tube-to-tube wear based on the degradation identified in Unit 3.

The third-phase inspection (July 2012) consisted of eddy current testing to measure the gaps between the AVBs and the tubes. Based on the gap measurements, an additional 104 tubes were examined in the U-bend region with the +PointTM coil. Inspections included the following inspection activities for each of the two replacement steam generators (SG 2E-088 and SG 2E-089) using site validated ECT techniques [7]:

• Bobbin Coil Examinations - All in-service tubes, full length tube-end to tube-end - 19,454 Inspections

• The review showed that the +PointTM probe had a slightly improved Probability of Detection (POD) over the bobbin coil. Rotating Coil Examinations - 5492 Inspections

• Tubesheet periphery and divider lane tubes (from 3" above to 1" below the top of the tubesheet), both legs, approximately 3 tubes in from the periphery and 2 tubes in from the divider lane - 4120 Inspections

• Hot Leg, Cold Leg, U-Bend Coil 2, Special Interest and Mag Bias - 377

• Secondary Side Visual Examinations

>> Post sludge lancing FOSAR examination at the top-of-tube-sheet (periphery and the divider lane)

>> Visual inspections of the upper tube bundle at the 7th TSP and AVB / retainer bar regions"

4. Attachment 3 by AREVA is only applicable to unit 3, therefore are not credited.

5. MHI inspections, although very useful per SCE Claims have not been reviewed and not credited.

6. It is assumed, that Westinghouse and Intertek are using the inspections performed by AREVA summarized in item 2 above.

7. SCE has not determined the true cause of Unit 3 Tube-to-Tube wear as required by NRC Confirmation Letter Action 1, which states, "Southern California Edison Company (SCE) will determine the causes of the tube-to- tube interactions that resulted in steam generator tube wear in Unit 3, and will implement actions to prevent loss of integrity due to these causes in the Unit 2 steam generator tubes. SCE will establish a protocol of inspections and/or operational limits for Unit 2, including plans for a mid-cycle shutdown for further inspections."

8.0 - NRC Information Notice No. 88-31: May 25, 1988, Steam Generator Tube Rupture Analysis Deficiency Description of Circumstances: On July 15,1987, a steam generator tube rupture (600 gallons per Q'linute) event occurred at North Anna Unit 1 shortly after the unit reached 100% power. The cause of the tube rupture has been determined to be high cycle fatigue. The source of the loads is believed to be a combination of a mean stress level in the tube and a superimposed alternating stress (The mean stress is produced by denting of the tube at the uppermost tube support plate, and the alternating stress is the result of out-of-plane deflection of the U-bend portion of the tube above the uppermost support plate, caused by flow-induced vibration). Denting also shifts the maximum tube bending 4

stress to the vicinity of the uppermost tube support plate. The rupture extended circumferentially 3600 around the tube. Based on available information, the staff concludes that the presence of all the following conditions could lead to a rapidly propagating fatigue failure such as occurred at North Anna: (1) Denting at the upper support plate, (2) A fluid-elastic stability ratio approaching that for the tube that ruptured at North Anna, and (3) Absence of effective AVB support.

Following the steam generator tube rupture at North Anna Unit 1 on July 15, 1987, the Virginia Electric and Power Company (VEPCO) modified the flow resistance of the steam generator downcomers at North Anna by the addition of flow baffle plates. This modification necessitated the reanalysis of certain design basis events including rupture of a steam generator tube. The new analysis utilized a revised Westinghouse method for calculating steam generator water mass and indicated that during the event, the water level on the secondary side could fall below the top of the.

steam generator tubes for a 10-minute period at the beginning of the event. '

Steam generator tube uncovery is significant because, if the break location becomes uncovered, a direct path might exist for fission products contained in the primary coolant to be released to the atmosphere without partition with the secondary coolant. VEPCO and Westinghouse reanalyzed the design basis steam generator tube rupture accident for Surry using the revised methods and determined that the steam generator tubes at Surry could also become uncovered even though the Surry plants were not modified by the addition of flow baffle plates.

The licensee further concluded that the offsite dose consequences exceeded those calculated in the Surry Updated Final Safety Analysis Report (UFSAR) because tube uncovery could produce a direct path for fission product release. Based on the Surry results, the analysis of steam generator inventory during a steam generator tube rupture at other plants may show that the steam generator tubes may uncover. Thus, for those plants where the steam generator tubes were thought to remain covered following tube rupture, the previously calculated safety analysis offsite doses might be exceeded and since the primary coolant activity limit in Technical Specifications is based upon the occurrence of this accident, the allowable technical specification limit may be too high.

