ML13169A379

LTR-13-0535 Bill Hawkins, Multiple E-mails San Onofre Nuclear Generating Station

ML13169A379
Person / Time
Site:	San Onofre
Issue date:	06/03/2013
From:	Hawkins B Public Commenter
To:	Macfarlane A, Magwood W, Apostolakis G, Brian Benney, Borchardt R, Dan Dorman, Howell A, Ryan Lantz, Leeds E, Ostendorff W NRC/Chairman, NRC/OCM, NRC/EDO, Office of Nuclear Reactor Regulation, Division of Operating Reactor Licensing, NRC/OCM/GEA, NRC/OCM/WCO, NRC/OCM/WDM, NRC Region 4
References
LTR-13-0535
Download: ML13169A379 (13)

Text

Joosten, Sandy From: Bill Hawkins <billlee123456@gmail.com>

Sent: Monday, June 03, 2013 9:17PM To: CHAIRMAN Resource; CMRAPOSTOLAKIS Resource; CMRMAGWOOD Resource; CMROSTENDORFF Resource; Hall, Randy; Leeds, Eric

Subject:

San Onofre void fractions, 50.59, Senator Boxer, Nunn, SCE & NRC Edison throws stones at Honorable Senator Barbara Boxer and with great confidence and arrogance quotes approval of San Onofre SGRP 10 CFR 50.59 via NRC Safety Evaluations and NRC AIT Report. Edison seeks to minimize damage from Dwight Nunn's Letters and rejection of recommendation to reduce high void fractions by SCE/AVB Team. The high void fractions of99.6% to maximize thermal output from RSGs caused a tube leak and unprecedented tube-to-tube wear and high cycle thermal fatigue cracks in Unit 3. The high void fractions of98.9% to maximize thermal output from RSGs caused high cycle thermal fatigue cracks in thousands of Unit 2 tubes. The behavior by Edison is against all the nuclear safety, public transparency, corporate business, its overriding obligations to its shareholders and public safety, democratic NORMS and is a blatant violation of Federal Regulations. Ratepayers need nuclear energy with safety, respect and reliabilty and not under the threat of a Fukushima. NRC charter is to respect and protect public safety and not join hands with SCE and throw stones at public.

Edison states, "Replacement of the steam generators is a replacement in-kind in terms of an overall fit, form, and function with no, or minimal, permanent modifications to the plant systems, structures, and components (SSCs)* The fact that the RSG were designed, fabricated and examined to the newer edition of the ASME Code is an enhancement over the OSGs."

SCE website states, "SCE advised the NRC that the San Onofre steam generators contained a number of different features from the precious design. In fact, safety evaluations prepared by the NRC in connection with amendments to the San Onofre license associated with the steam generator replacements described the most important of those changes in detail. At no time did SCE hide the differences from the NRC, nor did it seek to mislead the NRC concerning the applicability of Section 50.59 to the project. Any suggestion that seeks to draw from the November 2004 letter a contrary conclusion is simply incorrect and relies on the fundamental error of viewing Section 50.59 as applying to identical, or "like for like" replacements."

A like-for-like replacement is defined as the replacement of an item with an item that is identical. For example, the replacement item would be identical if it was purchased at the same time from the same vendor as the item it is replacing, or if the user can verify that there have been no changes in the design, materials, or manufacturing process since procurement of the item being replaced. If differences from the original item are identified in the replacement item, then the item is not identical, but similar to the item being replaced, and an evaluation (Such as 10CFR 50.59, Changes, tests or experiments .. ) is necessary to determine if any changes in design, material, or the manufacturing process could impact the functional characteristics (high void fractions, high steam flows, high fluid velocities, excessive tube vibrations, etc.) and ultimately the component's ability to perform its required safety function (SG Tube structural Integrity). If the licensee cannot demonstrate that the replacement item is identical and differs in design, or results in a design change, new test or experiment, which adversely affects RSG's functional characteristics (high Void fractions, high steam flows, Wrong AVBs for FEI) and ultimately the RSGs ability to perform its required safety function (RCS Barrier), then the licensee needs to inform NRC ASAP and proceed safely for a NRC Approved 50.90 License Amendment like Palo Verde.

San Onofre's Partner, Bigger and Older Brother, Palo Verde RSG NRC Approved 10 CFR 50.90 License Amendment- Generally, the RSGs differ from the Original Steam Generators (OSGs) as follows:

1

• The number of tubes is increased by 10%.

