ML13109A075

Submission by Friends of the Earth Supplementing Its 10 CFR 2.206 Petition and Responding to Southern California Edison'S 2.206 Response

ML13109A075
Person / Time
Site:	San Onofre, 05000231
Issue date:	02/06/2013
From:	Ayres R Ayres Law Group, Friends of the Earth
To:	Document Control Desk
References
2.206
Download: ML13109A075 (83)

Text

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Group February 6, 2013 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Docket Nos. 50-361 and 50-362 Friends of the Earth 10 C.F.R. 2.206 San Onofre Units 2 & 3 To the Members of the 2.206 Board:

On November 8, 2012, the Commission referred to the Executive Director for Operations the portion of a petition filed earlier by Friends of the Earth alleging that SCE violated 10 C.F.R. § 50.59 when it failed to seek an amendment to its operating license for the installation of replacement steam generators on Units 2 and 3 at San Onofre. This question was referred to the Executive Director of Operations.

On January 15, 2013, Southern California Edison (SCE) submitted a response to the 2.206 board regarding the question before the 2.206 panel considering Friends of the Earth's Petition. At a meeting of the panel on January 16, 2013, Friends of the Earth requested the opportunity to respond to Edison's submission. The attached document is that submission.

Sincerely, 1707 L St, N.W.

• Suite 850 e Washington, DC 20036

• Phone: (202) 452-9200 Web: www.ayreslawgroup.com

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE PETITION REVIEW BOARD Docket Nos. 50-361 and 50-362 Submission by Friends of the Earth Supplementing Its 10 C.F.R. 2.206 Petition and Responding to SCE's 2.206 Response I. BACKGROUND On January 31, 2012, San Onofre Nuclear Generating Station (San Onofre) experienced a steam generator tube leak in Unit 3 that resulted in the release of radioactive material into the environment. Prior to the leak in Unit 3, SCE had discovered excessive wear in Unit 2, which was offline for a refueling outage.

Subsequently, untimely degradation of the walls of many tubes was discovered in the replacement steam generators (RSG) for Units 2 and 3 after, respectively, less than two years and approximately eleven months.

These events were the result of design choices made by Southern California Edison (SCE or Edison) during the specification and construction of the four RSGs for San Onofre Units 2 and 3. In the early 2000s, SCE made a decision that it was time to replace the original steam generators (OSG) at San Onofre, which had given twenty-eight years of relatively trouble-free service.

Edison filed an application for authority to go forward with the RSGs with the California Public Utility Commission (PUC) on February 27, 2004. When SCE sought bidders for the job of building the RSGs, it specified that the RSGs should be designed and constructed such that no license amendment would be required under 10 C.F.R.

50.59.1 SCE asked that the supplier "guarantee in writing that the RSG design is licensable and provide all support necessary to achieve that end.",2 A Design Specification for RSG #2 required the supplier to provide "an engineering evaluation....

justifying that the RSGs can be replaced under the provision of 10 CFR 50.59 (without prior NRC approval)." 3 After the contract was let and Mitsubishi Heavy Industries (MHI) had begun initial work, SCE met with the NRC in June 2006. In this meeting, SCE asserted that the Arnie Gundersen, "2.206 Presentation San Onofre Units 2 and 3 Replacement Steam Generators" (before the NRC 2.206 Petition Review Board, January 16, 2013) (hereinafter "Gundersen Jan. 16, 2013 Presentation"), slide 6.

2 Id. at slide 8.

3 Id. at slide 9.

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RSGs would be built without a license amendment, but provided no 10 CFR 50.59 analysis to support this assertion to the regulators. 4 As the design process proceeded, SCE required the RSGs to be significantly changed by comparison to the OSGs. Edison required 377 more tubes in the RSG, which were to be placed in the center of the tube bundle. This decision required elimination of the stay cylinder 5 that was in the center of the bundle on the OSG. A new "broached plate" design was used for the tube supports, replacing the egg crate design of the OSG.6 Together these and other changes increased the steam quality, reducing the liquid water in the U-bend portions of the RSG as compared to the OSG.7 SCE did not seek a license amendment with regard to these design changes in the RSG.

Such a combination of changes was not made when other generating plants installed RSGs. At Palo Verde, for example, 10% more tubes were incorporated into the RSG but placed at the periphery, not center, of the tube bundle. Though the additional tubes produced 2.9% more heat, Palo Verde has experienced no fluid elastic instability (FEI) problems and virtually no tube damage. 8 Thus, the problem at San Onofre was a foreseeable consequence of certain design decisions made by SCE.

Since beginning operation in 2010/11, the foreseeable and foreseen high void fraction in the four new RSGs created a fluid elastic instability causing thousands of steam generator tubes to vibrate excessively and hit each other and the restraining structures around them. This vibration induced damage created a radiation leak in one RSG tube, and the weakening of seven other tubes that jeopardized the primary radiation containment barrier.

In January 2012, Unit 3 experienced a tube leak that lead to the discovery of extensive tube wear in both Units 2 and 3 after just a short period of operation. The tube degradation in each unit is unlike tube wear in any other replacement steam generators in U.S. plants at the same stage of their useful lives. San Onofre Unit 2 has 1595 degraded tubes; Unit 3 has 1806. Unit 2 has 4721 tubal wear indications; Unit 3 has 10,284. Unit 2 has 510 tubes plugged after one cycle of operation of the replacement steam generators; Unit 3 has 807. SCE and the Nuclear Regulatory Commission (NRC) have reported that 9% of the tubes in Unit 3's steam generators have greater than 10% through-wall wear indications. In Unit 2, 5% of the tubes show such wear.

4 Id. at slide 11.

5 Id. at slide 25.

6 Id. at slide 26.

7 Id. at slide 28.

8 Id. at slide 27.

9 Only one other RSG has suffered a similar amount of tube degradation, but the Advisory Committee on Reactor Safety concluded that the tube wear at that plant is "different than the form of degradation" at San Onofre. Advisory Committee for Reactor Safeguards, NRC, July 23, 2012, letter to R.W. Borchardt, Executive Director for Operations, NRC, "

SUBJECT:

Final Safety Evaluation Report Associated with the Florida Power and Light St. Lucie, Unit 2, License Amendment Request for an Extended Power Uprate," p.

4.

2

In March of 2012, an Augmented Inspection Team (AIT) was dispatched to San Onofre to investigate independently the tube degradation at San Onofre Units 2 and 3.

Both Edison's and AIT's investigations identified FEI as the immediate mechanical cause of the excessive tube wear. But neither determined the root cause of the premature and extensive tube degradation in the RSGs. Lacking such understanding, SCE has not proposed any action to permanently fix the problems of either Unit 2 or 3.

Edison did not perform 10 CFR §50.59 evaluations prior to predetermining that its changes in the steam generator design for San Onofre Units 2 and 3 would not require a license amendment process and NRC evaluation. Instead, Edison decided in 2004 that it would not request a license amendment for the RSGs. Edison then contractually bound Mitsubishi Heavy Industries (MHI) who had the winning bid to construct the RSGs, to avoid the 10 CFR §50.59 license amendment process. Edison met with the NRC in 2006 and informed the NRC that the 10 CFR §50.59 process would not apply, even though Edison had not even begun a §50.59 analysis to reach that conclusion. Indeed, the RSGs were completely fabricated before Edison's 10 CFR §50.59 analyses were even begun.

FoE filed a Petition to Intervene with the Nuclear Regulatory Commission on June 18, 2012. The Commission did not act on FoE's Petition until November 8, 2012, when it referred one issue raised by FoE's Petition to the Executive Director for Operations: whether Edison needed a license amendment to replace the steam generators at San Onofre Units 2 and 3. On December 10, 2012, NRC staff contacted counsel for FoE to begin the 2.206 process. On January 9, 2013 SCE filed a response to FoE's Petition in the 2.206 proceeding in which it argued that SCE could not have foreseen, and did not know, that the design of the RSG had increased the risk that FEI would cause premature tube wear on an unprecedented scale, as well as failure of some tubes.

The NRC held a public meeting on January 16, 2013 in which FoE made a presentation to the staff and answered questions about the Petition.

II.

SUMMARY

OF LEGAL POINTS FoE has provided significant new information to the 2.206 process in two primary veins. Foremost, FoE has provided independent analysis and conclusions by a nuclear engineering expert, Mr. Arnold Gundersen, who has offered a different understanding of the root cause of the tube wear at the San Onofre RSGs that has not been articulated by either the licensee or the NRC. While all parties have at this point identified FEI as the mechanicalcause of the tube wear, SCE and the NRC have not provided an analysis of the root cause of the FEI,as Mr. Gundersen has done.

Mr. Gundersen's initial technical analysis of the problems at San Onfore was borne out by the AIT, which validated his assessment that FEI was the mechanical cause of the tube wear. As SCE and the NRC have not yet offered an explanation for what is causing the FEI, Mr. Gundersen's analysis about the effect of the combination of design 3

changes at San Onofre constitutes significant new information to aid in the resolution of these issues in this 2.206 process.

The deficiency in the design of the RSGs has now been realized in their operation:

the design function of containing radioactive coolant from the primary loop has been shown to have been compromised by the changes SCE made, as Mr. Gundersen has said all along.

SCE has attempted to defend its failure to seek a license amendment by arguing that it had no knowledge of the deficiency associated with the design at the time it performed the 50.59 evaluations. But Edison's Vice President of Engineering has admitted to the NRC staff that this assertion is false.

SCE's 50.59 evaluations, which are based on this misrepresentation, are therefore invalid and cannot support SCE's argument that no license amendment was required.

This fact necessarily affects the NRC's conclusions as well: the NRC's determination that SCE's 50.59 analysis was adequate-which was contingent on SCE's representations about what it knew at the time it performed the evaluation-is therefore undermined and must be revised in light of this new information.

In sum, SCE's 50.59 screening, which determined that no license amendment was required, was not based on all of the information known to SCE at the time. Had SCE's evaluation accurately represented the state of its knowledge-i.e., that the design function of the RSGs would be adversely affected by the changes SCE sought to make to the design-SCE would have been required to seek a license amendment at the time.

NRC's review, as documented in the AIT Report, relied on SCE's misrepresentations and is therefore not based on complete and accurate knowledge about the licensee's activities. Accordingly, the NRC's conclusions to date cannot exonerate SCE for failure to obtain a license amendment. As this argument forms the basis of SCE's request that the Board deny FoE's 2.206 Petition, SCE's request must be denied.

As discussed further below, SCE is currently operating without a valid license.

Accordingly, FoE respectfully requests that the Board grant FoE's converted 2.206 Petition and officially suspend SCE's operating license until SCE receives a license amendment, as FoE requested in the January 16, 2013 Public Meeting with the Petition Review Board on this matter.

III. FOE'S PETITION SATISFIES THE CRITERIA FOR EVALUATION IN MANAGEMENT DIRECTIVE 8.11 The NRC's Management Directive 8.11 (MD 8.11), "Review Process for 10 CFR 2.206 Petitions," lays out the criteria to be used in determining whether to review a petition and whether to reject a petition.' 0 The NRC uses three criteria in most cases to 10Management Directive 8.11 (updated Oct. 25, 2000), available at http://pbadupws.nrc.gov/docs/

ML0417i/ML041770328.pdf 4

consider whether a petition meets the requirements of 10 C.F.R. §2.206. However, MD 8.11 provides that when a Presiding Officer refers a petition to intervene to the 2.206 process, as in this case, the petition is exempt from the first two criteria. The third criterion remains applicable and states:

There is no NRC proceeding available in which the petitioner is or could be a party and through which the petitioner's concerns could be addressed. If there is a proceeding available, for example, if a petitioner raises an issue that he or she has raised or could raise in an ongoing licensing proceeding, the staff will inform the petitioner of the ongoing proceeding and will not treat the request under 10 CFR 2.206.11 The only ongoing NRC proceeding involving FoE's Petition on San Onofre is the one before the Atomic Safety and Licensing Board (ASLB) concerning San Onofre Units 2 and 3. However, the question referred by the Commission to the ASLB for consideration in that proceeding is whether the process set into motion by the March 26, 2012 Confirmatory Action Letter (CAL) is a defacto license amendment proceeding and, if so, whether FoE's Petition to Intervene meets standing and admissibility requirements.

The ASLB has been clear that the scope of that proceeding does not include consideration of the main issue in this 2.206 process, which is whether Edison should have applied for a license amendment for the design and installation of the replacement steam generators.

Thus, FoE's Petition meets the third criterion for 2.206 review under MD 8.11.

MD 8.11 further describes criteria staff must use to determine whether to reject a petition. They are as follows:

1. The incoming correspondence does not ask for an enforcement-related action or fails to provide sufficient facts to support the petition but simply alleges wrongdoing, violations of NRC regulations, or existence of safety concerns...

2. The petitioner raises issues that have already been the subject of NRC staff review and evaluation either on that facility, other similar facilities, or on a generic basis, for which a resolution has been achieved, the issues have been resolved, and the resolution is applicable to the facility in question.. .These requests will not be treated as a 2.206 petition unless they present significant new information.

3. The request is to deny of license application or amendment...

4. The request addresses deficiencies within existing NRC rules... 12 Criteria three and four are not applicable to FoE's Petition since it neither asks the NRC to deny a license application or amendment, nor to address deficiencies in NRC rules.

FoE's Petition is not a general request for enforcement such as that described in the first criterion. The Petition specifically requests that the NRC require SCE to submit

" Id. at 11.

12Id. at 12 (emphasis supplied).

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a license amendment application for the design and installation of the RSGs.13 Moreover, the 23-page document, along with Mr. Gundersen's 17-page declaration, and additional information provided to the Board at the January 16, 2013 meeting and in this submission, give extensive and specific support for FoE's argument that SCE should have sought a license amendment under 10 C.F.R. § 50.59 before it replaced the steam generators. FoE also submits with this filing Mr. Gundersen's affidavit from the ASLB proceeding and three reports written by Fairewinds Associates, an energy consulting company where Mr. Gundersen is Chief 14 Engineer. These documents further elaborate why a license amendment is required.

Although the defects in San Onofre's RSGs have been subjected to some NRC inspection and review, FoE's Petition presents significant new information and thus must be allowed under criterion two. The issue before the NRC staff under this criterion is a threshold issue. In other words, whether a petition meets the criterion determines only whether the NRC will review the petition, not whether the petition succeeds on its merits.

As in the federal courts, the NRC should therefore construe criterion two broadly in favor of the petitioner when determining whether to review a petition. Under the Federal Rules of Civil Procedure, by analogy, district judges must construe the allegations of a complaint broadly in favor of the plaintiff when they decide a motion to dismiss for failure to state a claim, which similarly determines whether the court will hear evidence on a claim."5 NRC's regulations (at 10 C.F.R. Part 2) do not contain a definition of "significant new information." The dictionary definition of "information" is not simply data.

Webster's II New College dictionary defines "information" as "knowledgederived from study, experience, or instruction."'16 FoE's Petition provides precisely that. Its expert, Arnold Gundersen, has studied the available data and, based on his wealth of experience as a nuclear engineer, has provided the Petition Review Board with an accurate root cause analysis and conclusions regarding the effects of the combination of changes in design in the San Onofre RSGs that required Edison to apply for a license amendment under 10 C.F.R. §50.59. His analysis differs in a significant way from that in the NRC staff's review.

The documents submitted and the presentation made by Fairewinds to the Petition Review Board (2013-1-16 FOE 2-206 Fairewinds Presentation) contained considerable 13See, e.g. FoE Petition to Intervene at 3.

14 See Gundersen Affidavit (Jan. 9, 2013), Attachment 2 to Opening Brief of Petitioner Friends of the Earth, In re Southern California Edison Company (NRC Atomic Safety and Licensing Board 2013) (ASLBP No.