Ipiscussion: A postulated steam generator tube rupture is one of the design basis accidents analyzed ih plant Safety Analysis Reports (SARs). Using conservative assumptions of single failure and loss of offsite power, it must be shown that the offsite dose consequences will be limited to the guideline doses of 10 CFR 100 or a fraction of the guideline doses depending on the assumptions made for iodine spiking. The iodine in the reactor coolant may be previously dissolved from allowable operational fuel failure or may result from an iodine spike which is the sudden increase in coolant iodine concentration produced by the transient conditions during the accident. Mechanisms for transport of the iodine that exits the reactor system to the atmosphere are discussed in Standard Review Plan (NUREG-0800) Section 15.6.3. In determining the amount of iodine that is transported to the atmosphere, credit may be given for "scrubbing" of iodine contained in the steam phase and in the atomized primary coolant droplets suspended in the steam phase for release points which are below the steam generator water level.

The Surry UFSAR assumed that the break is always covered with water so that 99% of the iodine would remain within the steam generator coolant and only 1% would be released through the atmospheric relief valves. The break location is assumed to be always covered in the UFSAR calculations because an initial steam generator water mass that may be non-conservatively large was assumed in order to conservatively account for the possibility of overfill and because steam generator tube failures were thought only to occur close to the tube sheet. The North Anna tube rupture demonstrated that steam generator tube failures can occur near the top of the tube bundle. The 5

revised steam generator water mass calculations by Westinghouse with the assumption that the break occurs at the top of the tube bundle led to the conclusion that the break could be uncovered for a significant period of time. Tube uncovery occurs because of the level shrink that accompanies reactor trip/turbine trip during the tube rupture event. The tubes would be recovered by the flow of auxiliary feedwater into the ruptured steam generator and by the reactor coolant which would be added due to the ruptured tube; however, the amount of iodine released from the ruptured steam generator could be larger than that previously calculated.

9.0 - MHI SG Tube Rupture Mihama, Japan 1991: On February 9th, 1991, a heat transfer tube (SG tube) in a steam generator of the No.2 pressurized water reactor at the Mihama nuclear power station of the Kansai Electric Power Company broke off during a rated output operation. As a result, about 55 tons of primary cooling water leaked out from the SG tube into the secondary cooling loop, and the reactor was scrammed by operation of the ECCS (Emergency Core Cooling System). The failure of the SG tube was caused by fretting fatigue resulting from contact of the SG tube with the supporting plate for the SG tubes, because the AVB, which functions to prevent flow-induced vibration, was not inserted deep enough onto the SG tubes in the steam generator. The scale of the accident was ranked "level 3" on the international nuclear events scale (lNES). At 13:40, an alarm of a condenser air off take system went off during a rated output operation, warning that the coolant water level in the steam generator was decreasing. At 13:50, an automatic emergency shutdown of the reactor was triggered by the signal of decreasing pressure in the pressurizer. After seven seconds, the ECCS was automatically operated, and coolant water was flooded into the reactor by ~

high pressure injection pump. However, one main steam isolation valve and one pressurizer relief "

valve could not be operated by remote control. Therefore, the valve operation was carried out manually. The failed tube was removed from the heat exchanger, and the fracture surface was examined by a scanning electron microscope. Striations, which are a characteristic of fatigue failure, were observed on large portions of the fracture surface, and dimples showing tensile fracture were also observed. However, few traces of stress corrosion cracking and corrosion were found on the fracture surface of the tube. The failure of the tube was, therefore, hypothesized to be due to cyclic loading. The morphology of the fracture surface of the SG tube shows a typical example of the striations formed on the fracture surface of the SG tube. Examination of the other SG tubes near the failed tube showed traces of wearing formed by fretting due to contact between the tubes and the anti-vibration bars on the outer surfaces of the tubes. Stress amplitude of the failed tube estimated based on the striation spacing was found to be in the range of around 51 to 60 MPa. (8.4 Ksi).

Occurrence of cyclic loading in the SG tube that had failed was related to the insertion depth of the anti-vibration bar, AVB. The SG tubes were subjected to vibrations due to the flow of secondary coolant outside of the SG tubes. In order to prevent the flow-induced vibration, V-shaped AVBs were installed onto the opposite U-bent SG tubes near the upper part of the steam generator. However, the insertion depth of the AVB for the SG tubes was not enough, because the engineers who installed the AVB did not fully understand the importance of the AVB. In fact, no damage was founding in the SG tubes into which the AVB were inserted to sufficient depth as shown by the design guidelines.