• Primary and secondary water volumes are increased.

• The RSG dry weight is increased.

• The RSGs are taller, resulting in an increase to the main steam nozzle elevation.

• The upper level indication nozzle tap elevations are higher.

• The main feedwater (FW) nozzle elevation is higher.

• The snubber lugs are at the same elevation but now project from the shell cone.

• A new recirculation nozzle is added.

• A new upper blowdown nozzle is added.

• SG tube material is changed from Inconel 600 to Inconel 690.

• New computer programs used for RSG stress analyses are discussed in Section 5.5.2 of this report. (SCE Got Violation for using unapproved computer programs)

The RSGs are being designed and analyzed in accordance with the ASME Code for structural acceptability, thermal-hydraulic, U-bend fatigue, tube degradation, tube plugging, and repair requirements. The maximum allowable RSG tube wall degradation has been analyzed according to the requirements of Regulatory Guide 1.121 Flow-induced Vibrations & Fluid Elastic Instability The World's Foremost Renowned Professeur Titulaire, Michel J. Pettigrew, Ecole Polytechnique de Montreal, on the subject of fluid elastic instability and turbulence-induced vibration states in the 1970's, "It is concluded that, although there are still areas of uncertainty, most flow-induced vibration problems 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. There has been no case yet where vibration considerations have seriously constrained the designer." Dr. Pettigrew told the NRC Commissioners again in 2013 that San Onofre Replacement Steam Generators flat anti-vibration bars do not provide a positive restraint against fluid elastic instability.

Shahab Khushnood, Zaffar M. Khan, M. Afzaal Malik, Zafar Ullah Koreshi and Mahmood Anwar Khan wrote in 2003, "Flow-induced vibration is an important concern to the designers of heat exchangers subjected to high flows of gases or liquids. Two-phase cross-flow occurs in industrial heat exchangers, such as nuclear steam generators, condensers, and boilers, etc. Under certain flow regimes and fluid velocities, the fluid forces result in tube vibration and damage due. to fretting and fatigues of tubes. Prediction of these forces requires an understanding of the flow regimes found in heat exchanger tube bundles. Excessive vibrations under normal operating conditions can lead to tube failure. Relatively little information exists on two-phase vibration. This is not surprising as single-phase flow induced vibration; a simpler topic is not yet fully understood. Vibration in two-phase is much more complex because it depends upon two-phase flow regime, i.e. characteristics of two-phase mixture and involves an important consideration, which is the void fraction. The effect of characteristics of two-phase mixture on flow-induced vibration is still largely unknown. Two-phase flow experiments are much more expensive and difficult to carry out as they usually require pressurized loops with the ability to produce 2

two-phase mixtures. Although convenient from an experimental point of view, air-water mixture ifused as a simulation fluid, is quite different from high-pressure steam-water. A reasonable compromise between experimental convenience and simulation of steam-water two-phase flow is desired."

One Masters Research Student R. Viollette and Dr. Pettigrew state in a 2006 research paper, "Fluid elastic instability is the most important vibration excitation mechanism for heat exchanger, or steam generator type of tube bundles. It is so because of the very high vibrations amplitude that it can induce to the tubes, which can lead to rapid failure by fatigue or wear. Also, unlike vibrations induced by vortex shedding (vortex-induced vibrations), fluid elastic instability is not a self-limiting phenomenon: amplitude of vibrations does continue to increase with velocity past the critical onset of the instability." Dr. Pettigrew told in his 2006 paper that design of flat anti-vibration bars design need to be verified against fluid elastic instability.

Quotes from [Redacted] Anonymous Engineers One [Redacted] Root cause Leader stated, " I wish that these [Redacted] Engineers had made these changes one by one and tested them before implementing them in the RSG design."

Another [Redacted] Project Manager said, " I wish that these [Redacted] Engineers had duplicated Palo Verde Generators, went through an independent design review .. ~ .... "

Another [Redacted] Retired Manager said, "These whole design changes by these [Redacted] Engineers and

[Redacted] Management were geared only to maximize the thermal performance and profits or the new Replacement Generators because of change from Alloy 600 tube material in the OSGs to Alloy 690 in the RSGs." Even though Alloy 690 has better corrosion resistance than Alloy 600, Alloy 690 has a 10-12 less heat transfer coefficient than Alloy 600.