13-924-01 -CAL-BDO 1); Steam GeneratorFailuresat San Onofre: The Needfor a Thorough Root Cause Analysis Requires No Early Restart, Fairewinds Associates (Mar. 26, 2012); San Onofre CascadingSteam GeneratorFailuresCreated by Edison. Imprudent Design andFabricationDecisions CausedLeaks, Fairewinds Associates (April 12, 2012); and San Onofre's Steam GeneratorFailuresCould Have Been Prevented,Fairewinds Associates (May 14, 2012). These documents are included with this submission as Attachments 1-4, respectively.

15See, e.g., Hrubec v. Nat'l R.R. Passenger Corp., 981 F.2d 962, 963 (7th Cir. 1992) ("Over and over, appellate courts insist that a complaint not be dismissed unless no relief may be granted 'under any set of facts that could be proved consistent with the allegations.' ") (internal citations omitted).

16 WEBSTER'S II NEW COLLEGE DICTIONARY 569 (1995).

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new information necessary to the NRC's review and as requested by the NRC for its 2.206 Petition Review process.

For example, the technical analysis provided by Mr. Gundersen shows that the malfunctions in the RSGs were entirely foreseeable based on the combination of design features of the RSGs that differed substantially from the design of OSGs. He wrote in his May 31, 2012 declaration:

The maximum quality of the water/steam mixture at the top of the steam generator in the U-Bend region should be approximately 40 to 50 percent, i.e. half water and half steam. With the Mitsubishi design the top of the U-tubes are almost dry in some regions. Without liquid in the mixture, there is no damping 7 against vibration, and therefore a severe fluid-elastic instability developed.'

The July 18, 2012 AIT report acknowledges the correctness of Mr. Gundersen's analysis based on MHIl's evaluations of the causes of tube-to-tube wear.' 8 Unlike the AIT, Mr. Gundersen examined the RSGs' design changes in the aggregate, rather than each change independently. Mr. Gundersen concluded that the totality of the design decisions made by SCE-namely, to remove the stay cylinder from the center of the tube bundle, fill that space with an additional 377 tubes, change the tube lengths, and alter the support structure for the tubes-would inevitably cause the FEI that has in fact occurred in the San Onofre RSGs. By contrast, Edison's and the NRC's analyses identified FEI as the mechanical cause of the damage suffered by the tubes, but did not look further to determine what was responsible for the FEI. Mr. Gundersen's holistic approach to analyzing the root cause of the malfunctions at San Onofre has provided significant new information in the form of a more sophisticated understanding of the root causes of the damage to the San Onofre RSGs that was not articulated by the licensee or the Commission's investigations.

Mr. Gundersen's conclusion that the RSGs' design is the root cause of the tube degradation has been confirmed by analyses undertaken by SCE's contractors, which found that during the design process SCE and its contractors were greatly concerned about the potential for this kind of design to create FEI.

SCE and its contractors understood that the high steam void fraction designed into the San Onofre RSGs increased the risk of tube-damaging FEI. SCE foresaw, well before the RSGs were installed at San Onofre, the high likelihood that FEI would result from the design of the RSG. Despite this understanding, Edison rejected changes in the design to reduce the void fraction because the change was thought to impede the company's ability to claim in its 10 CFR §50.59 analysis that no license amendment was necessary for the RSGs.

17Declaration of Arnold Gundersen, May 31, 2012 ¶ 36 (hereinafter "Gundersen May 31, 2012 Declaration"). Attachment to FoE Petition to Intervene (June 18, 2012).

18 Letter to P. Dietrich from E. Collins re San Onofre Nuclear GenerationStation - NRC Augmented Inspection Team Report 05000361/2012007and 05000362/2012007)(July 18, 2012) at 18.

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Edison has admitted to the NRC that it knew in 2005 that its design would produce the excessive void fraction that caused the FEI and resulting premature tube damage to San Onofre Units 2 and 3. At the November 30, 2012 public meeting with Edison, Greg Werner, NRC Inspection & Assessment Lead, SONGS Project Branch, asked Edison's Vice President for Engineering, Tom Palmisano, the following question:

Just so we are clear, the under-prediction of the velocity by FIT III was not recognized - the problem of the model when it was changed from square pitch to triangular pitch a number of years ago - but the void fraction even under FIT-III while not predicting 99.6% was predicting 95% which was still high and was a matter of concern back in the 2005 timeframe ... that was a matter of concern a number of feasibility studies were conducted to try to lower the void fraction before the steam generators were fabricated but apparently it was not ....

Palmisano's response to the question admitted that Edison understood the void fraction was too high in the 2005 time period:

We... have asked MHI for a better explanation of that and we are looking at it ourselves because as you say the void fraction was high it was not predicted as high 99.5% it was high it was questioned, ultimately the calculations and the operating experience showed even with that void fraction the system should have been effective it was not - clearly that's a failure several reasons for that failure that have to be dealt with.

This statement demonstrates that Edison was fully aware at the design stage of the impact of the design changes it had already ordered on the RSGs (e.g. removal of the stay cylinder, additional tubes located in the center of the bundle, replacement of the egg crate tube supports with broached plate supports, etc.).

Such information has not been considered by NRC staff. For example, in the November 2012 AIT Report, NRC states that because MHI was still evaluating "the potential factors that contributed to the low flow velocities in FIT-III relative to the velocities calculated by the ATHOS model" the issue remains unresolved.' 9 The statement above shows that Edison's insistence on not applying for a license amendment assured that no change to the design of the RSGs was made to reduce the high likelihood of FEI.

Mr. Gundersen's submissions are significant new information that meet the criteria of MD 8.11. Thus, the Board should consider FoE's Petition.

19 Letter to P. Dietrich from E. Collins re San Onofire Nuclear Generation Station - NRC Augmented Inspection Team Report 05000361/2012010and 05000362/2012010)(Nov. 9, 2012) at 20.

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IV. FOE'S 2.206 PETITION SHOULD BE GRANTED BECAUSE SCE'S CURRENT OPERATING LICENSE IS NOT VALID ABSENT AMENDMENT In its presentation to the Petition Review Board on January 16, 2013, FoE clarified the enforcement action it seeks through its converted 2.206 petition: that the Board 1) order SCE to submit a license amendment request; and 2) officially suspend SCE's operating license until a license amendment is approved for the changes SCE made to Units 2 and 3 when it replaced the steam generators in 2009-11.

Suspending SCE's operating license is required because the license, absent amendment, is invalid: it does not account for the adverse effects of the design changes on the steam generators' design function, as required by 10 C.F.R. § 50.59. As explained in FoE's presentation to the Board and discussed further below, SCE's failure to seek a license amendment for these changes is a violation of 10 C.F.R. § 50.59 that requires suspending San Onofre's operating license until it is amended.

A. FoE Has Presented Information Showing that the Design Changes SCE Made to the RSGs Adversely Affected Their Design Function and Satisfied at Least One of the Eight Criteria of 10 C.F.R. § 50.59(c)(2)

In SCE's Response to FoE's converted 2.206 Petition, SCE contends that the changes to the RSGs did not have an adverse effect on their design function and were therefore screened out under 10 C.F.R. § 50.59.20 Thus, SCE asks the Board to accept their modeling exercise and ignore the actual facts of what has occurred at San Onofre.

SCE's assertion-that the design function was not adversely affected by the changes-has been proven incorrect by the leak in one tube during operation and the rupture of eight others under testing, as well as by the degradation of thousands of tubes at both units at San Onofre. This rapid degradation indicates that the design function of containing radioactive coolant from the primary loop has been compromised and the risk of accident increased. Moreover, this result was foreseeable at the time the design changes were made, as discussed further below.

i. Adverse Effect on Design Function According to Section 5.1 of the Final Safety Analysis Report as updated (UFSAR), as cited by SCE, the RSGs have two design functions: "(1) Function as a part of the reactor coolant pressure boundary (RCPB) as a barrier to the release of fission products; and (2) Transfer the heat generated in the reactor from the reactor coolant system into the secondary system.',2 1 In addressing the issue of design function, SCE offers a conclusory statement denying that the design changes identified by FoE had an adverse22 adverse effect on the RCPB function of the RSGs.2 ' SCE erroneously contends that FoE's technical expert, Mr. Gundersen, "provides no explanation or information to 20 SCE Response at 9.

21 id.

22 id.

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support those claims in the table [that "certain design changes would increase the probability or consequences of an accident, create the possibility of a new accident, or have other effects"]. 2324"Therefore," SCE concludes, "the table does not provide a basis for a 50.59 violation."

To the contrary, in several submissions to the NRC Mr. Gundersen has provided a detailed explanation of how the combination of design changes increased the probability or consequences of an accident, including in FoE's initial Petition to Intervene (now converted in part to this 2.206 Petition).2 SCE's claim that no design function was adversely affected by the design changes is in error. Indeed, SCE itself acknowledges an adverse result: in its response to FoE's 2.206 Petition, SCE states that the "adverse condition that later resulted in a tube leak was a deficiency associated with the design."2 6 This statement appears to undermine SCE's denial of no adverse effect.

To the extent that SCE is attempting to differentiate between an adverse "condition" and an adverse "effect" to explain this apparent discrepancy, such a distinction is wanting.2 7 As Mr. Gundersen has repeatedly explained,2 8 tube integrity is an integral part of the reactor coolant pressure boundary in that it serves as a barrier to the release of fission products. To the extent that, in SCE's words, a design "deficiency" created an "adverse condition" resulting in a tube leak-which SCE concedes-it is difficult to see how SCE can continue to maintain the position that the design changes did not adversely affect the design function, or that FoE has failed to make this argument.

Indeed, the differences between the RSGs and the original steam generators are precisely the changes Mr. Gundersen has repeatedly identified as affecting the design function of the RSGs. The combination of conditions related to the thermal hydraulic and tube supporting condition that are now understood to be related to in-plane FEI [the main tube wear mechanism] are as follows:

" Increase in tube bundle heat transfer surface area (11%)

• Increase in number of tubes (5%)

23 Id. at 10.

24 Id.

25 SCE misconceives FoE argument in its Response ("In some places, Mr. Gundersen appears to be taking the position that any changes in the UFSAR requires a license amendment .... That position is inconsistent with 10 CFR 50.59..."). Id. at 10. FoE well understands that a license amendment is not required for every change to the UFSAR; rather, its argument is that the specific combination of changes in question adversely affected a design function (specifically, that the changes affected the reactor coolant pressure boundary) and that this impact on the design function met at least one, if not more, of the eight criteria enumerated in 10 C.F.R. § 50.59(c)(2), as described by Mr. Gundersen in his declaration submitted with FoE's Petition to Intervene, Gundersen May 31, 2012 Declaration, supra note 17.

26 SCE Response at 9.

27 SCE attempts to distinguish what it terms "nonconformances" in the RSGs from design changes that resulted in adverse effects on design function-and therefore required a license amendment. This position is unavailing. As discussed in more detail below, SCE knew about the inherent design defects prior to the finalization of the design and therefore the results cannot be considered later-in-time "nonconforming" conditions.

28 See, e.g., Gundersen May 31, 2012 Declaration, supra note 17.

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• Removal of stay cylinder

• Change from lattice bars to trefoil broached tube support plates

" Change in tube support configuration in U region

• Change from CE to MHI moisture separators

" Power level/operating temperature/tube plugging margin The failure to consider systematically the combined effect of all of the design changes, as Mr. Gundersen has said, was a programmatic cause contributing to the RSG tube wear.

Regarding the technical argument advanced by SCE that a 10 C.F.R. § 50.59 evaluation is not required for changes made to structures, systems, or components whose design function is not specified in the UFSAR, this argument has been rebutted by the NEI 96-07 Guidelines, as endorsed in the AIT Report. Specifically, the AIT Report states:

Per NEI 96-07, changes affecting structures, systems, or components that are not explicitly described in the updated final safety analysis report can have the potential to adversely affect structure, system, or component design functions that are described and thus may require a 10 CFR 50.59 evaluation.29 Thus, while the updated FSAR "did not specify how the original steam generators relied on special design features such as the stay cylinder, tubesheet, tube support plates, or the shape of the tubes to perform the intended safety functions," 30 the licensee is still required to perform a 50.59 evaluation when the changes adversely affect the structure, system, or component design functions.

While SCE did perform a screening comparing the differences in subcomponents between the original steam generators and the RSGs as to whether the differences adversely affected the design function, by failing to consider the effect of the combined suite of changes it had made, SCE clearly arrived at the wrong conclusion, as discussed in detail above.

ii. Fabrication of the RSGs At various points in its 2.206 Response SCE also appears to suggest it was either the fabrication or a deviation from the procurement specifications that caused the RSG tube wear.31 SCE asserts, for example, that the "[tube] leak and unexpected tube wear are nonconforming conditions" such that "[i]f the RSGs had been designed and manufactured in accordance with the procurement specification, the leak and tube wear would not have occurred." 32 29 AIT Report at 36.

30 id.

31See also SCE Response at 10 ("FOE has not demonstrated that any of the design changes (ifproperly implemented) would have adversely affected a design function)" (emphasis added).

32 Id.

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This assertion does not stand up. As Mr. Gundersen stated to the Petition Review Board, the root cause of the tube wear was the RSG design-specifically, the design constraints put on Edison's contractors in an attempt not to trigger a license amendment under 50.59. Thus, the tube damage would have occurred regardless of any fabrication issues due to inherent defects in the design of the RSGs.3 3 Notably, except for SCE, no one has suggested that there were any issues with the fabrication process.

The crux of SCE's argument for why the company was not required to seek a license amendment and therefore why it is not in violation of 10 C.F.R. § 50.59 is that it properly screened out the changes because there was no adverse effect on any design function. SCE has now conceded the design function was indeed adversely affected (the "adverse condition that later resulted in a tube leak was a deficiency associated with the design and was not known at the time the 50.59 was performed"). The necessary conclusion that follows from that acknowledgement is that SCE did not properly screen the changes.

B. The Adverse Effects on the Design Function were Foreseeable and Known to SCE Prior to Finalizing the RSG Design SCE has asserted that the adverse condition later resulting in the tube leak and wear "was not known at the time the 50.59 evaluation was performed."3 4 SCE further states that the tube leak and unexpected tube wear are "nonconforming conditions" and that these "nonconformances were not known at the time the 50.59 evaluations were performed" and therefore "do not indicate any violation of 10 CFR 50.59."35 This claim was rebutted by Mr. Palmisano, Edison's own Vice President for Engineering, quoted previously, at the NRC's public meeting last fall. Mr. Palmisano admitted that Edison knew as early as 2005 that the excessive void fraction designed into the RSGs would cause FEI, which in turn caused significant damage to the tubes. Edison contractors were also aware of this problem. Nevertheless, SCE chose not to make design changes in the RSGs that might have prevented, or mitigated, the FEI because it would have required a license amendment under a 50.59 analysis.

C. SCE's Failure to Seek a License Amendment Constitutes a Violation of 10 C.F.R. § 50.59 SCE argues in its 2.206 Response to this Board that "[l]ater-identified errors in an evaluation or nonconformances do not mean that an earlier 50.59 evaluation, such as SCE's 50.59 analysis for the RSGs, was deficient or that a license amendment should have been obtained."36 This statement cannot be squared with what Edison's Vice President for Engineering said at the NRC public meeting last fall. This information 33 Gundersen Jan. 16, 2013 Presentation, supra note 1, at slide 30.

34 SCE Response at 9.

35id.

36 Id. at 12.

12

completely negates SCE's justification of its 50.59 analysis-i.e., that it is not deficient because the design errors were not identified until after the fact.