Accordingly, the SG tubes were subjected to flow-induced vibration and strongly contacted with the sixth supporting plate, so that the SG tubes incurred damage by fretting fatigue. Inspection of the AVB had not been carried out since installation. In order to provide an opportunity to learn from the accident that resulted in the leakage of primary coolant from the SG tube due to fretting fatigue, the damaged steam generator has been preserved in an exhibition at the Mihama station of the Kansai Electric Power Company. An exhibition is a good way to help everyone to good lessons from an accident. This accident was the first disaster in Japan that resulted in actuation of the ECCS due to leakage of primary coolant in the steam generator. Therefore, the accident caused social concern with nuclear reactors. The international nuclear events scale (lNES) is defined by the IAEA to assure coherent reporting of nuclear accident by different official authorities. The INES is characterized from level one to level seven. The level number increases with the scale of the accident. For example, 6

level one is a minor event, and level seven is major accident. The scale of the accident in 1979 resulting in the loss of coolant that occurred in Three Mile Island was ranked level five by the IAEA.

The accident reported here was ranked level three.

10. Cruas NPP: Between 2004 and 2006, three primary-to-secondary leaks occurred at the Cruas NPP: unit 1 in February 2004 and unit 4 in November 2005 and February 2006. The three leaks were all the result of a circumferential crack in the tube at the location where the tube passes through the uppermost tube support plate (TSP #8). Analyses carried out by EDF, further to the last two events, resulted in them being attributed to high cycle fatigue of steam generator tubes due to flow-induced vibration. The results of in situ examination initiated by the Cruas NPP operator showed that the flow holes of the uppermost Tube Support Plates (TSPs) were partially or completely blocked by corrosion products. This phenomenon is referred to in this paper as TSP "clogging-up" and it was considered potentially generic for EDF NPP fleet. For the Cruas leakages, it was established that the association ofTSP clogging-up and the specificity of the Cruas steam generator (central area in the tube bundle where no tubes are installed) were responsible for a significant increase in the velocity of the secondary fluid in the tube bundle central area. The high velocity of the fluid in this region increases the risk of fluidelastic instability for the tubes. A new fact caused the leakages observed on the Cruas units: the heavy build-up of deposits on the secondary side of the steam generator which changed the flow conditions in the center of the tube bundle. The deposits reduced or blocked water/steam flow through the quadrifoil-shaped holes in TSPs, forcing more water and steam into the center of the tube bundle, which caused the excessive vibration of the tubes near the center of the tube bundle. This excessive vibration due to fluid-elastic instability resulted in fatigue cracking of the tube.

7

Joosten, Sandy From: Vinod Arora [vinnie48in@gmail.com]

Sent: Thursday, April 11 , 2013 3:04 PM To: CHAIRMAN Resource; Borchardt, Bill; Leeds, Eric; Howell, Art; Dorman, Dan; Benney, Brian; Hall, Randy; Lantz, Ryan; R4ALLEGATION Resource

Subject:

Direct Response Requested From Honorable NRC Chairman, Dr. Macfarlane from a concerned American Citizen HAHN Baba April!!' 2013 at 2:58 pm Your comment is awaiting moderation.

Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

San Onofre NRC/SCE/MHIlPublic Awareness Series Please excuse me for any computer or human performance grammatical or spelling errors.

Good Morning

Dear Dr. Macfarlane,

I was in a meeting with Ted Craver few weeks ago. He was surrounded by Powerful Politicians, Bank CEO's, Mayor's and other Government Officials. He told the audience that Edison has monopoly franchise agreements with cities in Southern California to supply electric power. Edison owns all the transmission and distribution systems in Southern California with an investment of $20 Billion. According to the press reports, there is no shortage of power or problems with grid stability in Southern California. It appears that Edison controls CPUC and ISO; therefore, Edison will stay in the rate-base and keep supplying its generated or borrowed power to Southern California even with San Onofre Unit 2 Shutdown. Edison will continue to make money even with both San Onofre Units Shutdown, until the defectively designed and degraded RSGs are repaired or replaced. In the end, Edison will recover all the money from ratepayers, Insurance Companies and MHI.