10 CFR 50.59- Changes, tests, and experiments. See Notes 1, 2 & 3 below for NRC AIT evaluation of SONGS IOCFR 50.59 Changes)

(c) (2) A licensee shall obtain a license amendment pursuant to§ 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would:

(i) Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the final safety analysis report (as updated);

Utility Likely Answer N/A with no explanation 3

Comments: The [Redacted] Engineers evasively answered N/A and did not answer the following NRC 50.59 Inspection Manual safety questions required before making the change. (1) Systems and components affected by the change (What is the effect of the change on their capability to perform their specified or intended functions?); (2) Parameters of the accident analysis affected by the change (Are all the relevant design basis accidents and transients identified?); (3) Potential effects of system or component failure (i.e., the question, "what would happen if..." is explored and answered in the evaluation), and (4) How the evaluation criteria are met. The first criterion is if the CTE would result in more than a minimal increase in the frequency of an accident previously evaluated in the FSAR (as updated). The intent of the criterion is to allow changes to be made without approval unless there is a discernible, attributable increase in frequency of an accident. There must be some reason to believe that the CTE would result in a more than minimal impact upon the accident frequency (as because it affects the integrity of the reactor coolant system, or the ability of SSC to remove decay heat, or makes an initiating event more likely to occur). Licensees must still meet applicable regulatory requirements. As noted in NEI 96-07, departures from the design, fabrication, testing and performance standards in the General Design Criteria are not compatible with a "no more than minimal increase" standard.

Title 10 ofthe Code ofFederal Regulations (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- (1) Shall have "an extremely low probability of abnormal leakage ... and gross rupture" (GDC 14, "Reactor Coolant Pressure Boundary"), (2) Shall be designed with sufficient margin" (GDCs 15, "Reactor Coolant System Design," and 31, "Fracture Prevention of Reactor Coolant Pressure Boundary"), and (3) shall be of"the highest quality standards practical" (GDC 30, "Quality of Reactor Coolant Pressure Boundary")

Most of the successful steam generators operate at void fractions less than 98.5%, steam pressures> 900 psi and recirculation ratios >4. This ensures that high dry steam is not produced in localized areas ofU-tube bundle and tubes and tube supports are not subject to excessive vibrations/tube contact and tube dry-out due to fluid elastic instability, flow-induced vibrations and excessive fluid hydrodynamic pressures. The original San Onofre steam generators had circulation ratios of 3.3, void fractions of 96.1 %, steam pressures of 900 psi and did not experience fluid elastic instability in 28 years of operation. All the four San Onofre replacement units had poor circulation ratios of 3.3, void fractions ranging from 98.9-99.6%, steam pressures ranging from 833-942 psi, RCS flows between 74-7 6 Mlbs/hour, narrow tube to pitch tube diameter, excessive number of tubes (9, 727),

extremely tall tubes (average length of heated tube increased by 50 inches, equivalent to the addition of -650 tubes), 116,000 square feet oftube heat transfer area (increased from 104,000 in the OSGs) and virtually no in-plane restraints. Successful Steam Generator manufacturers control a combination of design and operational features to prevent tube/support damage from fluid elastic instability, flow-induced vibrations and excessive fluid hydrodynamic pressures. Because of the customer's (SCE) desire to increase thermal output from steam generator and profits, Steam Generator manufacturers do not design and test/experiment anti-vibration bar support systems in an operating steam generator to demonstrate its capability to prevent tube/support damage from high void fractions (fluid elastic instability), flow-induced vibrations and excessive fluid hydrodynamic pressures. This is against the 10CFR 50.59 rules and GDC 14, 15 and 30.