SCE was aware of the high probability that tube wear would occur as a result of the RSG design and decided to proceed despite this knowledge. The fact that SCE had this information available to it at the time it performed the 10 C.F.R. § 50.59 evaluations undermines the credibility of its analysis and SCE's reliance on it to argue that no license amendment should have been sought.

In its communication with this Board, Edison states, "[b]ased on the 50.59 evaluations, the Facility Change Report37stated that SCE concluded that the changes could be made without prior NRC approval.

But the Facility Change Report referred to was written in 2009, five years after the RSGs were procured in 2004. Edison informed the NRC in June of 2006 that their 10 CFR § 50.59 process indicated that prior NRC approval for the San Onofre RSGs would not be required, even though it had not even begun a 10 CFR § 50.59 analysis to reach that conclusion. The historical record indicates that early in the design phase for the RSGs and before the NRC was notified in 2006 both Edison and MHI were aware of extraordinarily high void fractions in the RSGs and failed to inform the NRC. The record shows that the RSGs were completely fabricated before the 10 CFR § 50.59 process was even begun.

Edison says that, "[t]he NRC staff already has reviewed whether the RSGs were appropriately evaluated under 10 CFR 50.59." (Edison Enclosure, Page 6) Edison has provided no documentation for its claim that the NRC reviewed a complete analysis of Edison's 2004 determination that a license amendment would not be required for San Onofre's Unit 2 and 3 Replacement Steam Generators. Additionally, Edison has provided no documentation for its claim that the NRC reviewed Edison's 2006 statements that the NRC approved Edison's conclusion that a complete 10 C.F.R. § 50.59 process was not necessary.

D. The NRC's Conclusions in the AIT Report Regarding SCE's Compliance with 50.59 Were Based on Incomplete Knowledge and Therefore Cannot Serve to Exonerate SCE for Failure to Seek a License Amendment SCE's request that the Board deny FoE's 2.206 Petition is based almost entirely on its claim that the NRC staff has already reviewed the issue of whether SCE properly evaluated the RSGs under 50.59 and found that the design changes were appropriately evaluated. 38 The NRC assessment included reviews by the Agency at the time the steam generators were replaced as well as the staff's subsequent review of SCE's 10 C.F.R. § 50.59 evaluations as documented in the July and November 2012 AIT Reports.

37 38 Id. at 2.

d. at 2, 6-8.

13

This argument is wholly negated by Mr. Palmisano's testimony and information available from other sources. As described above, SCE had knowledge of the defects inherent to the design of the RSGs prior to conducting the 50.59 evaluations. This information shows SCE's statements that the company had no knowledge of a "deficiency associated with the design" "at the time the 50.59 evaluation was performed" 39 to be false. Consequently, SCE's 10 C.F.R. § 50.59 screening and evaluations-which ignored this information-are fatally undermined by this revelation.

The same is true of the NRC's conclusions about the adequacy of SCE's analysis, as they are contingent on representations made by SCE that have turned out to be false.

To the best of FoE's knowledge, the NRC did not have the information provided by Mr. Palmisano and others available to it when it reviewed SCE's 50.59 evaluations, either at the time of the RSGs' replacement or in the course of conducting the AIT inspection. This information sheds an entirely new light on SCE's 50.59 evaluations and shows that SCE should have sought a license amendment at the time it performed the RSG replacement based on what it knew then. The NRC's determination that SCE complied with the requirements of 50.59-SCE's central argument in its Response to FoE's 2.206 Petition-is therefore invalidated by this information demonstrating that SCE was clearly on notice of the design defect-not retrospectively, but at the time it performed its 10 C.F.R. § 50.59 evaluations.

The AIT Report, which SCE contends "evaluated and rejected the very issues raised by FOE regarding design changes"' 40 states that "10 CFR 50.59 does not require the licensee to presume deficiencies in the design or fabrication."4 No such presumption is needed in this case: SCE knew about the design deficiencies. The AIT's determination that SCE reviewed the major design changes in accordance with 50.59's requirements is thus no longer accurate and must be revised.

SCE asks this Board to reject FoE's Petition on the grounds that 1) "the NRC's evaluation of the 50.59 process for the RSGs encompasses all of the issues raised by FOE" in the 2.206 Petition, and 2) that FoE does not identify any significant new information. Neither of these points is correct.

Had the NRC understood what Edison knew about the design deficiencies in the RSGs, and when the company knew about them, it would have been able to assess SCE's 50.59 screening and evaluation, as well as its subsequent defense thereof, from a differently informed perspective. Because this was not the case, however, FoE contends that the NRC review and its conclusions about the adequacy of SCE's 50.59 evaluations cannot serve to exonerate SCE.

39 4 Id. at 9.

1Id. at 10.

41 AIT Report at 36.

14

V. CONCLUSION For the reasons presented above, and in the statements made by Mr. Ayres and Mr. Gundersen before the Board on January 16, 2013, FoE urges the Board to: (1) order SCE to submit a license amendment request; and (2) suspend SCE's operating license until a license amendment is approved for the changes SCE made to the replacement steam generators at San Onofre Units 2 and 3 for which it should previously have sought a license amendment.

Respectfully submitted,

/Richard Ayres/

Richard E. Ayres Counsel to Friends of the Earth 15

Attachment 1 Gundersen Affidavit

CURRICULUM VITAE Arnold Gundersen Chief Engineer, Fairewinds Associates, Inc December 2012 Education and Trainine ME NE Master of Engineering Nuclear Engineering Rensselaer Polytechnic Institute, 1972 U.S. Atomic Energy Commission Fellowship Thesis: Cooling Tower Plume Rise BS NE Bachelor of Science Nuclear Engineering Rensselaer Polytechnic Institute, Cum Laude, 1971 James J. Kerrigan Scholar RO Licensed Reactor Operator, U.S. Atomic Energy Commission License # OP-3014 Qualifications- including andnot limited to:

" Chief Engineer, Fairewinds Associates, Inc

• Nuclear Engineering, Safety, and Reliability Expert

" Federal and Congressional hearing testimony and Expert Witness testimony

• Former Senior Vice President Nuclear Licensee

" Former Licensed Reactor Operator

• Atomic Energy Commission Fellow

• 40-years of nuclear industry experience and oversight o Nuclear engineering management assessment and prudency assessment o Nuclear power plant licensing and permitting - assessment and review o Nuclear safety assessments, source term reconstructions, dose assessments, criticality analysis, and thermohydraulics o Contract administration, assessment and review o Systems engineering and structural engineering assessments o Cooling tower operation, cooling tower plumes, thermal discharge assessment, and consumptive water use o Nuclear fuel rack design and manufacturing, nuclear equipment design and manufacturing, and technical patents o Radioactive waste processes, storage issue assessment, waste disposal and decommissioning experience o Reliability engineering and aging plant management assessments, in-service inspection o Employee awareness programs, whistleblower protection, and public communications o Quality Assurance (QA) & records Publications Published Lecture - The Lessons of the Fukushima DaiichiNuclear Accident published in the InternationalSymposium on the Truth of Fukushima Nuclear Accident and the Myth of Nuclear Safety, August 30, 2012 University of Tokyo, Iwanami Shoten Publishers, Tokyo, Japan Author - The Echo Chamber: Regulatory Captureand the Fukushima DaiichiDisaster,

Page 2 of 16 Lessons From Fukushima, February 27, 2012, Greenpeace, Co-author - Fukushima Daiichi."Truth And The Way Forward,Shueisha Publishing, February 17, 2012, Tokyo, Japan.

Co-author - FairewindsAssociates 2009-2010 Summary to JFC,July 26, 2010 State of Vermont, Joint Fiscal Office, (http://www.leg.state.vt.us/jfo/envy.aspx).

Co-author - Supplemental Report of the Public OversightPanel Regarding the Comprehensive Reliability Assessment of the Vermont Yankee Nuclear Power Plant July 20, 2010, to the Vermont State Legislature by the Vermont Yankee Public Oversight Panel.

Co-author - The Second Quarterly Report by Fairewinds Associates, Inc to the Joint Legislative Committee regarding buried pipe and tank issues at Entergy Nuclear Vermont Yankee and Entergy proposed Enexus spinoff. See two reports: FairewindsAssociates 2nd QuarterlyReport to JFCand Enexus Review by FairewindsAssociates.

Author - Fairewinds Associates, Inc FirstQuarterlyReport to the Joint Legislative Committee, October 19, 2009.

Co-author - Report of the Public Oversight Panel Regarding the Comprehensive Reliability Assessment of the Vermont Yankee Nuclear Power Plant,March 17, 2009, to the Vermont State Legislature by the Vermont Yankee Public Oversight Panel.

Co-author - Vermont Yankee Comprehensive Vertical Audit - VYCVA -Recommended Methodology to Thoroughly Assess Reliability and Safety Issues at Entergy Nuclear Vermont Yankee, January30, 2008 Testimony to Finance Committee Vermont Senate.

Co-author - Decommissioning Vermont Yankee - Stage 2 Analysis of the Vermont Yankee DecommissioningFund- The DecommissioningFund Gap, December 2007, Fairewinds Associates, Inc. Presented to Vermont State Senators and Legislators.

Co-author - Decommissioning the Vermont Yankee Nuclear Power Plant.- An Analysis of Vermont Yankee's DecommissioningFund andIts ProjectedDecommissioningCosts, November 2007, Fairewinds Associates, Inc.

Co-author - DOE DecommissioningHandbook, FirstEdition, 1981-1982, invited author.

Presentations & Media Fairewinds Energy Education Corp 501c3 presentations:

" A Mountain of Waste 70 Years High, Presentation: Old and New Reactors, University of Chicago, December 1, 2012

• Congressional Briefing September 20, 2012; invited by Representative Dennis Kucinich

• Presentations in Japan August/September 2012: Presentation at University of Tokyo (August 30, 2012), Presentation at Japanese Diet Building (members of the Japanese Legislature - August 31, 2012), Presentation to citizen groups in Niigata (September 1, 2012), Presentations to citizen groups in Kyoto (September 4 , 2012), Presentation to Japanese Bar Association (September 2, 2012), and Presentation at the Tokyo Olympic Center (September 6, 2012)

• Multi-media Opera: Curtain of Smoke, by Filmmaker Karl Hoffman, Composer Andrea Molino, and Dramatist Guido Barbieri, Rome, Italy (2012-5-21,22)

• Curtain of Smoke Symposium (2012-5-21), with Dr. Sherri Ebadi 2004 Nobel Laureate

• The Italian National Press Club Rome (2012-5-21) with Dr. Sherri Ebadi 2004 Nobel Laureate: the relationship between nuclear power and nuclear weapons

" Radio 3 Rome (2012-5-21) Discussion of Three Mile Island and the triple meltdown at Fukushima Daiichi (Japan),

Page 3 of 16

" Sierra Club Panel Discussions (2012-5-5): Consequences of Fukushima Daiichi with Paul Gunter and Waste Disposal with Mary Olson,

• Physicians for Social Responsibility Seattle (2012-3-17),

• Fukushima Daiichi Forum with Chiho Kaneko, Brattleboro, VT (2012-3-11),

• Physicians for Global Responsibility Vancouver (2012-3-11) Skype Video Lecture, University of Vermont (2 - 2011),

• Boston Nuclear Forum, Boston Library (6/16/11),

• Duxbury Emergency Management (6/15/11),

• Vermont State Nuclear Advisory Panel (VSNAP), Elder Education Enrichment,

• New Jersey Environmental Federation (5/14/11),

• Quaker Meeting House,

• Press Conference for Physicians for Social Responsibility (5/19/11),

" St. Johnsbury Academy - Nuclear Power 101.

Educational videos on nuclear safety, reliability and engineering particularly Fukushima issues.

Videos may be viewed @ fairewinds.org (501 c3 non-profit)

Expert commentary (many more unnamed): CNN (6), The John King Show (14), BBC, CBC, Russia Today, Democracy Now, KPBS (Radio & TV) VPR, WPTZ, WCAX, WBAI, CCTV, NECN, Pacifica Radio, CBC (radio & TV) (4), Rachel Maddow Show, Washington Post, New York Times, The Guardian,Bloomberg (print & TV), Reuters, Associated Press, The Global Post, Miami Herald,Tampa Times, Orange County Times, LA Times, Al Jazeera (print), The Tennessean, The Chris Martinson Show, Mainichi News, TBS Japan, Gendai Magazine, NHK television, Scientific American. Huffington Post (Paris) named Fairewinds.com the best go to site for information about the Fukushima Daiichi accident (5/9/11).

Patents Energy Absorbing Turbine Missile Shield- U.S. Patent # 4,397,608 - 8/9/1983 Committee Memberships Vermont Yankee Public Oversight Panel, appointed 2008 by President Pro-Tem Vermont Senate National Nuclear Safety Network - Founding Board Member Three Rivers Community College - Nuclear Academic Advisory Board Connecticut Low Level Radioactive Waste Advisory Committee - 10 years, founding member Radiation Safety Committee, NRC Licensee - founding member ANSI N-198, Solid Radioactive Waste Processing Systems Honors U.S. Atomic Energy Commission Fellowship, 1972 B.S. Degree, Cum Laude, RPI, 1971, 1st in nuclear engineering class Tau Beta Pi (Engineering Honor Society), RPI, 1969 - 1 of 5 in sophomore class of 700 James J. Kerrigan Scholar 1967-1971 Teacher of the Year - 2000, Marvelwood School Publicly commended to U.S. Senate by NRC Chairman, Ivan Selin, in May 1993 - "It is true.. .everything Mr. Gundersen said was absolutely right; he performed quite a service."

Page 4 of 16 Expert Witness Testimony and Nuclear Engineering Analysis and Consultin2 Expert Witness Report For Friends Of The Earth -_July 11, 2012 San Onofre's Steam Generators:Significantly Worse Thank All Others Nationwide Expert Witness Report For Friends Of The Earth - May 15, 2012 San Onofre's Steam GeneratorFailuresCouldHave Been Prevented,Fairewinds Associates Expert Witness Report For Friends Of The Earth - April 10, 2012 San Onofre CascadingSteam GeneratorFailuresCreatedBy Edison: Imprudent Design And FabricationDecisions CausedLeaks, Fairewinds Associates Expert Witness Report For Friends Of The Earth - March 27, 2012 Steam GeneratorFailuresAt San Onofre."The Need ForA Thorough Root Cause Analysis Requires No Early Restart Expert Witness Report For Greenpeace - February 27, 2012 Lessons From Fukushima: The Echo Chamber Effect, Fairewinds Associates Nuclear Regulatory Commission - December 21, 2011 Expert witness report to Atomic Safety and Licensing Board: PrefiledDirect Testimony of Arnold Gundersen Regarding ConsolidatedContention RK-EC-3/CW-EC-1 (Spent Fuel Pool Leaks)

New York State Department Of Environmental Conservation - November 15-16, 2011 Expert witness for Riverkeeper: hearing testimony regarding license extension application for Indian Point Units 2 and 3 - contention: tritium in the groundwater.

Nuclear Regulatory Commission - November 10, 2011 Expert witness report entitled: Fukushima and the Westinghouse-Toshiba APIO00, A Report for the AP1000 Oversight Group by FairewindsAssociates, Inc, and Video. Submitted to NRC by the AP1000 Oversight Group.

Nuclear Regulatory Commission - October 7, 2011 Testimony to the NRC Petition Review BoardRe: Mark I Boiling Water Reactors, Petition for NRC to shut down all BWR Mark 1 nuclear power plants due to problems in containment integrity in the Mark I design.