The question is why NRC Commission is sticking out its neck for San Onofre, which is an INPO 4 Plant, with the worst regulatory, safety, emergency preparedness, fire, cyber security, retaliation, discrimination, harassment, management and maintenance record. I am saying all this based on my firsthand knowledge as an ex-San Onofre employee, public safety expert and my observations/work as a high-energy line break, fire and emergency preparedness engineer and auditor. I am sure, that NRC will not go through so much public opposition defending another utility with such bad nuclear safety record as San Onofre. SCE Management to cover their own mistakes in the design of replacement steam generators are using their Powerful PR Machine, which is ridiculing Honorable Senator Barbara Boxer, Honorable Representative Ed Markey, Friends of the Earth, Dr. Joram Hopenfeld, Arnie Gundersen, John Large, Professor Daniel Hirsch, MHI, pressuring News Reporters not to publish anything adverse and paying Union Leaders to show up at Public Meetings.

NRC commission has not completed review of SCE Unit 2 Return To Service Reports, SCE Response to NRR RAI's, DAB Safety Team, Honorable Senator Barbara Boxer, Honorable Representative Ed Markey, Friends of the Earth, Dr. Joram Hopenfeld, Arnie Gundersen, John Large, Professor Daniel Hirsch' Allegations, NRC san Onofre Special Tube Inspections and yet NRC Commission had indicated approval of SCE' s Special License Amendment designed for subverting NRC Regulatory process, public participation and legal challenges from Public Safety Experts and Attorneys. From NRC's Commission eagerness and hasty actions, I conclude that NRC is not using its authority properly and using due diligence as an Independent Regulatory Commission in interest of Public Safety. This type of irresponsible NRC behavior directly conflicts and erodes Public Confidence in Public Statements made by you for the NRC Commission's Safety and Oversight Charter. It gives the clear and undisputed perception that NRC has not learnt any lessons from Three Mile Island, Browns Ferry, Davis-Besse, Fukushima, Chernobyl, Mihama Unit 2, Arkansas Units 1 and 2 Recent Events, SONGS 3 1

Accident and Dr. Joram Hopenfelds's concerns. 8.4 Million Southern Californians do not want a Fukushima in their backyards due to blind trust of NRC Commission in the Unsafe Gamble of Restarting of San Onofre Unit 2 for SCE Engineers to sharpen their pencils to learn from their continued mistakes. That is not how the concept ofNE!, INPO and NRC well managed, safe, clean and reliable nuclear power works. Please feel free to send me an email, if you need further assistace or have any questions. Sincerely ....HAHN Baba 2

Remsburg, Kristy From: Vinod Arora [vinnie48in@gmail,com]

Sent: Friday, April 12, 2013 2:33 AM To: CHAIRMAN Resource; Borchardt, Bill; Leeds, Eric; Howell, Art; Dorman, Dan; Benney, Brian; Hall, Randy; Lantz, Ryan; R4ALLEGATION Resource

Subject:

This is America, not Iran or North Korea, where captured politicians for their selfish motives and hidden agendas can erode NRC's Regulatory Authority, Safety Mission and Public Trust.

Follow Up Flag: Follow up Flag Status: Flagged HAHN Baba April 12,2013 at 2:26 am Your comment is awaiting moderation.

Sincere Thanks to NRC Chainnan, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

San Onofre NRC/SCEIMHIIPublic Awareness Series Please excuse me for any computer or human perfonnance grammatical or spelling errors.

Good Morning

Dear Dr. Macfarlane,

Phase 1 - Chernobyl, Fukushima, Three Mile Island, Mihama, North Ana, Davis-Besse, Turkey Point, Robinson and SONGS Unit 3 Nuclear AccidentslIncidentslEvents were caused by a combination of hardware, process and human perfonnance errors. Following the example of Arkansas Unit One Event on Easter Sunday, March 31, 2013, water from a fire main broken by a falling component or a fire damages some electrical equipment that supplies off-site power to the San Onofre Unit 2 Transfonners. Unit 2, which is operating at 70% power, automatically shuts down the reactor when off-site power is lost to all reactor coolant pumps.

Concurrent with reactor trip, Turbine and feedwater pumps trip.

Phase 2 - Per NRC Infonnation Notice No. 88-31, during the Phase 1 event, the water level on the secondary side could fall below the top of the steam generator tubes for a lO-minute period at the beginning of the event.

