ML13074A549

LTR-13-0222 - E-mail Ace Hoffman Provides Media Alert - San Onofre Retainer Bar Problems

ML13074A549
Person / Time
Site:	San Onofre
Issue date:	03/13/2013
From:	Hoffman A The DAB Safety Team
To:	Macfarlane A NRC/Chairman
References
LTR-13-0222
Download: ML13074A549 (8)

Text

OFFICE OF THE SECRETARY CORRESPONDENCE CONTROL TICKET Date Printed: Mar 14, 2013 16:29 PAPER NUMBER: LTR- 13-0222 LOGGING DATE: 03/14/2013 ACTION OFFICE:

0o AI~

AUTHOR: Ace Hoffman AFFILIATION:

ADDRESSEE: Chairman Resource

SUBJECT:

Provides Media Alert - San Onofre Retainer Bar Problems ACTION: Appropriate DISTRIBUTION: RF, SECY has Ack.

LETTER DATE: 03/13/2013 ACKNOWLEDGED No SPECIAL HANDLING: Lead office to publicly release 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after SECY's assignment, via SECY/EDO/DPC.

NOTES:

FILE LOCATION: ADAMS DATE DUE: DATE SIGNED:

Joosten, Sandy From: Capt.D [captddd@gmail.com]

Sent: Wednesday, March 13, 2013 12:56 PM To: Capt D

Subject:

Media Alert: San Onofre Retainer Bar Problems ==> This is an unanalyzed event, which requires a full safety justification.

Media Alert: San Onofre Retainer Bar Problems Media

Contact:

Don Leichtling (619) 296-9928 or Ace Hoffman (760) 720-7261 The DAB Safety Team has transmitted the above AllegationNiolation to the Offices of Chairman of the NRC, California Attorney General, Senator Barbara Boxer's Committee on Environment and Public Works (EPW), California Public Utilities Commission, Congressman Markey and the NRC San Onofre Special Review Panel.

Allegation/Violations The NRC has dispositioned AIT follow-up report dated 11/09/2012, "Item 3. "(Closed) Unresolved Item 05000362/2012007-03,

'Evaluation of Retainer Bars Vibration during the Original Design of the Replacement Steam Generators" as a non-cited violation in accordance with Section 2.3.2 of the NRC's Enforcement Policy." However, as shown below, SCE/MHI' s failure to verify the adequacy of the retainer bar design as required by SCE/MHI's procedures have resulted in plugging of several hundred tubes in the SONGS brand new replacement generators. This has resulted in these violations:

1. Failure to meet NRC Chairman Standards on Nuclear Safety by SCE,

2. Failure to meet Senator Boxer's Committee on Environment and Public Works (EPW) Standards on Nuclear Safety by SCE,

3. Failure to enforce SCE Edison Contract Document instructions to MHI by SCE,

4. Failure to meet SONGS Technical Specifications by SCE,

5. Failure to meet general design criteria (GDC) in Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities GDC 14, "Reactor Coolant Pressure Boundary" by SCE/MHI,

6. Failure to demonstrate that Unit 2 retainer bars will maintain tube bundle geometry at 70% power due to fluid elastic instability during a main line steam break (MSLB) design basis event, and

7. SCE/MHI took shortcuts by avoiding the 10 CFR 50.90 License Amendment Process under the false pretense of "like for a like" replacement steam generator. SCE added 377 more tubes, increased the average length of heated tubes and changed the thermal-hydraulic operation of the RSGs without proper safety analysis and inadequate 10CFR 50.59 Evaluation.

This intentional action to produce more thermal megawatts out of the RSGs compromised safety at SONGS Unit 2 due to the failure of 90 percent through wall thickness of a tube by the inadequate design of the retainer bar.

Recommended Actions:

NRC San Onofre Special Panel is requested to resolve the above listed Allegations/Violations within 30 days of receipt of this email and prior to granting SCE's permission to restart Unit 2.