New York State Department Of Environmental Conservation, October 4, 2011 PrefiledRebuttal Testimony Of Arnold Gundersen On Behalf Of PetitionersRiverkeeper, Inc.,

Scenic Hudson, Inc., And NaturalResources Defense Council, Inc. To The Direct Testimony Of Matthew J. Barvenik (Senior PrincipalGZA Geoenvironmental,Inc.) Regarding Radiological Materials

Page 5 of 16 Southern Alliance for Clean Energy (SACE) submission to TVA Board of Directors - August 3, 2011- Expert witness report entitled: The Risks of Reviving TVA 's Bellefonte Project, and Video prepared for the Southern Alliance for Clean Energy (SACE).

New York State Department Of Environmental Conservation, July 22, 2011 Prefiled Direct Testimony Of Arnold Gundersen On Behalf Of PetitionersRiverkeeper, Inc.,

Scenic Hudson, Inc., And NaturalResources Defense Council, Inc. Regarding Radiological Materials Nuclear Regulatory Commission - May 10, 2011 Comment to the proposedrule on the AP1000 Design CertificationAmendment Docket ID NRC-2010-0131 As noticed in the FederalRegister on February24, 2011 Retained by Friends of the Earth as Expert Witness.

Nuclear Regulatory Commission - May 10, 2011 Comment to the proposedrule on the API 000 Design CertificationAmendment Docket ID NRC-2010-0131 As noticed in the FederalRegister on February24, 2011 Retained by Friends of the Earth as Expert Witness.

NRC Advisory Committee on Reactor Safeguards (ACRS) - May 26, 2011 Lessons learned from Fukushima and Containment Integrity on the AP1000.

Vermont Energy Cooperative (VEC) - April 26, 2011 Vermont Yankee - Is It Reliablefor 20 more years?

Vermont State Nuclear Advisory Panel (VSNAP) - February 22, 2011 Testimony and presentation entitled the Vermont Yankee Public Oversight PanelSupplemental Report regarding management issues at the Vermont Yankee Nuclear Power Plant to the reconvened Vermont State Nuclear Advisory Panel.

Vermont State Legislature Senate Committee On Natural Resources And Energy February 8, 2011. Testimony: Vermont Yankee Leaks and Implications.

(http://www.leg.state.vt.us/jfo/envy.aspx)

Vermont State Legislature - January 26, 2011 House Committee On Natural Resources And Energy, and Senate Committee On Natural Resources And Energy Testimony regarding Fairewinds Associates, Inc's report: Decommissioningthe Vermont Yankee Nuclear Power Plant and Storing Its Radioactive Waste (http://www.leg.state.vt.us/jfo/envy.aspx). Additional testimony was also given regarding the newest radioactive isotopic leak at the Vermont Yankee nuclear power plant.

Vermont State Legislature Joint Fiscal Committee Legislative Consultant Regarding Entergy Nuclear Vermont Yankee Decommissioning the Vermont Yankee Nuclear Power Plant and Storing Its Radioactive Waste January 2011. (http://www.leg.state.vt.us/jfo/envy.aspx).

Page 6 of 16 U.S. Nuclear Regulatory Commission Advisory Committee on Reactor Safeguards (NRC-ACRS) AP 1000 Sub-Committee Nuclear Containment Failures:Ramificationsfor the API00 Containment Design, Supplemental Report submitted December 21, 2010. (http://fairewinds.com/reports)

Vermont State Legislature Joint Fiscal Committee Legislative Consultant Regarding Entergy Nuclear Vermont Yankee Reliability OversightEntergy Nuclear Vermont Yankee, December 6, 2010. Discussion regarding the leaks at Vermont Yankee and the ongoing monitoring of those leaks and ENVY's progress addressing the 90-items identified in Act 189 that require remediation. (http://www.leg.state.vt.us/jfo/envy.aspx).

U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board (NRC-ASLB)

DeclarationOfArnold Gundersen SupportingBlue Ridge EnvironmentalDefense League's Contention Regarding Consumptive Water Use At Dominion Power's Newly ProposedNorth Anna Unit 3 PressurizedWater Reactor in the matter of Dominion Virginia Power North Anna Power Station Unit 3 Docket No.52-017 Combined License Application ASLBP#08-863 COL, October 2, 2010.

U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board (NRC-ASLB)

DeclarationOfArnold Gundersen Supporting Blue Ridge Environmental Defense League's New Contention RegardingAPIO00 Containment Integrity On The Vogtle Nuclear Power Plant Units 3 And 4 in the matter of the Southern Nuclear Operating Company Vogtle Electric Generating Plant, Units 3&4 Combined License Application, Docket Nos. 52-025-COL and 52-026-COL and ASLB No. 09-873-01 -COL-BDO1, August 13, 2010.

Vermont State Legislature Joint Fiscal Committee Legislative Consultant Regarding Entergy Nuclear Vermont Yankee - July 26, 2010 Summation for 2009 to 2010 Legislative Year For the Joint Fiscal Committee Reliability Oversight Entergy Nuclear Vermont Yankee (ENVY) Fairewinds Associates 2009-20 10. This summary includes an assessment of ENVY's progress (as of July 1, 2010) toward meeting the milestones outlined by the Act 189 Vermont Yankee Public Oversight Panel in its March 2009 report to the Legislature, the new milestones that have been added since the incident with the tritium leak and buried underground pipes, and the new reliability challenges facing ENVY, Entergy, and the State of Vermont. (http://www.leg.state.vt.us/jfo/envy.aspx)

U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board (NRC-ASLB)

DeclarationOfArnold Gundersen Supporting Blue Ridge EnvironmentalDefense League's Contentions in the matter of Dominion Virginia Power North Anna Station Unit 3 Combined License Application, Docket No.52-017, ASLBP#08-863-01-COL, July 23, 2010.

Florida Public Service Commission (FPSC)

Licensing and construction delays due to problems with the newly designed Westinghouse AP1000 reactors in Direct Testimony In Re: Nuclear Plant Cost Recovery Clause By The Southern Alliance For Clean Energy (SA CE), FPSC Docket No. 100009-El, July 8, 2010.

U.S. Nuclear Regulatory Commission Advisory Committee on Reactor Safeguards (NRC-

Page 7 of 16 ACRS) AP 1000 Sub-Committee Presentation to ACRS regarding design flaw in AP 1000 Containment - June 25, 2010 Power Point Presentation: http://fairewinds.com/content/ap 1000-nuclear-design-flaw-addressed-to-nrc-acrs.

U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board (NRC-ASLB)

Second Declaration Of Arnold Gundersen Supporting Supplemental Petition Of Intervenors Contention 15: DTE COLA Lacks StatutorilyRequired Cohesive QA Program- June 8, 2010.

NRC Chairman Gregory Jaczko, ACRS, Secretary of Energy Chu, and the White House Office of Management and Budget API 000 Containment Leakage Report FairewindsAssociates - Gundersen, Hausler,4-21-2010.

This report, commissioned by the AP1000 Oversight Group, analyzes a potential flaw in the containment of the AP 1000 reactor design.

Vermont State Legislature House Committee On Natural Resources And Energy - April 5, 2010 Testified to the House Committee On Natural Resources And Energy - regarding discrepancies in Entergy's TLG Services decommissioning analysis. See Fairewinds Cost Comparison TLG Decommissioning(http://www.leg.state.vt.us/jfo/envy.aspx).

Vermont State Legislature Joint Fiscal Committee Legislative Consultant Regarding Entergy Nuclear Vermont Yankee - February 22, 2010 The Second Quarterly Report by Fairewinds Associates, Inc to the Joint Legislative Committee regarding buried pipe and tank issues at Entergy Nuclear Vermont Yankee and Entergy proposed Enexus spinoff. See two reports: FairewindsAssociates 2nd QuarterlyReport to JFCand Enexus Review by FairewindsAssociates. (http://www.leg.state.vt.us/jfo/envy.aspx).

Vermont State Legislature Senate Natural Resources - February 16, 2010 Testified to Senate Natural Resources Committee regarding causes and severity of tritium leak in unreported buried underground pipes, status of Enexus spinoff proposal, and health effects of tritium.

Vermont State Legislature Senate Natural Resources - February 10, 2010 Testified to Senate Natural Resources Committee regarding causes and severity of tritium leak in unreported buried underground pipes. http://www.youtube.com/watch?v=36HJiBrJSxE Vermont State Legislature Senate Finance - February 10, 2010 Testified to Senate Finance Committee regarding A Chronicle of Issues Regarding Buried Tanks and UndergroundPipingat VT Yankee. (http://www.leg.state.vt.us/j fo/envy.aspx).

Vermont State Legislature House Committee On Natural Resources And Energy - January 27, 2010 A Chronicle of Issues RegardingBuried Tanks and UndergroundPipingat VT Yankee.

(http://www.leg.state.vt.us/jfo/envy.aspx).

Submittal to Susquehanna River Basin Commission, by Eric Epstein - January 5, 2010 Expert Witness Report OfArnold Gundersen Regarding Consumptive Water Use Of The

Page 8 of 16 SusquehannaRiver By The ProposedPPL Bell Bend Nuclear PowerPlant In the Matter of RE:

Bell Bend Nuclear Power Plant Application for Groundwater Withdrawal Application for Consumptive Use BNP-2009-073.

U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board (NRC-ASLB)

Declarationof Arnold Gundersen Supporting Supplemental Petitionof Intervenors Contention 15: Detroit Edison COLA Lacks Statutorily Required Cohesive QA Program, December 8, 2009.

U.S. NRC Region III Allegation Filed by Missouri Coalition for the Environment Expert Witness Report entitled: Comments on the Callaway SpecialInspection by NRC Regarding the May 25, 2009 Failureof its Auxiliary FeedwaterSystem, November 9, 2009.

Vermont State Legislature Joint Fiscal Committee Legislative Consultant Regarding Entergy Nuclear Vermont Yankee Oral testimony given to the Vermont State Legislature Joint Fiscal Committee October 28, 2009.

See report: Qufarterly Status Report - ENVY Reliability Oversightfor JFO (http://www.leg.state.vt.us/jfo/envy.aspx).

Vermont State Legislature Joint Fiscal Committee Legislative Consultant Regarding Entergy Nuclear Vermont Yankee The First Quarterly Report by Fairewinds Associates, Inc to the Joint Legislative Committee regarding reliability issues at Entergy Nuclear Vermont Yankee, issued October 19, 2009.

See report: Quarterly Status Report - ENVY Reliability Oversightfor JFO (http://www.leg.state.vt.us/j fo/envy.aspx).

Florida Public Service Commission (FPSC)

Gave direct oral testimony to the FPSC in hearings in Tallahassee, FL, September 8 and 10, 2009 in support of Southern Alliance for Clean Energy (SACE) contention of anticipated licensing and construction delays in newly designed Westinghouse AP 1000 reactors proposed by Progress Energy Florida and Florida Power and Light (FPL).

Florida Public Service Commission (FPSC)

NRC announced delays confirming my original testimony to FPSC detailed below. My supplemental testimony alerted FPSC to NRC confirmation of my original testimony regarding licensing and construction delays due to problems with the newly designed Westinghouse AP 1000 reactors in Supplemental Testimony In Re: Nuclear Plant Cost Recovery Clause By The Southern Alliance For Clean Energy, FPSC Docket No. 090009-EI, August 12, 2009.

Florida Public Service Commission (FPSC)

Licensing and construction delays due to problems with the newly designed Westinghouse AP 1000 reactors in Direct Testimony In Re: Nuclear Plant Cost Recovery Clause By The Southern Alliance For Clean Energy (SA CE), FPSC Docket No. 090009-EI, July 15, 2009.

Vermont State Legislature Joint Fiscal Committee Expert Witness Oversight Role for Entergy Nuclear Vermont Yankee (ENVY)

Contracted by the Joint Fiscal Committee of the Vermont State Legislature as an expert witness

Page 9 of 16 to oversee the compliance of ENVY to reliability issues uncovered during the 2009 legislative session by the Vermont Yankee Public Oversight Panel of which I was appointed a member along with former NRC Commissioner Peter Bradford for one year from July 2008 to 2009.

Entergy Nuclear Vermont Yankee (ENVY) is currently under review by Vermont State Legislature to determine if it should receive a Certificate for Public Good (CPG) to extend its operational license for another 20-years. Vermont is the only state in the country that has legislatively created the CPG authorization for a nuclear power plant. Act 160 was passed to ascertain ENVY's ability to run reliably for an additional 20 years. Appointment from July 2009 to May 2010.

U.S. Nuclear Regulatory Commission Expert Witness Declaration regarding Combined Operating License Application (COLA) at North Anna Unit 3 DeclarationofArnold Gundersen Supporting Blue Ridge Environmental Defense League's Contentions (June 26, 2009).

U.S. Nuclear Regulatory Commission Expert Witness Declaration regarding Through-wall Penetration of Containment Liner and Inspection Techniques of the Containment Liner at Beaver Valley Unit I Nuclear Power Plant Declarationof Arnold Gundersen Supporting Citizen Power'sPetition (May 25, 2009).

U.S. Nuclear Regulatory Commission Expert Witness Declaration regarding Quality Assurance and Configuration Management at Bellefonte Nuclear Plant DeclarationofArnold Gundersen SupportingBlue Ridge EnvironmentalDefense League's Contentions in their Petitionfor Intervention and Request for Hearing,May 6, 2009.

Pennsylvania Statehouse Expert Witness Analysis presented in formal presentation at the Pennsylvania Statehouse, March 26, 2009 regarding actual releases from Three Mile Island Nuclear Accident. Presentation may be found at: http://www.tmia.com/march26 Vermont Legislative Testimony and Formal Report for 2009 Legislative Session As a member of the Vermont Yankee Public Oversight Panel, I spent almost eight months examining the Vermont Yankee Nuclear Power Plant and the legislatively ordered Comprehensive Vertical Audit. Panel submitted Act 189 Public Oversight Panel Report March 17, 2009 and oral testimony to a joint hearing of the Senate Finance and House Committee On Natural Resources And Energy March 19, 2009.

http://www.leg.state.vt.us/JFOiVermont%20Yankee.htm Finestone v FPL (11/2003 to 12/2008) Federal Court Plaintiffs' Expert Witness for Federal Court Case with Attorney Nancy LaVista, from the firm Lytal, Reiter, Fountain, Clark, Williams, West Palm Beach, FL. This case involved two plaintiffs in cancer cluster of 40 families alleging that illegal radiation releases from nearby nuclear power plant caused children's cancers. Production request, discovery review, preparation of deposition questions and attendance at Defendant's experts for deposition, preparation of expert witness testimony, preparation for Daubert Hearings, ongoing technical

Page 10 of 16 oversight, source term reconstruction and appeal to Circuit Court.

U.S. Nuclear Regulatory Commission Advisory Committee Reactor Safeguards (NRC-ACRS)

Expert Witness providing oral testimony regarding Millstone Point Unit 3 (MP3) Containment issues in hearings regarding the Application to Uprate Power at MP3 by Dominion Nuclear, Washington, and DC. (July 8-9, 2008).

Appointed by President Pro-Tem of Vermont Senate Shumlin (now Vermont Governor Shumlin) to Legislatively Authorized Nuclear Reliability Public Oversight Panel To oversee Comprehensive Vertical Audit of Entergy Nuclear Vermont Yankee (Act 189) and testify to State Legislature during 2009 session regarding operational reliability of ENVY in relation to its 20-year license extension application. (July 2, 2008 to present).

U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board (NRC-ASLB)

Expert Witness providing testimony regarding Pilgrim Watch's Petitionfor Contention 1 UndergroundPipes (April 10, 2008).

U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board (NRC-ASLB)

Expert Witness supporting Connecticut Coalition Against Millstone In Its Petition ForLeave To Intervene, Request For Hearing,And Contentions Against Dominion Nuclear Connecticut Inc. 's Millstone Power Station Unit 3 License Amendment Request ForStretch Power Uprate (March 15, 2008).