With tubes uncovered, this condition of ZERO Water in the Unit 2 defectively designed and degraded steam generators would cause fluid elastic instability (FEI), flow-induced random vibrations, excessive hydrodynamic pressures and Mitsubishi Flowering Effect and could conceivably cause the collapse of MHI Anti-vibration structure and failure of retainer bars.

Phase 3 - The faulted steam generator over-pressurizes due to 100% load rejection and leaking/ruptured tubes, and the main steam safety valves per SONGS procedures progressively open to prevent over-pressurizing the faulted steam generator and start releasing steam to the environment. The force of the flashing steam would create high-energy jets, lift loose parts and debris present in the steam generator. These adverse effects would cut holes into already degraded tubes and create additional loading on tube support plates (TSPs) due to heavy build-up of deposits on trefoil/quadrifoil-shaped holes from SG blowdown and crack high cycle fatigue U-bend tubes not supported by an Anti-Vibration Bars (A VB). Due to lack of in-plane restraints, large U-bends supported without A VBs and with clearances of 0.05 inches start to swing violently with large amplitudes (in plane velocities> 60 feet/sec.) and cause several tubes to leak and with double-ended ruptures in the mid-span, free span and at the junction of 7th tube support plates in a matter of minutes due to tube-tube wear and thousands of undetected macroscopic circumferential cracks. These cumulative adverse conditions in all likelihood would result in additional massive cascading of RSGs tube failures (tubes would excessively rattle or vibrate, hitting other tubes with violent impacts) due to extremely low tube-to-tube clearances and no in-plane effective anti-vibration bar support protection system. This Titanic and adverse effect would involve hundreds 1

of degraded and active SG tubes along with all the inactive (plugged /unstabilized) tubes causing an undetermined amount of simultaneous tube leaks/ruptures. The iodine in the reactor coolant assumed to be dissolved from allowable operational fuel failure or from an iodine spike produced by the transient conditions during the accident could be significantly larger than that previously approved NRC Limits.

Phase 4 - The accident would transport with steam 100% of the iodine contained in the 15,000 gallons of reactor coolant to the environment exceeding the federal regulations within 10 minutes during an extremely fast-paced transient beyond the operator control and failure of defense-in-depth actions. Core Damage Probability (CDP) and Large Early Release Probability (LERP) associated with multiple double-ended rupture of steam generator tubes will significantly increase than previously approved NRC SONGS FSAR Limits. The resulting doses would be significantly higher than the dose consequences analyzed in the SONGS UFSAR for the post-trip SLB event with a concurrent iodine spike. The postulated post-trip SLB with multiple tube ruptures and concurrent iodine spike Exclusion Area Boundary, Low Population Zone, and Control Room doses would be significantly higher than the post-trip SLB Control Room limit of 5 Rem TEDE, and the Exclusion Area Boundary and Low Population Zone limit of 2.5 Rem TEDE.

Phase 5 SCE DID Actions and unreliable operator actions to detect a leak and to re-pressurize the steam generators as claimed by Edison are not practical to stop a major nuclear accident in Unit 2 in progress in the first 5-15 minutes of the transient during the 5-month trial period.

Phase 6 - Federal regulators signaled on April 10, 2013 that running California's San Onofre nuclear power plant at reduced power would not pose a significant safety risk - a key step toward a possible restart of one of the idled reactors. The preliminary ruling from Nuclear Regulatory Commission staff represents a victory for operator Southern California Edison, which is pushing for a restart by June and has argued for months that the Unit 2 reactor is safe to run at lower power. But, Sen. Barbara Boxer, D-Calif., called plans to restart the plant before an investigation is complete "dangerous and premature."

Phase 7 Very Strong Street Rumors are that NRC Brilliant Engineers are being pushed to the side by an adamant and very power politician connected with Edison to ignore public questioning of Edison Mistakes. This is America, not Iran or North Korea, where captured politicians for their selfish motives and hidden agendas can erode NRC's Regulatory Authority, Safety Mission and Public Trust.

Phase 8 - Email to Obama Campaign Please be kind enough to deliver the following message to His Excellency, The President of United States. We need to close this Public Safety NRC Regulatory Loophole for the safety of 8.4 Million Southern Californians, a majority Democratic State.

2

Joosten, Sandy From: Vinod Arora [vinnie48in@gmail.com]

Sent: Saturday, April 13, 20137:20 PM To: CHAIRMAN Resource; Leeds, Eric; Borchardt, Bill; Lantz, Ryan; R4ALLEGATION Resource; arnie@fairewinds.com; Benney, Brian; Hall, Randy

Subject:

Reanalysis of NUREG-1919 requested for SONGS Unit 2 by NRR/RES HAHN Baba April 13. 2013 at 7:10 pm Your comment is awaiting moderation.

Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

San Onofre NRC/SCE/MHI/Public Awareness Series - Please excuse me for any computer or human performance grammatical or spelling errors.

SONGS Unit 3 RSGs massive and unprecedented tube failures and A VB/TSP degradation occurred due to fluid elastic instability, flow-induced random vibrations, Mitsubishi Flowering Effect and high cyclic fatigue under the following unique circumstances:

(1) U-tube bundle areas with high dry steam, double in-plane velocities (> 56 feet/sec, Dr. Pettigrew and others, 2006-2011) compared with out-of plane velocities assumed (28 feet/sec) to have been used in William Krotiuk 2002 Report TH calculations and predicted by Westinghouse !NRC IMHI IAREVA ATHaS Computer Models, (2) Lack of positive in-plane restraints and zero damping, (3) Large number of SONGS Unit 2 RSG U-bends with tube clearances of only 0.05 inches (Design 0.25 inches, Industry Norm> 0.25 inches)

(4) Excessive number of tubes with narrow tube pitch to tube diameter, (5) Low in-plane frequency tubes and retainer bars compared with MHI SGs higher in-plane frequency tubes and retainer bars (4) SONGS tubes being much longer than Westinghouse Model 51 steam generators (Average length of heated tube 730 inches) and other MHI SGs, (5) MHI RSGs unique floating tube bundle with degraded Retainer Bars, can collapse due to 100% tube uncovery for 10 minutes under MSLB SG Depressurization, Multiple SGTR SG over-pressurization and lifting ofSG Relief Valves, Combination of MSLB and SGTR Conditions, Release of 100% RCS Iodine to Environment (6) Large amount of uncertainties and unverified assumptions in MHI, AREVA, Westinghouse and Intertek's contact force (Zero for in-plane vibrations), wear rate and tube stress calculations (4.6 ksi versus 16-17 ksi) and computer modeling, and (7) Incomplete tube inspections in SONGS Unit 2. Incubating macroscopic circumferential cracks caused by fluid elastic instability, flow-induced random vibrations and high cycle thermal fatigue are extremely difficult to detect and be accurately sized by nondestructive evaluation techniques like X-ray, ultrasonic, and eddy current based bobbin coil probes, mechanically rotating pancake coil (RPC), etc., which have been used in 17,000 SONGS Tube inspections. State-of-the-art systems: Zetec MIZ-80 iD system, Tecnatom TEDDY+, Circular 'IE and TM, transmit-receive eddy current array probe C-3 and other specialized radiographic probes capable of detecting sub-surface cracks caused by high cycle thermal fatigue have not been used in the 170,000 SONGS Tube Partial and Limited Inspections as shown below for Unit 2 due to access problems in the most problematic inner-most sections of the U-Tube Bundle, the high-cost, lack of availability of highly specialized tools and contractors, radiation doses, and time considerations in a rush to start Unit 2. The inspection scope defectively designed and degraded SONGS Unit 2 RSGs should have covered 100% hot leg and cold leg tube inspections, 100% of dents or dings, 100% of tube inspections in the tight radius U-bends, 100% area of the Top of the Tube Sheet and Tube Support Plates. The combined effects of tube-to-tube wear and high cycle thermal fatigue 1

cracks have been witnessed by sudden tube ruptures in North Ana in 1987, MHI SG in Mihama, Japan in 1991, three tube leakages in French SGs between 2004 through 2006, 20 tube ruptures/leakages in SGs between 1980 2000 in USA, and SONGS 3 in 2012.

In light of SONGS Unit 3 massive tube failures and safety concerns of 8.4 Southern Californians, NRC staff needs to reanalyze NUREG-19l9, published 2009, "Resolution of Generic Safety Issue 188: Steam Generator Tube Leaks or Ruptures Concurrent with Contairunent Bypass from Main Steam Line or Feedwater Line Breaches" before approval of SONGS New License Amendment. This is a one-time Public Safety and Knowledge Test for Brilliant NRC Staff and applies only to SONGS Unit 2 Defectively Designed and Degraded Unit 2 Steam Generators. Excellent PR move recommended for NRC Staff to restore its degraded public image and allegations of a SCE Captured Agency. Dr. Macfarlane, Senator Barbara Boxer and His Excellency, President of the United States would like it.... Thanks 2