Governing Standards, Expectations and Criteria

1. NRC Chairman, Dr. Macfarlane has publicly stated, "SCE is responsible for the work of MHI and its sub-contractors."

2. Excerpt of a Edison Contract Document received from a SONGS Anonymous Insider states, "The Supplier shall prepare and 1

submit for Edison's approval a ... [Redacted]... demonstrating compliance of the RSG design with all SONGS ... [Redacted]....

The report shall include an engineering evaluation, including all necessary analyses and evaluations, justifying that the RSGs can be replaced under the provisions of 10 CFR 50.59 (without prior NRC ... [Redacted]...). The report format shall follow the guidelines of ... [Redacted]... in order to facilitate preparation of the 10 CFR 50.59 evaluation. The 10 CFR 50.59 evaluation shall be performed by Edison. Specifically, the ... [Redacted]... shall include, as a minimum, the following: Description of the RSG impact on the existing systems, structures, and components. Detailed calculations addressing the RSG impact on the UFSAR analyses, ....[Redacted]... The calculations shall include a Summary of Transients Analysis that shall evaluate the RSG impact on each event. When required by the Summary of Transients, event specific calculations shall be performed. All evaluations, analyses, and calculations shall be consistent with the latest SONGS analyses of record, evaluation methodologies, analysis processes and computer codes existing at the time of performance. The Supplier shall achieve acceptable results by minimizing any reduction in operating margins (e.g., increasing Reactor Over Power Margin [ROPM] requirements). If the Supplier determines that reduction of operating margins is necessary, it will inform Edison as soon as practical, so that the impact can be mutually agreed to, such that there is no, or minimal, impact on plant operation. All evaluations, analyses and calculations performed by subcontractors shall be provided to Edison for review, and all Edison comments shall be resolved in a manner acceptable to Edison prior to Supplier internal document approval. All evaluations, analyses and calculations, including computer code input and output, in their entirety, will be provided to Edison for future Edison use."

3. SONGS Technical Specification Section, 5.5.2.1, "Steam Generator (SG) Program", subsection c, "Provisions for SG tube repair criteria", Item 1 states, "Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 35% of the nominal tube wall thickness shall be plugged."

4. Title 10 of the Code of FederalRegulations (10 CFR), "Energy," establishes the fundamental regulatory requirements for the integrity of the SG tubes. Specifically, the general design criteria (GDC) in Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," state that the RCPB-shall have "an extremely low probability of abnormal leakage... and gross rupture" (GDC 14, "Reactor Coolant Pressure Boundary")

"shall be designed with sufficient margin" (GDCs 15, "Reactor Coolant System Design," and 31, "Fracture Prevention of Reactor Coolant Pressure Boundary")

shall be of "the highest quality standards practical" (GDC 30, "Quality of Reactor Coolant Pressure Boundary")

shall be designed to permit "periodic inspection and testing.. .to assess.. structural and leak-tight integrity" (GDC 32, "Inspection of Reactor Coolant Pressure Boundary")

Background

NRC AIT follow-up report dated 11/09/2012, "Augmented Inspection Team Follow-up Inspection (ADAMS Accession Number ML 2012010)", page 10 states, Item 3. "(Closed) Unresolved Item 05000362/2012007-03, "Evaluation of Retainer Bars Vibration during the Original Design of the Replacement Steam Generators" "In February 2012, the licensee identified wear indications in Unit 2 replacement steam generators at the tube locations in contact with the retainer bars. Some of the indications showed excessive wear with a maximum degradation of 90 percent through wall. The team identified that the design of the replacement steam generators did not expect any potential vibration concerns in the area of the tube bundle where the retainer bars were located..."