U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board (NRC-ASLB)

Expert Witness supporting Pilgrim Watch's PetitionFor Contention 1: specific to issues regardingthe integrity of PilgrimNuclear Power Station's undergroundpipes and the ability of Pilgrim'sAging Management Programto determine their integrity. (January 26, 2008).

Vermont State House - 2008 Legislative Session House Committee on Natural Resources and Energy - Comprehensive Vertical Audit: Why NRC Recommends a VerticalAudit for Aging Plants Like Entergy Nuclear Vermont Yankee (ENVY)

• House Committee on Commerce - Decommissioning Testimony Vermont State Senate - 2008 Legislative Session

• Senate Finance - testimony regarding Entergy Nuclear Vermont Yankee Decommissioning Fund

• Senate Finance - testimony on the necessity for a Comprehensive Vertical Audit (CVA) of Entergy Nuclear Vermont Yankee

• House Committee on Natural Resources and Energy - testimony regarding the placement of high-level nuclear fuel on the banks of the Connecticut River in Vernon, VT U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board (NRC-ASLB)

MOX Limited Appearance Statement to Judges Michael C. Farrar (Chairman), Lawrence G.

McDade, and Nicholas G. Trikouros for the "Petitioners": Nuclear Watch South, the Blue Ridge Environmental Defense League, and Nuclear Information & Resource Service in support of

Page 11 of 16 Contention 2: Accidental Release of Radionuclides, requestinga hearing concerningfaulty accident consequence assessments made for the MOX plutoniumfielfactory proposedfor the Savannah River Site. (September 14, 2007).

Appeal to the Vermont Supreme Court (March 2006 to 2007)

Expert Witness Testimony in support of New EnglandCoalition'sAppeal to the Vermont Supreme Court Concerning: DegradedReliability at Entergy Nuclear Vermont Yankee as a Result of the Power Uprate. New England Coalition represented by Attorney Ron Shems of Burlington, VT.

State of Vermont Environmental Court (Docket 89-4-06-vtec 2007)

Expert witness retained by New England Coalition to review Entergy and Vermont Yankee's analysis of alternative methods to reduce the heat discharged by Vermont Yankee into the Connecticut River. Provided Vermont's Environmental Court with analysis of alternative methods systematically applied throughout the nuclear industry to reduce the heat discharged by nuclear power plants into nearby bodies of water and avoid consumptive water use. This report included a review of the condenser and cooling tower modifications.

U.S. Senator Bernie Sanders and Congressman Peter Welch (2007)

Briefed Senator Sanders, Congressman Welch and their staff members regarding technical and engineering issues, reliability and aging management concerns, regulatory compliance, waste storage, and nuclear power reactor safety issues confronting the U.S. nuclear energy industry.

State of Vermont Legislative Testimony to Senate Finance Committee (2006)

Testimony to the Senate Finance Committee regarding Vermont Yankee decommissioning costs, reliability issues, design life of the plant, and emergency planning issues.

U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board (NRC-ASLB)

Expert witness retained by New England Coalition to provide Atomic Safety and Licensing Board with an independent analysis of the integrity of the Vermont Yankee Nuclear Power Plant condenser (2006).

U.S. Senators Jeffords and Leahy (2003 to 2005)

Provided the Senators and their staffs with periodic overview regarding technical, reliability, compliance, and safety issues at Entergy Nuclear Vermont Yankee (ENVY).

IOCFR 2.206 filed with the Nuclear Regulatory Commission (July 2004)

Filed 10CFR 2.206 petition with NRC requesting confirmation of Vermont Yankee's compliance with General Design Criteria.

State of Vermont Public Service Board (April 2003 to May 2004)

Expert witness retained by New England Coalition to testify to the Public Service Board on the reliability, safety, technical, and financial ramifications of a proposed increase in power (called an uprate) to 120% at Entergy's 31-year-old Vermont Yankee Nuclear Power Plant.

Page 12 of 16 International Nuclear Safety Testimony Worked for ten days with the President of the Czech Republic (Vaclav Havel) and the Czech Parliament on their energy policy for the 21 st century.

Nuclear Regulatory Commission (NRC) Inspector General (IG)

Assisted the NRC Inspector General in investigating illegal gratuities paid to NRC Officials by Nuclear Energy Services (NES) Corporate Officers. In a second investigation, assisted the Inspector General in showing that material false statements (lies) by NES corporate president caused the NRC to overlook important violations by this licensee.

State of Connecticut Legislature Assisted in the creation of State of Connecticut Whistleblower Protection legal statutes.

Federal Congressional Testimony Publicly recognized by NRC Chairman, Ivan Selin, in May 1993 in his comments to U.S. Senate, "It is true.. .everything Mr. Gundersen said was absolutely right; he performed quite a service."

Commended by U.S. Senator John Glenn for public testimony to Senator Glenn's NRC Oversight Committee.

PennCentral Litigation Evaluated NRC license violations and material false statements made by management of this nuclear engineering and materials licensee.

Three Mile Island Litigation Evaluated unmonitored releases to the environment after accident, including containment breach, letdown system and blowout. Proved releases were 15 times higher than government estimate and subsequent government report.

Western Atlas Litigation Evaluated neutron exposure to employees and license violations at this nuclear materials licensee.

Commonwealth Edison In depth review and analysis for Commonwealth Edison to analyze the efficiency and effectiveness of all Commonwealth Edison engineering organizations, which support the operation of all of its nuclear power plants.

Peach Bottom Reactor Litigation Evaluated extended 28-month outage caused by management breakdown and deteriorating condition of plant.

SpecialRemediation Expertise:

Director of Engineering, Vice President of Site Engineering, and the Senior Vice President of Engineering at Nuclear Energy Services (NES) Division of Penn Central Corporation (PCC)

Page 13 of 16

• NES was a nuclear licensee that specialized in dismantlement and remediation of nuclear facilities and nuclear sites. Member of the radiation safety committee for this licensee.

• Department of Energy chose NES to write DOE DecommissioningHandbook because NES had a unique breadth and depth of nuclear engineers and nuclear physicists on staff.

• Personally wrote the "Small Bore Piping" chapter of the DOE's first edition Decommissioning Handbook, personnel on my staff authored other sections, and I reviewed the entire Decommissioning Handbook.

Served on the Connecticut Low Level Radioactive Waste Advisory Committee for 10 years from its inception.

• Managed groups performing analyses on dozens of dismantlement sites to thoroughly remove radioactive material from nuclear plants and their surrounding environment.

• Managed groups assisting in decommissioning the Shippingport nuclear power reactor.

Shippingport was the first large nuclear power plant ever decommissioned. The decommissioning of Shippingport included remediation of the site after decommissioning.

• Managed groups conducting site characterizations (preliminary radiation surveys prior to commencement of removal of radiation) at the radioactively contaminated West Valley site in upstate New York.

Personnel reporting to me assessed dismantlement of the Princeton Avenue Plutonium Lab in New Brunswick, NJ. The lab's dismantlement assessment was stopped when we uncovered extremely toxic and carcinogenic underground radioactive contamination.

Personnel reporting to me worked on decontaminating radioactive thorium at the Cleveland Avenue nuclear licensee in Ohio. The thorium had been used as an alloy in turbine blades.

During that project, previously undetected extremely toxic and carcinogenic radioactive contamination was discovered below ground after an aboveground gamma survey had purported that no residual radiation remained on site.

Additional Education Basic Mediation Certificate Champlain College, Woodbury Institute 28-hour Basic Mediation Training September 2010 Teaching and Academic Administration Experience Rensselaer Polytechnic Institute (RPI) - Advanced Nuclear Reactor Physics Lab Community College of Vermont - Mathematics Professor - 2007 to present Burlington High School Mathematics Teacher - 2001 to June 2008 Physics Teacher - 2004 to 2006 The Marvelwood School - 1996 to 2000 Awarded Teacher of the Year - June 2000 Chairperson: Physics and Math Department Mathematics and Physics Teacher, Faculty Council Member Director of Marvelwood Residential Summer School Director of Residential Life The Forman School & St. Margaret's School - 1993 to 1995 Physics and Mathematics Teacher, Tennis Coach, Residential Living Faculty Member

Page 14 of 16 Nuclear Engineering Work Experience 1970 to Present Expert witness testimony in nuclear litigation and administrative hearings in federal, international, and state court and to Nuclear Regulatory Commission, including but not limited to: Three Mile Island, US Federal Court, US NRC, NRC ASLB & ACRS, Vermont State Legislature, Vermont State Public Service Board, Florida Public Service Board, Czech Senate, Connecticut State Legislature, Western Atlas Nuclear Litigation, U.S. Senate Nuclear Safety Hearings, Peach Bottom Nuclear Power Plant Litigation, and Office of the Inspector General NRC.

Nuclear Engineering, Safety, and Reliability Expert Witness 1990 to Present

• Fairewinds Associates, Inc - Chief Engineer, 2005 to Present

• Arnold Gundersen, Nuclear Safety Consultant and Energy Advisor, 1995 to 2005

• GMA - 1990 to 1995, including expert witness testimony regarding the accident at Three Mile Island.

Nuclear Energy Services, Division of PCC (Fortune 500 company) 1979 to 1990 Corporate Officer and Senior Vice President - Technical Services Responsible for overall performance of the company's Inservice Inspection (ASME XI),

Quality Assurance (SNTC lA), and Staff Augmentation Business Units - up to 300 employees at various nuclear sites.

Senior Vice President of Engineering Responsible for the overall performance of the company's Site Engineering, Boston Design Engineering and Engineered Products Business Units. Integrated the Danbury based, Boston based and site engineering functions to provide products such as fuel racks, nozzle dams, and transfer mechanisms and services such as materials management and procedure development.

Vice President of Engineering Services Responsible for the overall performance of the company's field engineering, operations engineering, and engineered products services. Integrated the Danbury-based and field-based engineering functions to provide numerous products and services required by nuclear utilities, including patents for engineered products.

General Manager of Field Engineering Managed and directed NES' multi-disciplined field engineering staff on location at various nuclear plant sites. Site activities included structural analysis, procedure development, technical specifications and training. Have personally applied for and received one patent.

Director of General Engineering Managed and directed the Danbury based engineering staff. Staff disciplines included structural, nuclear, mechanical and systems engineering. Responsible for assignment of personnel as well as scheduling, cost performance, and technical assessment by staff on assigned projects. This staff provided major engineering support to the company's nuclear waste management, spent fuel storage racks, and engineering consulting programs.

Page 15 of 16 New York State Electric and Gas Corporation (NYSE&G) - 1976 to 1979 Reliability Engineering Supervisor Organized and supervised reliability engineers to upgrade performance levels on seven operating coal units and one that was under construction. Applied analytical techniques and good engineering judgments to improve capacity factors by reducing mean time to repair and by increasing mean time between failures.

Lead Power Systems Engineer Supervised the preparation of proposals, bid evaluation, negotiation and administration of contracts for two 1300 MW NSSS Units including nuclear fuel, and solid-state control rooms. Represented corporation at numerous public forums including TV and radio on sensitive utility issues. Responsible for all nuclear and BOP portions of a PSAR, Environmental Report, and Early Site Review.

Northeast Utilities Service Corporation (NU)- 1972 to 1976 Engineer Nuclear Engineer assigned to Millstone Unit 2 during start-up phase. Lead the high velocity flush and chemical cleaning of condensate and feedwater systems and obtained discharge permit for chemicals. Developed Quality Assurance Category 1 Material, Equipment and Parts List. Modified fuel pool cooling system at Connecticut Yankee, steam generator blowdown system and diesel generator lube oil system for Millstone. Evaluated Technical Specification Change Requests.

Associate Engineer Nuclear Engineer assigned to Montague Units 1 & 2. Interface Engineer with NSSS vendor, performed containment leak rate analysis, assisted in preparation of PSAR and performed radiological health analysis of plant. Performed environmental radiation survey of Connecticut Yankee. Performed chloride intrusion transient analysis for Millstone Unit 1 feedwater system. Prepared Millstone Unit 1 off-gas modification licensing document and Environmental Report Amendments 1 & 2.

Rensselaer Polytechnic Institute (RPI) - 1971 to 1972 Critical Facility Reactor Operator, Instructor Licensed AEC Reactor Operator instructing students and utility reactor operator trainees in start-up through full power operation of a reactor.

Public Service Electric and Gas (PSE&G) - 1970 Assistant Engineer Performed shielding design of radwaste and auxiliary buildings for Newbold Island Units 1

& 2, including development of computer codes.

Page 16 of 16 Media Featured Nuclear Safety and Reliability Expert (1990 to present) for Television, Newspaper, Radio, & Internet - Including, and not limited to:

CNN: JohnKingUSA, CNN News, Earth Matters; DemocracyNow, NECN, WPTZ VT, WTNH, VPTV, WCAX, RT, CTV (Canada), CCTV Burlington, VT, ABC, TBS/Japan, Bloomberg: EnergyNow, KPBS, Japan National Press Club (Tokyo), Italy National Press Club (Rome), The Crusaders, Front Page, Five O'Clock Shadow: Robert Knight, Mark Johnson Show, Steve West Show, Anthony Polina Show, WKVT, WDEV, WVPR, WZBG CT, Seven Days, AP News Service, Houston Chronicle, Christian Science Monitor, Reuters, The Global Post, International Herald, The Guardian, New York Times, Washington Post, LA Times, Miami Herald, St. Petersburg Times, Brattleboro Reformer, Rutland Herald, Times-Argus, Burlington Free Press, Litchfield County Times, The News Times, The New Milford Times, Hartford Current, New London Day, Vermont Daily Briefing, Green Mountain Daily, EcoReview, Huffington Post, DailyKos, Voice of Orange County, AlterNet, Common Dreams, and numerous other national and international blogs Public Service, Cultural, and Community Activities 2009 to Present -Fairewinds Energy Education Corp 501(C)3 non-profit board member 2005 to Present - Public presentations and panel discussions on nuclear safety and reliability at University of Vermont, Vermont Law School, NRC hearings, Town and City Select Boards, Legal Panels, Local Schools, Television, and Radio.

2007-2008 - Created Concept of Solar Panels on Burlington High School; worked with Burlington Electric Department and Burlington Board of Education Technology Committee on Grant for installation of solar collectors for Burlington Electric peak summer use Vermont State Legislature - Public Testimony to Legislative Committees Certified Foster Parent State of Vermont - 2004 to 2007 Mentoring former students - 2000 to present - college application and employment application questions and encouragement Tutoring Refugee Students - 2002 to 2006 - Lost Boys of the Sudan and others from educationally disadvantaged immigrant groups Designed and Taught Special High School Math Course for ESOL Students - 2007 to 2008 NNSN - National Nuclear Safety Network, Founding Advisory Board Member, meetings with and testimony to the Nuclear Regulatory Commission Inspector General (NRC IG)

Berkshire School Parents Association, Co-Founder Berkshire School Annual Appeal, Co-Chair Sunday School Teacher, Christ Church, Roxbury, CT Washington Montessori School Parents Association Member Marriage Encounter National Presenting Team with wife Margaret Provided weekend communication and dialogue workshops weekend retreats/seminars Connecticut Marriage Encounter Administrative Team - 5 years Northeast Utilities Representative Conducting Public Lectures on Nuclear Safety Issues Personal Married to Maggie Gundersen 1979. Two children: Eric, 32, president and founder of MapBox and Development Seed, and Elida, 29, paramedic in Florida. Enjoy sailing, walking, cross-country skiing, yoga, and reading. ---- End

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

)

In the Matter of )

) Docket Nos. 50-361-CAL & 50-362-CAL SOUTHERN CALIFORNIA EDISON COMPANY)

) ASLBPNo. 13-924-01-CAL-BDO1 (San Onofre Nuclear Generating Station, )

Units 2 and 3) ) January 9, 2013 GUNDERSEN AFFIDAVIT Arnold Gundersen , being duly sworn, state:

(Print Name)

1. My name is Arnold Gundersen and I reside at 125 Northshore Drive, Burlington, Vermont.

2. My CV is attached. I have both Bachelor's and Master's degree in nuclear engineering. I was an Atomic Energy Commission Fellow, a Licensed Reactor Operator, and I hold one nuclear plant patent.