The changes in void fractions due to numerous unanalyzed and untested design and operational changes not only increased significantly the frequency of occurrence of a steam generator tube leak in Unit 3 but destroyed 4

the Edison's newly constructed Billion Dollar 21st Century Safest and Innovative Steam Machines in less than 2 years of operation due fluid elastic instability, flow-induced vibrations and excessive fluid hydrodynamic pressures. In addition, 8 tubes in Unit 3 failed structural and leakage integrity criteria during their in-situ

.pressure main steam line break pressure testing. The high void fractions ultimately destroyed the RCS Barrier protected by the tubes. This was one of the largest steam generators ever built and represented a significant increase in size from those that Mitsubishi Heavy Industries has built in the past. It required Mitsubishi Heavy Industries to evolve a new design beyond that which they currently have available. Such design evolutions required a careful, well thought approach that should have fully evaluated the risks inherent in creating a new and significantly larger steam generator. Such design evolutions challenged the capability of existing models and engineering tools used for proven steam generator designs. Success in developing a new and larger steam generator design required a full understanding of the risks inherent in this process and putting in place measures to manage these risks. Void fraction was an important thermal- hydraulic parameter, related to the probability of tube dry out occurring during power operation (the higher the void fraction, the higher the probability of dry out). Tube dry out is an undesirable phenomenon as it may eventually result in tube cracking. AVB team recognized that the design for SONGS RSGs resulted in higher steam quality (void fraction) 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 design consequences and the SCE/MHI 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 CFR 50.59." In Unit 2, the RSGs were installed and tested in 2009110 and in Unit 3 in 2010/11. The RSG post-installation test results met acceptance criteria for all specified test parameters except measurements of void fractions to confirm the safety of new design, thus wasting all the money, time and the effort put into their fabrication.

(ii) through vii)

AnswerN/A (viii) Result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses.

Answer NO

1. RCS Structural Integrity: Changing from ANSYS to ABAQUS does not constitute a departure from a method of evaluation described in the UFSAR used in establishing the design basis or in the safety analyses.

2. Seismic Analysis ofReactor Vessel Internals: Changing from Topical Report CENPD-178 to CENPD-178-P does not constitute a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

3. Tube Wall Thinning Analyses: Changing from CEFLASH, STRUDL and ANSYS to Manual Calculations and ANSYS: RSG analysis considered LOCA loads, DBE loads, and maximum MSLB primary-to-secondary differential pressure across the tubes simultaneously, which is conservative.

Implement the activity per plant procedures without obtaining a License Amendment.

5

Likely Wrong Answer by Utility X Request and receive a License Amendment prior to implementation.

Results in more than a discernible, attributable increase in the frequency of occurrence of steam generator tube leakage/cracking/rupture an accident previously evaluated in the final safety analysis report (as updated).-

This is the Right Answer Notes:

1. NRC AIT Report States, "Based on the updated final safety analysis report description of the original steam generators, the steam generators major design changes were appropriately reviewed in accordance with the 10 CFR 50.59 requirements. However, further review is required related to the change in methodology used for the steam generator stress analysis calculations. An NRC senior reactor analyst performed a preliminary risk assessment. The risk is composed oftwo parts: (1) a non-consequential steam line break that induces a steam generator tube rupture, specifically involving the degraded tubes; and (2) an elevated risk of a tube rupture as an initiating event. Although in this case the degraded condition of the tubes was manifested as a small primary to secondary leak, it is possible that a full-blown rupture could have been the first indication. It should be noted, this is a preliminary assessment of the risk requiring additional information and inspection to ascertain whether a performance deficiency exists. This does not include or preclude regulatory or enforcement action by the NRC. The NRR technical specialist reviewed SCE's 10 CFR 50.59 evaluation and found two instances that failed to adequately address whether the change involved a departure of the method of evaluation described in the updated final safety analysis report. This issue is identified as URI 05000362/2012007-10, "Evaluation of Departure ofMethod of Evaluation for 10 CFR 50.59 Processes."

2. NRC AIT Team Response - ABAQUS Instead of ANSYS 24: The NRC follow-up AIT Report states, "The inspectors determined that the NRC had approved using ABAQUS for reactor coolant system structural integrity analyses. Therefore, the inspectors determined that the change from ANSYS to ABAQUS did not require the licensee to obtain a license amendment prior to implementing the change. However, the inspectors noted that the 10 CFR 50.59 written evaluation for this change, did not state that ABAQUS had been approved by NRC for the intended application. Therefore, the inspectors determined that the evaluation was not adequate, in that it did not provide a correct basis for the licensee's determination that the change from ANSYS to ABAQUS did not require a license amendment prior to implementing the change. Title 10 CFR 50.59(d)(l) requires that the licensee maintain records of changes in the facility that "include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment ... ". The licensee's failure to provide an appropriate basis for the licensee's determination that the change from ANSYS to ABAQUS did not require a license amendment prior to implementing the change therefore constituted a violation of 10 CFR 50.59(d)(1). Because this violation impacted the regulatory process, the inspectors assessed it in accordance with the NRC Enforcement Policy, as directed by MC 0612, Appendix B, "Issue Screening". NRC Enforcement Manual contains specific processes and guidance for implementing this Policy. NRC Enforcement Manual, Section 2.1 O.D.6 states, in part, that minor violations include the failure to meet I 0 CFR 50.59 requirements that involve a change to the final safety analysis description where there was no reasonable likelihood that the change would ever require NRC approval per 10 CFR 50.59. As described above, the change from ANSYS to ABAQUS did not require a license amendment prior to 6