MHI Root Cause Analysis, pages 23 through 26 state, "The design function of the retainer bar is to support the AVB assembly during manufacturing and prevent excessive AVB assembly movement during operational transients. The retainer bar must be strong enough to support the AVB assembly and fit within the physical constraints of the U-bend. The tubesheet drilling pattern is one of the first design decisions made for a new steam generator and it is at that time that each tube location along the periphery of the tube bundle is established. The tube bundle design thus determines the retainer bar's length and thickness. At SONGS, in order to accommodate the increased number of tubes, the retainer bars are relatively long and thin as compared to the retainer bars in other SGs designed by MHI, resulting in their having low natural frequencies. The engineer responsible for the retainer bar design did not recognize the need to analyze the retainer bar for flow induced vibration because no such analysis had been performed on previous MHI SG designs. The design control procedure for this design activity did not identify this issue, nor was it recognized during the design review process. During operation, the secondary flow velocity and steam quality (void fraction) created turbulent flow conditions capable of causing high amplitude vibration if the retainer bar natural frequency was low enough, which turned out to be the case. The high amplitude vibration resulted in the retainer bar contacting some tubes and causing tube wear. The design control process did not provide sufficient direction to assure that an evaluation of the need for an analysis of flow induced vibration of the retainer bar was performed and verified. Basis: The evaluation team concluded that the fundamental reason for the retainer bar FIV was the lack of clear direction in the MHI design procedures to 2

require an evaluation to determine the different analyses and the level of analysis that were required for the RSG design in light of changes in the SONGS RSG design from previous MHI steam generator designs."

In October 2012 MHI reported directly to the NRC its safety concerns about the retainer bars (Part 21 - Steam Generator Tube Wear Adjacent to Retainer Bars, October 5 2012, NRC Region 1, Defects and Non-Compliance, 10 CFR 21.21 (d)(3)(i) : "The Steam Generator tube wear adjacent to the retainer bars was identified as creating a potential safety hazard. The maximum wear depth is 90%

of the tube thickness. The cause of the tube wear has been determined to be the retainer bars' random flow-induced vibration caused by the secondary fluid exiting the tube bundle. Since the retainer bar has a low natural frequency, the bar vibrates with a large amplitude. This type tube wear could have an adverse effect on the structural integrity of the tubes, which are part of the pressure boundary. The plugging of the tubes that are adjacent to the retainer bars was performed. MHI has recommended to the purchaser

[SCE] to remove the retainer bars that would have the possibility of vibration with large amplitude or to perform the plugging and stabilizing for the associated tubes..."

John Large states, "The continuous retaining bar wraps around the tube bundle to which is fixed the outboard ends of the AV bars. The retaining bar is pulled in, wrapped around the tube bundle by the hairclip-like retainer bar, this being captured in situ by being threaded through the first two rows of tubes, and held in this position by friction between the retainer bar and the inboard top surfaces of the AV bars. Because the tube-to-tube clearance tightens towards the apex of the U-bend, one-half of the total restraint assemblies require a smaller diameter retainer bar in order to fit between the tube rows."

Five tubes in Unit 2 and 3 steam generators experienced through wall wear due to retainer bar random vibration because of inadequate design. These five tubes, which exceeded > 35% wall thickness, were plugged in accordance with SONGS Technical Specifications. One tube, which experienced maximum degradation of 90 percent through wall, could have leaked, if Unit 2 was not shutdown for refueling. SCE/MHI's failure to verify the adequacy of the retainer bar design as required by Procedure S0123-XXIV-37.8.26 resulted in the plugging of several hundred tubes in these brand new replacement generators.

Deviations or Defects

1. NRC AIT follow-up report dated 11/09/2012, "Augmented Inspection Team Follow-up Inspection (ADAMS Accession Number ML 2012010)", page 14 states, "The licensee did not meet Procedure S0123-XXIV-37.8.26 requirements to ensure the design of the retainer bar was adequate with respect to the certified design specification. Specifically, the licensee failed to ensure that there was sufficient analytical effort in the design methodology of the anti-vibration bar assembly to support the conclusion that tube wear would not occur as a result of contact with the retainer bars due to flow-induced vibration. The inspectors determined that the requirements for flow-induced vibration in the certified design specification, along with the expectations in Procedure S0123-XXIV-37.8.26, provided sufficient information to reasonably foresee the inadequate design of the retainer bars during the review and approval of design Calculations S023-617-1-C749 and S023-617 C157, including the associated design 3

drawings provided by Mitsubishi. The associated violation for this performance deficiency is described in Section 40A7 of this report. [SCE identified]