3. My pertinent experience related to the Steam Generator matters being considered by this ASLB Proceedings include but are not limited to:

3.1. As the Senior Vice President of Inspection Services, I was responsible for a group of approximately 200 personnel performing ASME III and ASME XI non-destructive piping inspections at nuclear plants throughout the United States. These personnel used inspection techniques identical to those used on the San Onofre tube inspections.

3.2. As the Senior Vice President of Engineering Services, I was responsible for the development of the first ever modern steam generator nozzle dams that were sold to approximately 40 nuclear reactors in the US and Asia. Dams of a similar design are in use in San Onofre's Replacement Steam Generators (RSG).

4. Friends of the Earth (FoE) has retained me to provide my expert opinion on several Factual Issues that this Atomic Safety Licensing Board directed FoE to consider.

Page 2 of 14 Issue #1: Does the Final Safety Analysis Report (FSAR) analyze a steam generator (S/G) tube failure event?

5. Yes, the FSAR does address a steam generator tube failure event.

6. One specific example of where steam generator tube integrity is addressed is in the San Onofre Technical Specifications' that are part of the FSAR.

6.1. Specifically, page 505 of the Technical Specifications has as a Limiting Condition of Operation "Steam Generator tube integrity shall be maintained":

5.5.2.11 Steam Generator (SG) Program (continued)

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

Structural integrity performance criterion: All in- service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a 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.

6.2. A second example where steam generator integrity is addressed on page 510 of the San Onofre Technical Specifications that states that the limiting design basis accident is a "double ended rupture of a single tube":

The steam generator tube rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to ... the leakage rate associated with a double-ended rupture of a single tube.

7. Eight replacement steam generator tubes failed their pressure tests in 2012 and more than 1,000 others have been plugged.

8. Therefore, a review of the evidence makes it clear that the San Onofre Replacement Steam Generator tube damage discovered in 2012 was so severe and extensive that both reactors have been operating in violation of their NRC FSAR license design basis as defined in their Technical Specifications.

1 http://pbadupws.nrc.gov/docs/ML1125/ML11251A100.pdf

Page 3 of 14

9. The Main Steam Line Break with radiological leakage through the steam generator tubes is one of the bounding conditions in emergency plan evaluation and the extent of steam generator tube failures directly impacts the FSAR analysis.

10. The Replacement Steam Generator (RSG) modifications at San Onofre increased both the likelihood of equipment failure and the radiological consequence of such failure and therefore directly affect the FSAR Current Design Basis.

11. In a Pressurized Water Reactor (PWR), the Containment barrier includes the steam generator tube sheet and the steam generator tubes. Edison modified the San Onofre Units 2 and 3 tube sheets by removing the "stay cylinder" from the original Combustion Engineering design and modified the tubes by adding 377 additional tubes to each RSG.

Therefore, by taking this action, Edison chose to modify the San Onofre containment design by installing the radically different Replacement Steam Generators.

12. General Design Criteria 50 of 10 C.F.R. § 50 Appendix A (Containment design basis.)

states: "This margin shall reflect consideration of(l) the effects ofpotential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators.... (2) the limited experience and experimental data availablefor defining accidentphenomena and containmentresponses.... "

13. The rapid and extraordinarily severe wear that resulted in the 2012 failures of all of Edison's San Onofre Replacement Steam Generators was the result of Edison's 2005 decision to radically change the RSG design and to claim that the Part 50.59 licensing process did not apply. These unlicensed unapproved design changes to the containment boundary violated General Design Criteria (GDC) 50 and therefore the FSAR must be amended to reflect Edison's significant modifications.

14. General Design Criteria 16 of 10 C.F.R. § 50 Appendix A (Containment design) states:

"Reactorcontainment and associatedsystems shall be providedto establish an essentially leak-tight barrieragainst the uncontrolledrelease of radioactivity to the environment and to assure that the containment design conditions importantto safety are not exceededfor as long as postulatedaccident conditions require."

15. The degraded condition of the tubes in the RSGs at San Onofre make it clear that Edison had violated GDC 16 and that Edison's modifications to the containment boundary must undergo the rigorous review of a formal FSAR license amendment process including the requisite public hearings.

16. In my opinion, San Onofre's RSG modifications violated both GDC 16 and GDC 50 and created an unanalyzed accident the significance of which was not considered in its Final Safety Analysis Report.

17. In order to determine whether the consequences or severity of accidents analyzed in the Final Safety Analysis Report (FSAR) may be affected by any proposed change activity, the NRC regulations require that plant design changes be implemented through the 10 C.F.R. § 50.59 process. This process is used to evaluate whether any changes to plant design or operation require prior NRC approval.

Page 4 of 14

18. The nuclear industry realized that the FSAR itself might lack sufficient details on proposed changes; therefore, the nuclear trade organization Nuclear Energy Institute (NEI) developed a set of specific guidelines for utilities and energy companies to follow in order to account for deficiencies in the each FSAR. The NRC approved the use of the NEI process.

19. One of the cornerstones to the NEI guidelines is determining if the proposed changes might have an adverse impact on plant safety. Adverse safety consequence is the driving factor for requesting NRC approval of a 50.59 change, not merely the "like-for-like" changes claimed by Edison.

20. While the NRC Augmented Inspection Team (AIT) briefly described how Edison addressed its 50.59 requirements, the evidence shows that Edison did not comply with the NEI guidelines for implementing 50.59.

21. Published reports indicate that the strategic decision made by Edison that the 50.59 process would not be applied to the RSGs was made by corporate officials before any engineering personnel had actually performed the 50.59 engineering analysis.

Consequently, Edison made a management decision to claim that the 50.59 process did not apply and therefore San Onofre was not required to seek NRC approval for the proposed changes at San Onofre Units 2 and 3. The Edison decision to ignore the 50.59 process for San Onofre's steam generators, enabled to avoid modification of its FSAR commitments as well as avoid analysis on steam generator performance and accidents

22. Proper operation of a steam generator is a major safety issue for each PWR. In addition to providing the containment barrier to radioactivity and producing steam, the steam generator has many other important safety functions. Therefore any RSG design changes clearly have potential safety consequences that are acknowledged in the FSAR.

Consequently, any design and/or fabrication change made to the steam generator must be thoroughly evaluated for its safety implications.

22.1. The RSG is the major component in the plant that contributes to safety during transients and accidents.

22.2. The RSG 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.

23. The NRC has acknowledged the fact that Edison employed a new methodology not reviewed or recognized in the FSAR to calculate the heat transfer, velocities, levels and water/steam distribution on the secondary side of both the Unit 2 and Unit 3 Replacement Steam Generators. And to date, the NRC has released no findings regarding the full impact of Edison's unreviewed and undocumented changes to its FSAR as a result of such radical design and fabrication changes to San Onofre's RSGs.

24. The overall performance of the Original Steam Generators was based upon a one dimensional computer code known as CRIB described in the FSAR, while the design and performance of the RSGs was based upon an unreviewed and un-benchmarked three dimensional code known as FIT-IlI which is not described in the FSAR.

Page 5 of 14

25. Knowing the standards applied and benchmarked for the RSG computer codes CRIB and FIT-III is critical information in the FSAR because the RSG computer code determines the thermal hydraulic performance during normal and accident conditions.

26. The AIT report indicated that the change to the FIT-Ill evaluation methodology was not discussed as part of Edison's 50.59 screening because the details of thermal hydraulic models used for the design of the OSG were not discussed in the original FSAR.

27. It should have been obvious to Edison that FIT-Ill has not been benchmarked and had not been previously used in licensing procedures showing that the use of FIT-Ill might have an adverse effect on the FSAR safety analysis thus necessitating the entire license amendment review and public hearing process.

28. As noted by the AIT, Edison approved the use of FIT-III code even though the code was not benchmarked nor identified as acceptable in the FSAR. Consequently, Edison operated San Onofre without knowing the uncertainties in the Replacement Steam Generators' performance characteristics. Predicted liquid levels, pressure drops, vibrations, and temperatures at both Units 2 and 3 were all subject to unknown uncertainties during both normal and abnormal operations.

29. In my opinion, by approving the use of an un-benchmarked and untested design tool like FIT-Ill, Edison did not did not meet the requirements expected from a nuclear licensee.

Use of an un-benchmarked computer code that is not included in the FSAR protocol demands a formal FSAR license amendment process including the requisite public hearings.

30. The AIT makes no reference to a NRC review or lack of review of the requisite 50.59 screening evaluation or whether the NEI criteria involving safety significance were included in Edison's analysis.

31. Design changes of the magnitude created by Edison to the San Onofre RSGs should have triggered a Request for Additional Information from the NRC. No RAI was issued by the NRC, because Edison never notified the NRC of the significant modifications its San Onofre operating license.

32. The AIT reported that FIT-Ill predictions differed considerably in comparison to an Electric Power Research Institute developed code named ATHOS. FIT-1Il predicted lower flow velocities and void fractions that were not conservative compared to ATHOS.

The AIT Report neglected an analysis of the root cause of the critical differences between FIT-III and ATHOS, and the negative impact such lax calculational modeling had on the design, fabrication, and successful operation of the San Onofre RSGs. Had Edison sought the required FSAR license amendment, comparisons between FIT-Ill and ATHOS would have been identified six years ago.

33. The AIT did not address the possibility that the lack of conservatism in FIT-Ill predictions, in addition to causing tube vibrations, could also result in non-conservative predictions of the behavior of the steam generator pressure vessel and associated main steam piping during accident conditions that are required to be analyzed in the FSAR.

Page 6 of 14

34. The AIT noted that the non-conservatisms in FIT-III are a contributor to the failure by Edison to adequately calculate the San Onofre RSG tube vibrations.

34.1. But equally important, the AIT failed to address that FIT-Ill could also create non-conservative predictions of the behavior of the steam generator pressure vessel and associated main steam piping during accident conditions that are required to be analyzed in the FSAR.

34.2. Such a conclusion implies that damage to the steam generator pressure vessel itself, and not just the tubes, might have occurred at San Onofre and remains unanalyzed by either Edison or the NRC.

35. The probability of an accident exceeding the plant's Current Design Basis is increased by the radically different Edison Replacement Steam Generators.

35.1. For example, uncertainties in predicting the thermal hydraulic performance of the steam generator nozzle may lead to stratification and early fatigue failures in the steam generator itself or associated main steam piping.

35.2. Hence, the operational risks involved in operating the San Onofre RSGs have created a licensing condition that should have been addressed as part of an FSAR license amendment and hearing process.

36. It is my professional opinion that Edison should have applied for the 50.59 process so that the FSAR license amendment evaluation and public hearings would have occurred six years ago, prior to creating an accident scenario and facing losses that by the end of this process will easily total more than $1 Billion.

37. The seriousness of the licensing and safety impact of the damaged RSGs at San Onofre cannot be overstated or underestimated.

37.1. Any Design Basis Accident (DBA) as defined in the FSAR needs to be accurately modeled in order to protect public health and safety.

37.2. The FSAR's DBA analysis including the extent of tube leakage in the event of a Main Steam Line Break significantly impacts the design and implementation of Emergency Evacuation Plans.

38. In the event of a steam line break accident in the San Onofre Replacement Steam Generators with the degraded condition of the tubes, an accident would have occurred that is more severe than any design basis accident scenario previously analyzed by Edison in the FSAR.

39. Such a DBA steam line break accident would render the San Onofre emergency plan totally inadequate and most likely cause an evacuation of a large portion of Southern California.

Page 7 of 14

40. Edison dramatically increased the radiation risk to the public as a result of San Onofre with Replacement Steam Generators that were extremely flawed beginning with their original design. The fact that 8 tubes failed the pressure tests in Unit 3 indicates that those tubes would have failed during a main steam line break (MSLB).

41. It is uncertain if a reactor operator would have been able to shut the plant down without melting the core. A simultaneous rupture of 8 tubes would have caused a primary to secondary leak of radioactive coolant of about 5000-6000 gallons per minute. This leakage would have begun to drain the nuclear core as well as releasing radioactive primary coolant to the atmosphere.

42. The ability of a reactor operator to control the water level in the affected steam generator with this high leakage rate and keep the nuclear reactor core cooled has never been analyzed or tested. An accident of this magnitude is outside ANY reactor's Current Design Basis (CDB).

43. The evidence presented by Edison and the NRC AIT shows that the real reason San Onofre had to plug 1300 tubes (and not just the eight that failed the pressure test) was that the San Onofre units were operating outside their Current Design Basis as defined in the FSAR and were in an unanalyzed, unlicensed condition.

44. Not only have Edison's modifications to the RSGs increased the severity of an accident, but also the Replacement Steam Generator modifications have increased the likelihood of a main steam line break. Even the NRC's AIT concluded that the probability of a MSLB was double what it had been with the OSG's.

45. In my opinion, thermal stratification and changes in the outlet steam flow from the Replacement Steam Generators would have induced stresses in the main steam piping that would likely increase the probability of a MSLB even beyond the NRC's conclusion.

46. Therefore, both the probability and the consequences of an accident have increased beyond those in the FSAR and the plant's Current Design Basis as a result of Edison's replacement team generator modifications.

47. The evidence clearly shows that Edison has been operating outside the design basis of its Final Safety Analysis

48. The modifications to the Replacement Steam Generators at San Onofre and the fact that eight tubes failed critical pressure tests significantly raises the potential for radiation bypassing the containment during severe accidents such as a main steam line break accident (MSLB), station blackout (SBO) and anticipated transients without scram (ATWS) events. This situation violates General Design Criteria 16 and 50 and thus triggers the commencement of a formal FSAR license amendment process including the requisite public hearings.

Page 8 of 14 ISSUE # 2 Figure 4-3 in the report entitled "Operational Envelope for Large U-bend Steam Generators, SONGS U2C17 Steam Generator Operational Assessment for Tube-to-Tube Wear" [hereinafter Tube-to-Tube Report] compares the bulk velocity ratio and void fraction ratio to several successfully operating large S/Gs, and it notes that "[alt 100% power, the thermal-hydraulic conditions in the u-bend region of the SONGS replacement [S/Gs] exceed the past successful operational envelope for U-bend nuclear [S/Gs] based on presently available data." Tube-to- Tube Report at 17.

How similar to the SONGS S/Gs are these other S/Gs? Do the other steam generators, for example, use alloy 670 tubes and have similar spacing, similar support structures, etc.?

49. The Combustion Engineering (CE) designed original steam generators (OSG) are not at all similar to the Mitsubishi RSGs, nor are the Mitsubishi RSGs similar to any other steam generators with which Edison is attempting to make a comparison.

49.1. No other Replacement Steam Generator design in the country has been modified in the extreme manner that those at San Onofre Units 2 and 3 have been altered.

49.2. Combustion Engineering built the OSG's at San Onofre. Because CE used only two steam generators, these OSG's were very large and had a tight tube pitch. To assure proper water flow the OSG's had egg crate tube support plates with a region at the center with no tubes and no heat load where a "stay cylinder" was located.