implementing the change, so with respect to section 2.1 O.D.6 of the NRC Enforcement Manual, there is no reasonable likelihood that the change from ANSYS to ABAQUS would ever require NRC approval. Therefore, in accordance with the NRC Enforcement Manual, the inspectors determined that the licensee's change from ANSYS to ABAQUS was a minor violation of 10 CFR 50.59(d)(l ).

3. NRC AIT Team Response- ANSYS Instead of STRUDL and ANSYS- The inspectors' review of this issue did not identify a violation of 10 CFR 50.59. The inspectors noted that the licensee had used ANSYS to calculate tube stresses for the both the original steam generators and replacement steam generators, so their use of ANSYS for this purpose did not constitute a change from the method described in the Updated Final Safety Analysis Report. The inspectors noted that for the original steam generators, the licensee had analyzed a combination of events using STRUDL to calculate displacement histories for those events. This provided additional margin for analysis done by ANSYS. However, for the replacement steam generators, the licensee had analyzed for the most limiting event, and had sufficient margin, so STRUDL was not needed. Based on this, the inspectors determined that the licensee had changed from using ANSYS and STRUDL to analyze several events for the original steam generators, to using only ANSYS to analyze a single limiting event for the replacement steam generators. Therefore, because the licensee did not change the method described in the Updated Final Safety Analysis Report, the inspectors concluded that the licensee did not need to obtain a license amendment prior to implementing that change.

7

Joosten, Sandy From: Bill Hawkins <billlee123456@gmail.com>

Sent: Wednesday, June OS, 2013 11:15 AM To: CHAIRMAN Resource; CMRAPOSTOLAKIS Resource; CMRMAGWOOD Resource; CMROSTENDORFF Resource; Benney, Brian; Howell, Art; Leeds, Eric; Borchardt, Bill; Dorman, Dan; R4ALLEGATION Resource; Lantz, Ryan; Hall, Randy

Subject:

San Onofre Unit 2 Sad Saga Continued- SCE/MHI/NRC/Public Awareness Series Challenging and Lingering Questions for NRC Commission, NRR, NRC AIT, NRC Office of Inspector General, NRC Office of Investigations, Senator Barbara Boxer, Congressman Ed Markey, ASLB and Special San Onofre Panel

1. It is the unanimous decision of unbiased Independent Safety and SONGS Insider 50.59 Experts that SONGS 10CFR 50.59 was not performed properly by SCE. If NRC had reviewed these changes in detail through a 50.90 License Amendment like Palo Verde SGRP, then the problems with Unit 3 FEI high void fractions, untested design and unanalyzed operational changes could have been averted. With a little respect for the Federal Regulations, this Billions of Dollars embarrassing fiasco for NRC Commission, SCE, MHI and entire nuclear industry could have been prevented. SCE is setting a bad and dangerous example for NRC Commission, Nuclear Industry, NEI, INPO, EPRI and other INPO I and Fleet Utilities. SONGS is an INPO 4 Plant, with the worst management, maintenance, fire/industrial safety and worker treatment/discrimination/retaliation record in the entire US Nuclear Fleet.

2. Root Cause for Units 3 and 2 have not been determined because:

a. Uncertainties with computer modeling for in-plane fluid elastic instability, and

b. Operational differences between Units 2 & 3 have not been analyzed in detail by NRC AIT consistent with its reports and SONGS Procedures, that is why, SCE, MHI, AREVA, Westinghouse, NRC Staff & Commissioners have not arrived at a clear and unanimous conclusions on Unit 2 FE! and Tube-to-AVB contact Forces. What were the correct contact force, which prevented tube-to-tube wear in Unit 2. 2N, 10 Nor 30N? What was the effect ofTube-to-AVB Gaps and Mitsubishi Flowering Effect on Unit 2 FEI? MHI Quarter Bundle, Statistical Simulation, and Testing data results are based on hideous data and is incomplete and totally unsatisfactory.