2. MHI Root Cause Analysis, UES-20120254, Rev.0, page 7 states, "The design control process did not provide sufficient direction to assure that an evaluation of the need for an analysis of flow induced vibration of the retainer bar was performed and verified." [MHI Identified]

3. Failure to meet general design criteria (GDC) in Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities GDC 14, "Reactor Coolant Pressure Boundary" by SCE/MHI. [DAB Identified]

4. Failure to demonstrate that Unit 2 retainer bars will maintain tube bundle geometry at 70% power due to fluid elastic instability during a main line steam break (MSLB) design basis event. [DAB Identified]

5. SCE/MHI took shortcuts by avoiding the 10 CFR 50.90 License Amendment Process under the false pretense of "like for a like" replacement steam generator. SCE added 377 more tubes, increased the average length of heated tubes and changed the thermal-hydraulic operation of the RSGs without proper safety analysis and inadequate 10CFR 50.59 Evaluation. This intentional action to produce more thermal megawatts out of the RSGs compromised safety at SONGS Unit 2 due to the failure of 90 percent through wall thickness of a tube by the inadequate design of the retainer bar. [DAB Identified]

NRC Resolution The above referenced report pages 26 & 27 states, "40A7 Licensee Identified Violations: The following violation of very low safety significance (Green) was identified by the licensee which met the criteria of the NRC Enforcement Policy for being dispositioned as a Non-Cited Violation.

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures shall be established to provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculation methods, or by the performance of a suitable testing program.

Contrary to the above, on December 11, 2007, and July 3, 2008, the licensee failed to establish design control measures during the review of Mitsubishi's design Calculations S023-617-1-C749 and S023-617-1-Cl 57, respectively, to verify or check the adequacy of the retainer bars' design with respect to the susceptibility of the smaller diameter retainer bars to flow-induced vibration. Specifically, the licensee failed to ensure that there was sufficient analytical effort in the design methodology of the anti-vibration bar assembly to support the conclusion that tube wear would not occur as a result of contact with the retainer bars due to flow-induced vibration. Consequently, the smaller diameter retainer bars vibrated during normal operation causing wear on the adjacent tubes, which challenged the integrity of the reactor primary coolant system boundary. The inspectors determined that the licensee's failure to verify the adequacy of the retainer bar design as required by Procedure S0123-XXIV-37.8.26 was of very low safety significance (Green) based on Inspection Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," and Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At-Power," because the finding did not involve a degraded steam generator tube condition where one tube could not sustain 3 times the differential pressure across a tube during normal full power, steady state operation and. none of the replacement steam generators violated the "accident leakage" performance criterion in plant Technical Specifications as a result of the retainer bar vibration. The licensee also implemented actions to inspect all affected tubes in Unit 2 and 3 and remove from service all those tubes surrounding the smaller retainer bars that could wear due to vibration of the retainer bar. Because this violation has been determined to be of very low safety significance (Green) and has been entered in the licensee's corrective action program as Nuclear Notification NN 201843216, it will be dispositioned as a non-cited violation in accordance with Section 2.3.2 of the NRC's Enforcement Policy."

Unit 2 Main Steam Line Break The DAB Safety Team Expert Panel has concluded that SONGS Unit 2 Replacement Steam Generators (RSG) are in worse shape now than when certified by SCE and their three NEI Qualified, "U.S. Nuclear Plant Designers." The accident scenario of concern consists of two events: (1) a non-isolable secondary system break or rupture that is outside containment; and (2) a coupling of this break with the rupture of, or significantly increased leakage from, affected SG tubes. Even at 70% power operations, if a steam line break outside containment were to occur in Unit 2, the depressurization of the steam generators with the failure of a main steam isolation valve to close would result in 100% void fraction in the degraded U-Tube bundle and the straightleg portion between the Tube Support Plates. This condition of ZERO Water in the steam generators would cause fluid elastic instability (FEI) and flow-induced random vibrations (FIV), which would then result in massive cascading SG tube failures, 4