49.3. Mitsubishi Heavy Industry, the fabricator of the Replacement Steam Generators (RSG), is a Westinghouse licensee and is not prepared to manufacture the tight tube pitch and the egg crate tube supports of the San Onofre RSG design.

50. Edison instructed Mitsubishi to replace the OSG egg crate design with broached tubes and to remove the OSG stay cylinder to add additional tubes to an area where there formally was no heat load. Edison also instructed Mitsubishi to add many other modifications to the RSG that are simply too numerous to list in this affidavit.

51. To the best of my knowledge and belief, no other steam generator in the nation is as large as those at San Onofre with broached tube supports, a tight Combustion Engineering tube pitch, and no stay cylinder. Therefore, comparing San Onofre to "several other successfully operating large S/G's" is simply not a valid engineering or scientific comparison.

52. My professional experience shows that the actual root cause of the steam generator tube degradation is the 2005 strategic decision by Edison to remove the stay cylinder, change the tube sheet, change the tube support structures and add an additional 400 tubes in the Replacement Steam Generator design while still claiming that this significant design modification was a "like-for-like" replacement. These changes have created Replacement Steam Generators unlike any other in the nation.

53. Adding almost 400 additional tubes to the central location where the stay cylinder had been previously located increased the heat load where it was already the highest.

Page 9 of 14

54. At the same time, Edison removed the egg crate tube supports and replaced them with broached tube supports that reduced cooling flow.

55. These three changes (additional tubes, removal of stay cylinder and egg crate removal) caused a unique and unanalyzed heat load to the interior of the Replacement Steam Generators that will continue to cause the tubes to vibrate and fail even after some have been plugged.

56. The center section of the original San Onofre steam generators contained a key structural element called a "stay cylinder" and no steam generator tubes. In 2005 or early 2006, Edison made a management decision to eliminate this vital support pillar and add additional tubes in its place.

57. In the original steam generator design, there was no heat input in this central area of the steam generator, because there were no tubes to add the heat. When Edison added almost 400 tubes (4% of the tubes) to the center of the tube bundle in the San Onofre Replacement Steam Generators, Edison effectively increased the power distribution to the center of the steam generator.

58. This radical and unanalyzed design change moved 4% of the heat to the inside of the tube bundle while reducing the heat by 4% to the outside of the tube bundle.

59. Adding this heat to the center of the bundle was then exacerbated by removing the egg crate tube supports and replacing them with a broached tube support plate design that further reduced flow to the center of the steam generator.

60. As the NRC confirmed in its AIT report, a large steam void has developed near where the additional tubes were added in the Replacement Steam Generators (called fluid elastic instability) that allows many types of excess vibrations to occur.

61. Fairewinds review of Figure 1 below from Edison's Condition Report clearly shows that the location within the steam generators where the steam "fluid elastic instability" has developed is precisely the region where the extra heat created by the 400 new tubes would create an excess of steam and various vibrational modes.

62. While 4% may seem like a small change, it is not. Each San Onofre reactor generates a total thermal output of approximately 3400 megawatts of heat. If one mathematically converts 4% of 3400 megawatts of heat, it equals 135 megawatts, or to illustrate it differently: 180,000 horsepower of thermal heat that was transferred from the outside of the tube bundles to the center.

Page 10 of 14 2

63. Contour Of Steam Quality Figure 1 0.10

(. It 0,111

0. 22 0.26 0.14 0, i39 0.13 0 VT1 0.5b

0. 6')

64. This data shows that a significant quantity of additional heat has been transferred to an area that previously had no tubes. That heat must be removed from this central area, yet Edison also reduced flow by replacing the egg crate supports with broached supports.

65. These design changes by Edison created too many steam bubbles that are causing various vibration modes and degradation in all four steam generators.

66. The data reviewed shows that the decision by Edison, to add almost 400 tubes to the center of the four Replacement Steam Generators, changes flow patterns by removing the stay cylinder. This decision by Edison also reduces flow by removing the egg crate tube supports and created the excess heat that is the causative factor in the fluid elastic instability in the Replacement Steam Generators at San Onofre.

67. No other steam generator in the United States was ever modified in a similar fashion and therefore comparisons to other steam generators at other reactors is not relevant or applicable.

2 Condition Report: 201836127,Revision 0, 5/7/2012, Figure 2: Contour of steam quality at the height of the maximum quality in U-bend region for Thot = 598"F (Figure 8.1-2 (a) in Reference [2]), Page 74.

Page 11 of 14 Issue #3: Figure 5-1 in the Tube-to-Tube Report compares the same parameters as in Figure 4-3, but for operation at 70% power. It appears from Figure 5-1 that the bulk fluid velocity for SONGS is at the high end of the experiential range. Given the likely differences between the SONGS generators and those cited in the discussion, can one conclude that operation at 70% power is conservative?

68. The request by Edison to operate San Onofre Unit 2 at 70% power is not a conservative decision.

69. To focus on Fluid Elastic Instabilities and tube-to-tube interactions is to miss the significant problems with the defective San Onofre Replacement Steam Generators.

70. Fluid Elastic Instability (FEI) causes the tubes to vibrate abruptly at large amplitudes, so it would be imperative that the velocity is maintained below the critical values that create dynamic instabilities. Both the NRC's AIT and Edison's Cause Report neglect the criticality of accurate predictions in the relationship between power and local velocities would be required to restart Unit 2.

71. However, Vortex Induced Vibrations (VIV) and Turbulence Induced Vibration (TIV) might be created if San Onofre Unit 2 were allowed to operate at reduced power, and once again, the NRC and Edison have neglected to review and acknowledge these scenarios.

72. Significant tube damage from fatigue and wear during relatively long periods of operation can cause FEI, VIV, and TIV. Therefore the restart of San Onofre Unit 2 should not be considered because Edison and the NRC reviewed and addressed these issues in their pro-forma reviews.

73. Additionally, properly scaled physical mockups of the San Onofre Replacement Steam Generators, not inadequate computer simulations, are needed and must be required to accurately assess tube wear and vibrational risk created by the possible operation of Unit 2.

74. Computer codes cannot operate and be assessed with out a full-scale mockup prepared by which to provide benchmarks for the computer codes. Once a complete assessment of full-scale mockups is completed, then the computer codes should have the capability to predict local heat transfer rates, pressure drops, void fraction, and velocities.

75. Focusing on measuring and plugging tubes that have become thinner as a result of internal vibrations does not verify San Onofre's RSGs. Edison is attempting to avoid the serious and necessary scientific analysis that would determine which unplugged tubes have become cracked from vibrations and yet are not deemed thin enough to require plugging.

76. Thus, prior to considering the restart of San Onofre Unit 2 at reduced power, Edison and the NRC must also prove to the public that the undetected cracks, which may have been already produced, will not suddenly fail during an unanticipated swing in reactor

Page 12 of 14 conditions (called an operational transient in the nuclear industry) and/or a design basis accident (DBA) that the plant must be built to withstand.

77. Restart of San Onofre Unit 2 should not be considered unless both Edison and the NRC are able to clearly demonstrate that the relationship between plant power and tube vibration is well understood and that FEI, VIV, or TIV will not add to tube wear and create additional safety risks.

Issue #4: Section 8.0 in the Tube-to-Tube Report states that "Itihe desired margin is a projected maximum stability ratio of 0.75 with 0.95 probability at 50% confidence over the next inspection interval of 5 months." Tube-to-Tube Report at 104. Does a confidence level of 50% meet the reasonable assurance requirement in the regulations?

78. In my opinion, a confidence level of 50% does not provide reasonable assurance of anything related to nuclear safety.

Issue #5: Throughout the Tube-to-Tube Report, the term "operational assessment" is used. How is the term "operational assessment" different than or the same as the terms "test" and "experiment" used in 10 C.F.R. § 50.59?

79. Operating the damaged San Onofre Unit 2 at reduced power is an experiment by Edison on steam generators that are unlike any other steam generators that have been designed and fabricated anywhere in the world. The term "operational assessment" is a euphemism employed by Edison to avoid meeting its regulatory requirements.

80. Edison has already acknowledged to the NRC that a research experiment, not an "operational assessment", will be performed and at San Onofre Unit 2 during its proposed five-month period of reduced power operation.

81. Unfortunately, the official transcript of the December 18 meeting between the NRC and Edison is not yet publically available, but Michael Blood of the Associated Press quotes Edison consultant Mike Short as saying research will be performed on tube vibrations when the plant operates at 70% power. Specifically, according to AP: "Short said the data collected by the system could be used in future research examining vibrations picked up by the monitors." 3

82. I note that this pattern of avoiding the intent of the NRC's regulation by relying on euphemism and carefully parsing words is a persistent mode of operation by Edison dating back to its earliest licensing decision to knowingly avoid the rigorous 50.59 process for the Replacement Steam Generators at San Onofre.

3 San Onofre: Edison backpedals on claim that retooling will aid safety, Associated Press, December 18, 2012, http://www.ocregister.com/news/plant-381083-edison-unit.html

Page 13 of 14 List of documents 4 used to conduct my analysis and arrive at my opinions:

1. San Onofre Technical Specifications http://pbadupws.nrc.gov/docs/ML1125/ML11251A100.pdf

2. General Design Criteria 50 of 10 C.F.R. § 50 Appendix A

3. General Design Criteria 16 of 10 C.F.R. § 50 Appendix A

4. Nuclear Energy Institute (NEI) 50.59 guidelines http://pbadupws.nrc.gov/docs/ML0037/ML003771157.pdf

5. Edison Management Strategic Decision Not To Implement 50.59: Improving Like-For-Like Replacement Steam Generators by Boguslaw Olech of Southern California Edison and Tomouki Inoue of Mitsubishi Heavy Industries, Nuclear Engineering International, January 2012, page 36-38. http://edition.pagesuite-professional.co.uk/launch.aspx?referral=other&pnum=36&refresh-=KOs3 a21 GRQ61`%20

&EID=af75ecbl-5b23-49be-9dd6-d806f2e9b7b5&skip=&p=36

6. NRC SAN ONOFRE REPLACEMENT STEAM GENERATOR AIT REPORT:

http://pbadupws.nrc.gov/docs/ML1218/ML12188A748.pdf

7. STEAM GENERATOR FAILURES AT SAN ONOFRE: THE NEED FOR A THOROUGH ROOT CAUSE ANALYSIS REQUIRES NO EARLY RESTART, Fairewinds Associates, Monday, Mar 26, 2012:

http://www.fairewinds.com/content/steam-generator-failures-san-onofre

8. SAN ONOFRE CASCADING STEAM GENERATOR FAILURES CREATED BY EDISON: IMPRUDENT DESIGN AND FABRICATION DECISIONS CAUSED LEAKS, Fairewinds Associates, Monday, Apr 9, 2012 http ://www.fairewinds.com/content/san-onofre-cascading-steam-generator-failures-created-edison

9. SAN ONOFRE'S STEAM GENERATOR FAILURES COULD HAVE BEEN PREVENTED, Fairewinds Associates Monday, May 14, 2012 http://www. fairewinds.com/content/san-onofre's-steam-pgenerator-failures-could-have-been-prevented

10. SAN ONOFRE'S STEAM GENERATORS: SIGNIFICANTLY WORSE THAN ALL OTHERS NATIONWIDE, Fairewinds Associates, Tuesday, Jul 10, 2012 http://www .fairewinds.com/content/san-onofre' s-steam-generators-significantly-worse-all-others-nationwide 4 No documents were provided by Edison, and no documents are covered in any confidentiality agreement between the parties.

THE UPS STORE BURLINGTON Q001 01/10/2013 15:28 FAX 8026511699 Page 14 of 14

11. Steam Generator Steam Quality Graph copied from Edison Condition Report:

201836127, Revision 0, 5/7/2012, Figure 2: Contour of steam quality at the height of the maximum quality in U-bend region for Thot = 598"F (Figure 8.1-2 (a) in Reference [2D, Page 74.

Arnold Gundersen, MSNE, RO Chief Engineer, Fairewinds Associates, Inc Arnold Gundersen (Name)

Subscribed to and sworn before me this.J _day of -- an pv*qz- . O Notary Public:

My commission expires: 6X '/1-0*/..... j&..

Attachment 2 Steam GeneratorFailuresat San Onofre: The Needfor a Thorough Root Cause Analysis Requires No Early Restart Fairewinds Associates March 26, 2012

op I U I U A

I XL'

FAIREWINDS ASSOCIATES STEAM GENERATOR FAILURE AT SAN ONOFRE EXECUTIVE

SUMMARY

Following 28 and 29 years of operation, the two San Onofre Nuclear Generating Station reactors owned by Southern California Edison (Edison) are unable to safely generate the necessary electricity for the people of California.

An investigation conducted by Fairewinds Associates has identified that a series of major modifications to the internal design of replacement steam generators in both San Onofre Units 2 and 3 are likely the cause of excessive wear, leaks and pressure test failures in the steam generator tubes. Despite Edison's rush to make an early restart of at least Unit 2 if not Unit 3, and the apparent relaxed approach of the NRC as to their role in the timing of any start up by Edison, Fairewinds Associates recommends that both San Onofre Unit 2 and Unit 3 remain shut down until the "root causes" of the nuclear power plant's rapid tube failures are understood and repaired, reliability is assured, and radioactive releases are prevented.

Page 1 I Fairewinds Associates I Steam Generator Failure at San Onofre

FAIREWINDS ASSOCIATES STEAM GENERATOR FAILURE AT SAN ONOFRE BACKGROUND There are 104 nuclear power plants generating electricity in the United States (US).

Pressurized Water Reactors (PWR's), like San Onofre, account for almost 70% of all US reactors; the remaining 30% of nuclear power reactors are Boiling Water Reactors (BWR's).

The San Onofre nuclear power reactor is a very unique design originally built by Combustion Engineering (CE) and is very different from the Westinghouse or Babcock & Wilcox nuclear power reactor designs. While most of the Westinghouse U-Tube PWR designs have three or four steam generators, all of the CE nuclear reactors use only two steam generators.

Because there are only two steam generators in this Combustion Engineering design, each steam generator is 50% larger than those built by Westinghouse for a similar reactor power output. In fact, only the 14 CE PWR nuclear reactors, out of the 104 nuclear reactors in the US, have this very unique and extra large sized steam generator system. This means that the replacement steam generators at San Onofre are some of the largest steam generators that have ever been designed or manufactured.

Unlike a Boiling Water Reactor (BWR), the water that cools the nuclear core inside a PWR never boils. In order to prevent boiling the cooling water is pressurized to more than 2,000 pounds per square inch (psi). However, in order to make a turbine spin and generate electricity, the nuclear power plant must produce steam. In a PWR, a steam generator is used to transfer heat from the pressurized, radioactive water that cools the reactor to the steam that turns the turbine and is supposed to be non-radioactive. To accomplish this engineering feat, the hot pressurized reactor water is pushed through thousands of U-shaped tubes inside the steam generator in order to remove the heat and by that process create non-radioactive steam on the outside of those same tubes to spin the turbine and generate electricity.

Page 2 Fairewinds Associates I Steam Generator Failure at San Onofre

FAIREWINDS ASSOCIATES STEAM GENERATOR FAILURE AT SAN ONOFRE STATUS The San Onofre reactors have significant problems because their newly installed steam generators have extensive degradation and are unable to perform their design function of containing the radioactive water in the facility. Concerned about the safety of these plants, Senator Barbara Boxer (D-CA), Chair of the Environment and Public Works Committee, sent a letter February 8, 2012, to the Chairman of the Nuclear Regulatory Commission (NRC), Dr. Gregory Jaczko, requesting that the NRC review and report on the safety conditions at the San Onofre nuclear plant due to the recently discovered problems related to tubes that carry radioactive water at the facility. Senator Boxer asked the NRC to assess the conditions at the plant, which is located in San Clemente, California, to determine if further action is needed.