3. SCE, MHI, AREVA, Westinghouse and NRC have not addressed the synergic effects of tube-to-tube wear and high cycle metal fatigue on thousands of worn and damage tubes.

4. SCE/MHI explanation of tube-to-AVB contact forces and FE! fatigue stress levels in Unit 2 is incomplete and totally unsatisfactory.

5. Unit 2 Tubes have not been inspected for internal incubating cracks.

6. SCE/NRC explanation of new License Amendment to operate at 70% power is incomplete and totally unsatisfactory. Unit 2 restart at reduced power is an unapproved test or experiment with the health and safety of 8.4 Million Southern Californians soley for the purpose of keeping SCE in Business.

7. SCE/MHI/NRC still do not understand that you cannot design anti-vibration bars to prevent the adverse effects of high void fractions of 99.6% as warned by Dwight Nunn, Dr. Pettigrew and other researchers. Therefore, the best advice is to operate the completely re-built replacement steam generators at void fractions of 96%, steam pressures > 950 psi, 1500 MWt and recirculation ratios >5.

8. SCE Operator Actions and the reliability/readiness of Unit 2 Diesels and Auxiliary Feedwater System to mitigate the consequences of a DBA with multiple tube ruptures are highly questionable.

9. SONGS Emergency Plan is totally inadequate during a DBA with multiple tube ruptures. NRC Reasonable Assurance cannot be provided by outdate Probability and Computer Models. Even a small accident would put SCE/MHI out of commission and destroy the a

public trust and credibility of NRC Commission. Is it really worth the risk? Think Honestly and with cool mind, please ask yourself these questions after Fukushima, Chernobyl, Three Mile Island, David-Bessee and SONGS 3. Actions speaks louder than false safety sermons.

1

10. US Justice Department needs to investigate SCE, NRC and MHI Key players under oath to determine the truth as demanded by Honorable Senator Barbara Boxer and Congressman Ed Markey.

Press Reports: "The former head of the Nuclear Regulatory Commission said Southern California Edison's plans to restart San Onofre at 70 percent power do not inspire him with confidence. A question specifically about the proposal to restart the San Onofre nuclear power plant elicited a skeptical response from Jaczko. The plant has been offline since January 2012 after a small radiation leak and the NRC is currently considering a proposal to restart one reactor at 70 percent for five months. "Just looking at it as an outsider now,"

Jaczko said, "the approach that is being taken is not one that instills tremendous confidence in me because the approach is for operation at reduced power. In principal, what you should see is design modifications and changes that allow operations at the licensed power levels." In other words: a fix. Jaczko said the restart proposal creates doubt in his mind that there's a complete understanding of what's wrong with the faulty steam generators at San Onofre.

Arnie Gundersen, said the probability of a nuclear power plant accident is in fact much higher than estimates, becuase five units have melted down in the past 35 years: one at Three Mile Island, one at Chernoble and three at Fukushima." We are dealing with a technology that can have 40 great years and one bad day," Gundersen said, "and that one bad day can destroy a country."

Peter Bradford, a commissioner with the NRC at the time of the Three Mile Island accident, said the regulatory agency pledged to become more transparent but in fact the opposite has happened. He pointed out little was done after Fukushima to improve safety at U.S. power plants, whereas there was heightened security after the Boston bombings. Bradford called for appointing nuclear regulatory commissioners who have a record of protecting public safety, as well as a record of technical experience with the industry."

2

Joosten, Sandy From: Bill Hawkins <billlee123456@gmail.com>

Sent: Saturday, June 08, 2013 1:11 PM To: CHAIRMAN Resource; CMRAPOSTOLAKIS Resource; CMRMAGWOOD Resource; CMROSTENDORFF Resource; Benney, Brian; Hall, Randy; Lantz, Ryan; R4ALLEGATION Resource; Borchardt, Bill; Howell, Art; Dorman, Dan; Leeds, Eric

Subject:

SCE Announces Shutdown of San Onofre. Afraid of investigations, Dwight Nunn, NOT NRC, ASLB, Public, CPUC and Customers To: SONGS Employees and Supplemental Workers It is with a heavy heart that I share with you SCE' s decision to permanently retire both Units 2 and

3. I recognize this difficult announcement is something none of us wanted to hear, but our decision is absolutely the right thing to do. The tough reality is that the recent Atomic Safety and Licensing Board decision creates significant additional uncertainty regarding our ability to get to an NRC decision to restart Unit 2 this year. This is not good for our customers, our investors and the regiOn.