involving hundreds of degraded active SG tubes. Fluid elastic instability (FEI) and flow-induced random vibrations (FIV) can progress through a buffer zone of plugged and staked tubes to reach pressurized, in-service tubes and create additional SG tube failures. The resulting SG secondary side blow-down could further increase tube leakage due to resonance vibrations within the affected SG tube bundle. The retainer bar restraint will not be able to contain the tube bundle geometry during a main line steam break (MSLB) design basis event and collapse of the Anti-vibration bar structure would resulting in additional SG tube failures. With an undetermined amount of simultaneous tube leaks/ruptures, approximately 60 tons of very hot high-pressure radioactive primary reactor coolant would leak into the secondary system. The release of this amount of radioactive primary coolant, along with an additional approximately 150 tons of steam into the environment in the first five minutes from a broken steam line would EXCEED the SONGS NRC approved safety margins. So, in essence, the operation of these RSG's will become a vehicle for a potential nuclear accident waiting to happen. Any tube leaks/failures under the above described conditions, would allow significant amounts of radiation to escape to the atmosphere and a major nuclear accident would easily result causing much wider radiological consequences leading to even a potential nuclear meltdown of the reactor! Since these events would happen at an extremely fast pace, no credit is assumed in the first 5 -15 minutes of the main steam line break accident for: (1) Enhanced Unit 2 Defense-In-Depth Actions - SCE Restart Plan Enclosure 2, Item 9.0, and (2) The differential pressure across the SG tubes necessary to cause a rupture will not occur if operators prevent RCS re-pressurization in accordance with their Emergency Operating Procedures - Enhanced Unit 2 Defense-In-Depth Actions - SCE Restart Plan , Item 5.2,2, Probabilistic Risk Analysis.

The Operational Assessments reports prepared by AREVA, MHI and Westinghouse "conflictand contradict'each other on the causes and extent of degradation pertaining to the SONGS Unit 2 Replacement Steam Generators. It is the DAB Safety Team Expert Panel, former NRC Staff and SONGS Concerned Insiders opinion is that these reports are full of holes based on incomplete inspection data, under-conservative computer modeling and are "Smoke & Mirrors", because: (1) SCE Engineers have either not provided, or they are withholding all the information to these parties because of "The consequences of being Wrong, Terminated or Fired", (2) Due to competing and proprietary interests between the three NEI qualified, "US Nuclear Plant Designers", (3) Time Pressure exerted by SCE in their rush to Restart Unit 2, and (4) Since nobody knows what really happened, all the Parties have a shared interest to Operate Unit 2 at reduced power as a "Test Lab to conduct Unapproved Experiments" to determine, "What really went wrong with SCE's unit 3 design, so SCE can determine the Root Cause, corrective actions, repair and test plans required to return both units 2 and 3 to full power operations all at their ratepayers expense." Now SCE is trying to restart the damaged Unit 2 bypassing the normal requirement to apply for a license amendment, which would entail a higher degree of scrutiny by the NRC and the opportunity for the public to request an evidentiary hearing.

According to conversations with SONGS Insiders as early as August 2012, this whole restart exercise is an ill-conceived and pre-planned move by SCE/MHI and its Political/Financial Mentors to keep the unsafe Unit 2 operating at reduced power, so SCE can stay in the California Public Utilities Commission rate base as a base-load and grid voltage stabilizer station. The ultimate plan is to replace/repair both Units steam generators in 5 years with unsafe Unit 2 in operation, until Unit 3 steam generators can be rebuilt and be put back in operation. Once Unit 3 is in operation, unsafe Unit 2 can be shutdown, repaired and put back in service. There are new anti-vibration bar systems currently being designed and tested by MHI in Japan to rebuild these steam generators. The effort is described in the following excerpt from NRC Inspection Report, which is posted on NRC website www.nrc.gov, "Report of NRC Vendor Inspection at Mitsubishi Heavy Industries, Ltd", 11/30/2012, http://pbadupws.nrc.cov/docs/ML1233/ML12333A144.pdf "The U.S. Nuclear Regulatory Commission (NRC) inspection team observed various activities associated with the mock-up tests of a portion of the upper tube bundle. The activities being done by Mitsubishi Heavy Industries, Ltd. (MHI) were conducted to determine if a design modification to repair the San Onofre Nuclear Generating Station (SONGS) steam generators was feasible. The design modification testing consisted of anti-vibration bar insertion tests using three different designs. The three different anti-vibration bar designs were:

• Thicker - inserted between and parallel to existing anti-vibration bars

• 30 Degree - inserted at a 30 degree angle to existing anti-vibration bars, forming intersections with existing anti-vibration bars

- Comb - shaped like a comb and will be inserted into the bundle on every other row and then rotated 90 degrees, locking tubes into place between the "teeth" of the comb.

In discussions with MHI personnel, they indicated that the thicker anti-vibration bar will likely be the least difficult to insert and the comb anti-vibration bar the most difficult to insert due to slight differences in the gaps and arrangement of the tubes. MHI conducted the testing at their steam generator manufacturing facility in Kobe, Japan."

Press reports dated March 7, 2013, state, "Shuttered since early last year, the San Onofre Nuclear Generating Station (SONGS) is steadily working toward safely restarting - although you wouldn't know it from the shrill political complaints lobbed its way. On Feb. 6, Sen. Barbara Boxer (D-CA), chair of the Senate Committee on Environment and Public Works, and Rep. Edward Markey 5

(D-MA), a candidate for his state's open Senate seat, wrote to the NRC, urging further scrutiny. The normally unflappable Ted Craver, Edison's CEO, took special offense at the legislators' spurious claims. "We bristle so," Craver declared during a company earnings call, "when elected officials issue press releases suggesting that SCE was aware of design problems with the replacement steam generators when they were installed at San Onofre. This is just not accurate. And it injects politics into a process that should be free from it." Meanwhile, closer to home, Assemblywoman Toni Atkins (D-San Diego) recently sent a letter of her own to the NRC. "More than 8 million Southern Californians," Atkins wrote, "many of whom are my constituents, live within a 50-mile radius of SONGS and they are entitled to know whether they are safe." [Source: UT- San Diego, March 7, 2012]

Consequences and Impact Claims Sen. Boxer: San Onofre Report Shows Shortcuts: Federal regulators on Friday released parts of a once-confidential report at the center of a dispute between California Sen. Barbara Boxer and the company that runs the troubled San Onofre nuclear power plant. However, sections of the 64-page report released by the Nuclear Regulatory Commission were redacted, and it wasn't immediately clear if the issues highlighted by Boxer were included. The Democratic Senator said last month that the study suggests operator Southern California Edison took shortcuts that compromised safety at the seaside plant, which was shut down more than a year ago after a tube break released a trace of radiation. Edison has said the Senator is off the mark.

[Source: LA times, March 3, 2013]

Southern California Edison Comments on MHI Evaluation of San Onofre Nuclear Plant Steam Generators, Posted March 8, 2013 - 2:00 p.m. PDT, Media

Contact:

Media Relations, (626) 302-2255 ROSEMEAD, Calif., March 8, 2013 - An evaluation by Mitsubishi Heavy Industries (MHI) made public today cites ineffective tube supports, dry steam and high steam flow velocity as causes of excessive wear in the steam generators MHI supplied to Southern California Edison's (SCE) San Onofre Nuclear Generating Station. SCE previously disclosed these same causes based on its own investigation, and the Nuclear Regulatory Commission's (NRC) augmented inspection team report last July found that MHI's use of faulty computer modeling in the design process caused MHI engineers to inadequately predict the dryness of the steam, measured by void fraction, in the replacement steam generators. MHI repeatedly reassured SCE of the efficacy of the design. During the design phase of the project, MHI advised SCE that, based on its own review and analysis, the maximum void fraction that MHI expected to occur was acceptable, did not require additional design changes or measures, and that the replacement steam generators would perform as warranted. "SCE's own oversight of MHI's design review complied with industry standards and best practices," said Pete Dietrich, SCE senior vice president and chief nuclear officer. "SCE would never, and did not, install steam generators that it believed would impact public safety or impair reliability." In fact, MHI states in its root cause report (page 41), that its analysis of conditions in the steam generator during the design phase (which calculated void fraction and steam flow velocity) concluded that the thermal hydraulic conditions in the San Onofre steam generators were acceptable, and specifically that there was no need to reduce void fraction.