Be proud, but never satisfied!

Pete There may be lots of questions yet to be answered about Southern California Edison's permanent shutdown of its San Onofre nuclear plant, but here are a couple about which there's no doubt.

Who's responsible? Edison, 100%. Accept no argument that it did the best it could in overseeing a

$700-million generator replacement project, but accidents happen. This wasn't an accident: It was the product of what Edison claims was its rigorous and negligent oversight of contractors. MHI was unable to build a steam generator specified by the inexperienced Edison Steam Generator Designers. On top ofthat Edison Engineers prepared defective 10CFR 50.59, subverted NRC regulatory process, ignored recommendations of SCEIMHI A VB Joint team established by Dwight Nunn, and misdirected MHI, Westinghouse, AREVA and Intertek in preparation ofUnit 2 Return to Service Reports. They used and abused any body they could find to achieve their end goal, but failed and abandoned the San Onofre Sinking Ship in Panic.

How much should Edison's customers pay for the misengineering and mismanagement that led to mothballing a hugely important generating station? That's easy. The answer is nothing. Not a dime.

SONGS Management has been misleading the public since the inception of Steam Generator Replacement Project. Their focus has always been on profits/production and preaching false sermons of their overriding obligation to safety and achieving excellence in operations. They have indulged in systematic retaliation of workers reporting nuclear safety concerns regarding steam generators, cyber security program, fire/safety, discrimination and harassment. SONGS Unit 3 1

Root cause was rejected in early June 2012 by SONGS Insiders and they were warned about MHI. SONGS Management were warned by many insiders that the Unit 3 Root Cause was a result of design deficiencies and changes as a result of 11% increase in heat transfer area of the tubes due to change of Alloy 600 from to Alloy 690 and evolutionary untested A VB design. Dwight Nunn's 2004 and 2005 letters warned about high void fractions and the capabilities ofMHI to build such massive steam generators and evolutionary AVBs capable of handling high viid fractions and tube fractions. Good SONGS SNO's like Dwight Nunn and Ross Ridenoure were kicked out by SCE and resigned abruptly without explanation.

It is not the Atomic Safety Board, NRC, MHI, Independent Safety Experts, CPUC or Public, which created significant additional uncertainty regarding SCE's decision to get to an NRC decision to restart Unit 2 this year. It is the inept, inefficient and cunning SONGS Senior Leadership Team, which was focussed on making money and bonuses for themselves, and subverting the regulatory process and not worried about plant safety, workers or the public. Justice Department, NRC Office of Inspector General and Investigations should continue their investigations into allegations of wrong doing by SONGS Senior Leadership Team. In the end, SONGS Senior Leadership Team was so afraid of these investigations, that they decided to abandon the ship by announcing Shutdown of SONGS blaming ASLB, NRC and Economics and coming with a new excuse, "This is not good for our customers, our investors and the region." Since when SCE was worried about the customers, safety and the region. These guys did not have the courtesy of informing NRC, MHI, SONGS Workers, CPUC and SDG&E, their supporters throgh this crisis before announcing the decision.

Tho~ght ~fJhe ~?YO~~~*~~n~~rs**()*f:"(JHit::2 ~estatt E~erill1entiss].i~d:::~~e* day before Sanofre Panicky,and Sudd~n Shutdown Ariilounc.ement Continuous monitoring of primary-to-secondary leaks led to three shutdowns at the Cruas NPP: unit 1 in February 2004 and unit 4 in November 2005 and February 2006. 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 of TSP 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. Based on this preliminary analysis, EDF has implemented preventive measures 2

(stabilizing and plugging of tubes in the central area of the tube bundle deemed sensitive to high cycle fatigue risk).