Additionally, SCE never rejected a proposed design change to address void fraction based on its impact on compliance with 10 CFR 50.59. "At no time was SCE informed that the maximum void fraction or flow velocities estimated by MHI could contribute to the failure of steam generator tubes," said Dietrich. "At the time, the design was considered sound." SCE is disappointed that MHI decided on its own to redact some information in its evaluation about the flaws in the computer codes. However, the NRC publicly disclosed the computer code flaws three months before MHI completed its evaluation. In addition, the corrective actions and other statements included in the evaluation make it evident that there were problems with the computer modeling that failed to predict conditions that led to the tube-to-tube wear. http:llwww.sonqscommunity.com/newsroom.asp#030813b Experts react to the above comments:

NRC has dispositioned as a non-cited violation in accordance with Section 2.3.2 of the NRC's Enforcement Policy." However, SCE/MHI' s failure to verify the adequacy of the retainer bar design as required by SCE/MHI's procedure resulted in plugging of several hundred tubes in brand new replacement generators. John Large states, "However, MHI's advice to either plug the local tubes and/or remove the retainer bars at risk raises two issues unique to the retainer bar and its sub-assembly: (i) Plugging of the at-risk tubes is not a satisfactory solution because it is the retainer bar that vibrates via random fluid flow processes at sub FEI critical velocity levels - these are likely to continue in play or, indeed, exacerbate at the proposed U2 restart at 70% power, leading to through-tube abrasion, the detachment of tube fragments, lodging at other unplugged and in-service tube localities, resulting in the so-called 'foreign object' tube wear; ii) MHI's recommendation that those retainer bars at risk of large-amplitude fluid flow excited vibration should be removed is, of course, dependent upon reliable analysis to identify the at-risk assemblies; and, importantly, and (iii) this restraint system probably also serves to contain the tube bundle geometry during a main line steam break (MSLB) design basis event, so any change or removal of the retaining bar assemblage would require a full safety justification."

NRC Offices of Nuclear Reactor Regulation (NRR) repeats the issues raised by John Large in RAI 15 and SCE's response to NRR RAI 15 is completely unsatisfactory, because during a main line steam break, the whole tube bundle will experience fluid elastic instability (FEI). Based on a MHI Root Cause Evaluation, Dr. Pettigrew's research paper in 2006 (ratio of in-plane critical velocity and out-of plane critical velocity, Air Water Mixture = 2.7) and another research paper published in 2011 (ratio of in-6

4 It I plane critical velocity at start of U-bend and out-of plane critical velocity, Steam Water Mixture = 2.0), it is reasonable to assume that due to FEI, the fluid velocities in the in-plane direction will be double of out-of plane fluid velocities as opposed to 1.5 assumed by Mitsubishi. Therefore, the retainer bar restraint will not be able to contain the tube bundle geometry due to the exponential fluid velocities in the in-plane direction and/or the increased stresses during a main line steam break (MSLB) design basis event and can potentially cause collapse of the Anti-vibration bar structure resulting in additional SG tube failures. This is an unanalyzed event, which requires a full safetyjustification.

The DAB Safety Team: Don, Ace and a BATTERY of safety-conscious San Onofre insiders plus industry experts from around the world who wish to remain anonymous. These volunteers assist the DAB Safety Team by sharing knowledge, opinions and insight but are in no way responsible for the contents of the DAB Safety Team's reports. We continue to work together as a Safety Team to prepare additional San Onofre Papers, which explain in detail why a SONGS restart is unsafe at any power level. For more information from The DAB Safety Team, please visit the link above.

Our Mission: To prevent a Trillion Dollar Eco-Disaster, like Fukushima, from happening inthe USA.

Copyright March 13, 2013 by The DAB Safety Team. All rights reserved. This material may not be published, broadcast or redistributed without crediting the DAB Safety Team. The contents cannot be altered without the Written Permission of the DAB Safety Team Leader and/or the DAB Safety Team's Attorney.

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