AREVA states, "Out-of-plane fluid-elastic instability has been observed in nuclear steam generators in the past and has led to tube bursts at normal operating conditions. Given identical designs, Unit 2 must be judged, a priori, as susceptible to the same TTW degradation mechanism as Unit 3 where 8 tubes failed structural integrity requirements after 11 months of operation. Based on the extremely comprehensive evaluation of both Units, supplemented by thermal hydraulic and FIV analysis, assuming, a priori, that TTW via in-plane fluid-elastic instability cannot develop in Unit 2 would be inappropriate. The nominal distance between extrados and intrados locations of neighboring U-bends in the same plane ranges from 0.25 inches to 0.325 inches due to the tube indexing. There are 36 U-bends in Unit 2 SG E-088 and 34 in SG E-089 with a separation less than or equal to 0.050 inches."

The circulation ratio of the replacement steam generator secondary-side fluid (ratio of riser mass flow-rate to steam outlet mass flow rate) at 70% power is- 4.9. A higher circulation ratio limits concerns regarding heat transfer performance, generator sludge management, corrosion product transfer, and tube dry-out.

Based on recent Mitsubishi Testing conducted in Japan, tube-to-AVB contact force more than 30N is required to counteract the adverse effects of in-plane fluid-elastic instability. Unit 2 AVBs only have 2N contact force, which cannot stop tube-to-tube wear and tube burst at 70% Unit 2 normal power operations, if in-plane and out-of-plane fluid-elastic instability develops due to abnormal operation occurrences, main steam line breaks, inadvertent equipment errors and other plant transients.

Let us assume, hope and pray for the benefit of 8.4 Million Southern Californians, IPC, State of California, CPUC, MHI, SCE and NRC, the probability of occurrence of these events is very low and nothing happens. But as stated above, 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 problem lies that in these U-bends, even at 70% power and a circulation ratio of 4.9, localized areas with very poor circulation ratio and no flow zones (Flow areas blocked by SG debris and corrosion products) can develop resulting in very high void fractions. With no tube damping and insufficient contact forces, in-plane fluid-elastic instability and out-of-plane vibrations can develop, as we witnessed in Unit 3. Just like Unit 3, now, the tubes will start moving in the in-plane direction and hit other worn and plugged/stabilized tubes with low clearances and cause tube-to-tube wear.. Also, the tubes in other non in-plane FEI areas will also start moving in the out-of-plane direction, hitting already damaged AVBs with sharp corners (Zero Radius) resulting in the existing incubating cracks in the tubes to grow at a undefined rate.

Now the tubes are wearing and cracks are growing without the knowledge of the operator, because there is no instrumentation installed in the SGs as a part of the NRC 3

Confirmatory Letter to warn/alarm the operator, as to what is going on about this kind of event. This event can occur at any time and propagate during the Unit 2, 5-month experiment window. Now, one, two or more than 5 tubes can potentially leak and/or rupture and the operator gets sudden warning/alarms through existing radiation monitors and proposed temporary N-16 detectors located on the main steam lines.

Shift Manager has only 15 minutes to diagnose, trouble shoot, declare the event and notify the Offsite Agencies for activation of the SONGS Emergency Plan. Before, Shift Manager can call for additional help, activate TSC, OSC, EOF, JIC or start taking actions to mitigate the consequences of a nuclear accident in progress, the reactor trips, turbine trips, main steam lines over pressurizes due to sudden turbine load rejection. The main steam lines atmospheric valves and/or main steam line relief valves will instantaneously open to prevent the main steam line from over pressurization and start dumping the un-partitioned radioactive coolant containing iodine with steam into the environment. In less than 15 minutes, 60 tons of radioactive coolant contained in the faulty and un-isolatable steam generator, will leak to the environment, melt the fuel in the reactor and release offsite doses in excess of Control Room limit of 5 Rem TEDE, and the Exclusion Area Boundary and Low Population Zone limit of 2.5 Rem TEDE.

Based on the NRC Studies, Independent Safety Experts Observations and observation of SONGS Emergency Plan Drills, San Onofre Emergency Plan is not proven to notify, shelter (Plus Kl Tablets) and evacuate the transients, disabled residents, affected families and children within the 10 mile zone during rush traffic hours in the event of above described a sudden large early frequency release radiological accident. A nuclear fallout from San Onofre can shutdown completely the business at Los Angeles and Long Beach Harbors and chock the already fragile economy of Southern California.

4