Only one month later on March 13, 2012, Chairman Jaczko responded, "The root cause of the tube leak has not yet been determined. ....NRC approval is not required for the licensee [Edison] to restart Units 2 and 3."

Steam generator tube degradation, like that which San Onofre is experiencing, causes a significant nuclear safety risk by substantially increasing the likelihood of an accident that releases radioactivity into the environment. Unfortunately, a leak or disintegration of one or more tubes would cause the radioactive water to escape the containment.

Because there is a 1,000-pound-per-square-inch (psi) pressure difference between the high-pressure radioactive side of the tubes and the lower pressure steam that then leaves the containment, a leak will inevitably release radioactivity into the environment.

Gross failure of one or more of the steam generator tubes could create a nuclear design basis accident and cause the nuclear reactor core to lose a portion of its cooling water.

However, the unique concern of degraded steam generator tubes is that uncontrolled radiation releases from a tube break do not remain inside the containment building and instead leak out of the facility and into public areas via atmospheric dump valves and steam generator blowdown.

Page 3 I Fairewinds Associates I Steam Generator Failure at San Onofre

FAIREWINDS ASSOCIATES STEAM GENERATOR FAILURE AT SAN ONOFRE DESIGN CHANGES CREATE SIGNIFICANT RISK OF FAILURE The design engineers for the San Onofre reactors believed that steam generator tubes would last for the lifetime of the nuclear reactor without any appreciable leakage. Because steam generators that hold those tubes were considered permanent components and would never need replacement, the nuclear containment buildings were never provided with access doors large enough to allow for the removal of degraded steam generators.

Even though tubes were expected to last for the entire lifetime of each PWR, this has not proven to be the case. For example, the first steam generators were replaced at the Surry 2 reactor in Virginia in 1979 after only seven years of operation. The vast majority of PWR steam generators have required replacement due to significant degradation.

As a result of tube deterioration and degradation uncovered several years ago, Southern California Edison (Edison) decided to replace each of the steam generators at both of the San Onofre reactors. A review of the published literature shows that the specifications of four steam generators are identical and they were purchased together under a single contract. Edison signed a contract for these steam generators with Mitsubishi Heavy Industries. This challenging construction project required that a hole be cut in the side of the nuclear containment to remove and replace the old steam generators. It was akin to cracking someone's chest and performing a heart transplant.

It now appears that after new steam generators were installed at San Onofre Unit 2 and Unit 3, the new tubes began to seriously degrade very quickly. Technicians first detected the unanticipated problems of significant wear in the tubes during the Unit 2 refuelling outage in January 2012. The wear-rate for these steam generator tubes is extraordinary because tube thickness has been reduced by as much as 30% in less than two years. With their typical lack of transparency, Southern California Edison, San Onofre, and the NRC were not forthcoming to the public on the extent of the significant degradation in Unit 2's steam generator tubes. While Unit 2 was shutdown for refuelling, San Onofre Unit 3 was operating at full power when it experienced a complete perforation of one steam generator tube that allowed highly radioactive water from inside the reactor to mix with the non-radioactive water that turns the turbine. As a consequence, an uncontrolled release of radiation into the environment ensued, and San Onofre Unit 3 was also forced to shutdown due to steam generator failure.

Page 4 I Fairewinds Associates I Steam Generator Failure at San Onofre

FAIREWINDS ASSOCtATES STEAM GENERATOR FAILURE AT SAN ONOFRE ROOT CAUSE The tube failure and ensuing radiation release at San Onofre Unit 3 made the public keenly aware that both San Onofre Units 2 and 3 are experiencing significant degradation and malfunction of their new steam generators. The public, including Senator Barbara Boxer, has demanded to know the extent of safety ramifications for the San Onofre Unit 2 and 3 steam generators and their leaking tubes.

What has changed that caused the leak?

What can the standard engineering practice of root cause analysis determine as the cause of such severe short-term steam generator degradation?

The most obvious change is that the old steam generators, that operated for more than 25-years, were recently replaced with new ones built by Mitsubishi Heavy Industries.

Why did the original design last for 25-years while the new design failed in only two years?

What did Edison and Mitsubishi modify in the new design that was different from the original design?

Page 5 Fairewinds Associates I Steam Generator Failure at San Onofre

FAIREWINDS ASSOCIATES STEAM GENERATOR FAILURE AT SAN ONOFRE A report1 by Southern California Edison and t Steam outlet Mitsubishi published in January 2012 describes in great detail numerous changes to the original steam generator design.

Fairewinds review of the Edison/MHI report determined that the four most critical changes likely to be a cause of the current tube leaks at San Onofre 2 and 3 are:

,Tube

1. The tube alloy was changed, Plate

2. Reactor flow rate was changed,

3. More steam generator tubes were added, and

4. Modifications were made to the

egg crate" that holds the tubes separate in Unit 2 and Unit 3.

Primary Primary WMle 0"~e

1. Improving Like-For-Like Replacement Steam Generators by Boguslaw Olech of Southern California Edison and Tomouki Inoue of Mitsubishi Heavy Industries, Nuclear Engineering International, January 2012, page 36-38. http://edition.pagesuite-professional.

co.uk/launch.aspx?referral=other&pnum=36&refresh=KOs3a2 1GRq6I &EID=af75ecb 1-5b23-49be-9dd6-d806f2e9b7b5&skip=&p=36

2. http://products.asminternational.org/fach/data/fullDisplay.do?database=faco&record=843&trim=false Page 6 1 Fairewinds Associates I Steam Generator Failure at San Onofre

FAIREWINDS ASSOCIATES STEAM GENERATOR FAILURE AT SAN ONOFRE While each of these changes is significant if reviewed individually, taken together they created a large risk of tube failure at the San Onofre reactors. The significant increase in the number of Mitsubishi steam generator tubes and the large flow rate of radioactive water through these tubes were impacted by this simultaneous change and combination of untested materials and techniques. Fairewinds believes that vibration within the tubes in both Unit 2 and Unit 3 were due to the simultaneous implementation of untested manufacturing and design changes made by the Edison/MHI to the replacement steam generators.

While Edison and San Onofre consider these steam generator replacements at San Onofre as a like-for-like replacement, such a distinction is actually part of a procedure that San Onofre developed in order to avoid the requisite NRC oversight of a steam generator replacement process. Several years prior to the design and installation of the new San Onofre steam generators, Edison/San Onofre completed the 10CFR50.59 review process of replacement steam generators. The Edison/San Onofre 10CFR50.59 review process of the replacement steam generators enabled Edison/San Onofre to have a so-called pre-review, so that the design and manufacture of the replacement steam generators at San Onofre did not receive any actual NRC oversight or technical review. The San Onofre application of the 10CFR50.59 review portrayed the steam generator replacement project as a like-for-like replacement1 that therefore would not require a thorough NRC review and approval process.

As a result of the design and manufacturing changes implemented by Edison and Mitsubishi to the original San Onofre steam generator tubes and related components, both Units 2 and 3 have experienced extraordinarily rapid degradation of their steam generator tubes. Fairewinds believes that if the original steam generators had been replaced with duplicates (like-for-like) as regulators allow, the problems that San Onofre is currently experiencing would have been dramatically reduced or entirely eliminated. The extensive changes made by Edison/Mitsubishi to the new San Onofre steam generators are hardly a like-for-like change and are the likely cause of problems in both Unit 2 and Unit 3.

1. Improving Like-For-Like Replacement Steam Generators by Boguslaw Olech of Southern California Edison and Tomouki Inoue of Mitsubishi Heavy Industries, Nuclear Engineering International, January 2012, page 36-

38. http://edition.pagesuite-professional.co.uk/launch.aspx?referral=other&pnum=36&refresh=KOs3a21GRq61

&EID=af75ecb 1-5b23-49be-9dd6-d806f2e9b7b5&skip=&p=36 Page 7 I Fairewinds Associates I Steam Generator Failure at San Onofre

FAIREWINDS ASSOCIATES STEAM GENERATOR FAILURE AT SAN ONOFRE URGENT EVALUATION REQUIRED Unfortunately, progress on evaluating the extent of the problems at both San Onofre Units 2 and 3 has been slow, due in part to Edison's purchase of only one set of the steam generator nozzle dams required for tube inspections. Steam generator nozzle dams prevent reactor water from leaking into the bottom of the steam generators when inspections are taking place. Consequently, when only one set of dams is available, both units cannot be simultaneously inspected. The decision to procure only one set of nozzle dams indicates a penny wise and pound foolish procurement policy that has made it technically impossible for Edison to simultaneously conduct these critical steam generator examinations of both San Onofre Units 2 and 3 without draining both vessels below their nozzles and limiting the movement of nuclear fuel in each reactor.

Simple inspections, conducted by using Eddy Current tests, indicate that more than 300 tubes in both units show unacceptable wear rates that require further evaluation. (Eddy Current inspections are non-intrusive and send electrical signals through the pipe wall to get a rough approximation of wall wear and thickness.) Now these 300 damaged tubes must be removed from service by welding them closed (plugged) prior to resuming plant operation. Additionally, within this very small sample, Edison has pressure tested only a limited number of tubes in San Onofre 3 and apparently only tested a single tube in Unit 2. This is one tube out of more than 19,000. Moreover, at least eight tubes in San Onofre Unit 3 have completely failed the limited pressure tests performed by Edison, while many thousands remained untested. A pressure test is a destructive test designed to cause a degraded tube that is too weak to sustain necessary pressure to fail under the testing and repair scenario rather than during operation. Such testing weeds out defective tubes on the edge of failure.

Even though Unit 3 experienced a gross tube leak and Unit 2 did not, it is important to note that these inspections showed that more tubes in Unit 2 were degraded than in Unit 3. It is also important to note that both Unit 2 and Unit 3 were designed and manufactured to the same specifications. Since Unit 2 operated somewhat longer than Unit 3, it is not surprising that Unit 2 should exhibit more degradation than Unit 3 as well.

Page 8 I Fairewinds Associates I Steam Generator Failure at San Onofre

FAIREWINDS ASSOCIATES STEAM GENERATOR FAILURE AT SAN ONOFRE INADEQUATE INFORMATION AND TRANSPARENCY As is typical of the NRC and the nuclear industry, they have not been forthcoming to the public or its elected representatives with important details concerning where the leaking and degraded tubes at San Onofre have been detected. San Onofre engineers should have precise maps detailing the degraded and leaking tubes as well as the exact location of the leak(s) in each tube. Such data is just one piece of critical information required in conducting a thorough root cause analysis of the problem and determining an accurate solution.

By failing to pressure test a significant sample of tubes in San Onofre Unit 2, Edison will be unable to determine the full extent of this formidable safety issue. In response, Edison has attempted to focus political and media attention on Unit 3, while trying to obscure the reality, which is that Unit 2 has the same overarching problems as Unit 3. San Onofre Unit 2 has the same steam generators, the same significant wear in the tubes, and the same ongoing failed operational issues as reactor Unit 3. Therefore, Fairewinds believes that in order to prevent radiation releases and assure ongoing long-term reliability, Edison must keep San Onofre Unit 2 shutdown until thorough and systematic tube pressure tests and a root cause analysis have been completed.

Furthermore, a complete chemical analysis of a selection of individual tubes in each of the San Onofre reactors, conducted either by Southern California Edison or an independent outside team of consultants, is the only accurate engineering method available to ascertain if the tube failures are due to metallurgical problems or mechanical wear. If Edison is to accurately determine whether the problems at San Onofre are due to metallurgical insufficiencies or mechanical wear, orthodox engineering methodology requires that San Onofre technicians physically remove (pull) a selection of tubes and examine them. The U shape and long size of these tubes preclude replacement or repair, and therefore the hole from which they are removed must be plugged by welding the hole Page 9 I Fairewinds Associates I Steam Generator Failure at San Onofre

FAIREWINDS ASSOCIATES STEAM GENERATOR FAILURE AT SAN ONOFRE shut. In order to answer long-term reliability and safety concerns, the metallurgy of the tubes must be compared to the old design and fabrication. Moreover, tubes from both San Onofre Unit 2 and 3 must be removed and thoroughly examined in order to compare any subtle differences in fabrication between the two units.

Unfortunately, it appears that the mobilization of an NRC Augmented Inspection Team to only Unit 3 is an effort by the NRC and Edison to obfuscate the issue at San Onofre and not conduct an orthodox, thorough, and requisite engineering root cause analysis.

Without a thorough examination of the tubes in San Onofre Unit 2 the cause of the tube thinning will remain unresolved creating a significant safety issue. If the NRC allows either San Onofre reactor to restart without a thorough root cause analysis and another tube or tubes were to fail, radioactive releases might be significantly larger than those that occurred after the January 2012 tube leak. Such an accident would cause implementation of the California emergency evacuation plan and closing of the San Clemente beach and Interstate 1-5, potentially for an extended period of time.

Aerial image of San Onofre Nuclear Plant proximity to US Interstate 5. Courtesy Google Images http://dailygoogleearth.files.

wordpress.com/2011/05/

sanonofre nuclear.jpg Page 10 I Fairewinds Associates I Steam Generator Failure at San Onofre

FAIREWINDS ASSOCIATES STEAM GENERATOR FAILURE AT SAN ONOFRE According to the published literature, the replacement steam generators for Units 2 and 3 have identical specifications. Therefore, allowing San Onofre Unit 2 to restart on the mistaken belief that it is somehow different that Unit 3 before the root cause of the problem is definitively known defies logic. The NRC has failed to adequately protect public health and safety during a very similar incident involving serious cracking on a BWR. The Quad Cities reactor experienced severe vibration induced cracking of its steam dryer in 2002. In a BWR, the steam dryer is a major component in the reactor. In a PWR, such as at San Onofre, the steam dryer is an integral part of the steam generator.

The owner of the plant and the reactor designer believed that they understood the problem and had made the appropriate repairs. They then started the reactor back up.

One year later in 2003, the steam dryer had cracked yet again. In fact the second cracks were worse than the first.

If San Onofre Unit 2 is allowed to start up prior to a complete root cause analysis, steam induced dryer cracking like that at Quad Cities may occur in the steam generators at San Onofre. In 2002, Quad Cities told the NRC that the repairs would successfully solve the first failure. In the Preliminary Operating Experience Report OE16403, issued after the second steam dryer failure, the NRC said that:

1. After the first failure, "Several teams of Exelon Nuclear, General Electric and industry experts are assembled to ...determine the

...corrective actions."

2. Following the second steam generator failure, the NRC said that the second failure was caused when "GE Nuclear Energy and the licensee did not foresee this phenomenon."

What might San Onofre fail to foresee as the true problem in its rush to start Unit 2 back up?

Allowing either San Onofre Unit 2 or Unit 3 reactors to restart before the root cause of the problem is definitively known defies logic.

Page 11 I Fairewinds Associates I Steam Generator Failure at San Onofre

FAIREWINDS AssoCIATES STEAM GENERATOR FAILURE AT SAN ONOFRE The Edison should not be rushing to restart these San Onofre reactors based upon a hunch or 'preliminary conclusion' that a safety problem may have been resolved and hoping that a root cause analysis, once finally attempted, may support an initial guess. If Edison restarts San Onofre Unit 2 or Unit 3, it will be impossible to conduct a thorough root cause analysis.

The residents of southern California will be left wondering when the next break will occur and if that one break will cause a significant radiation release. Therefore, in conclusion and despite the Nuclear Regulatory Commission's (NRC) refusal to exert its regulatory authority on when the plants are permitted to start up again, Fairewinds Associates recommends that both San Onofre Unit 2 and Unit 3 remain shut down until the root cause of each nuclear reactor rapid steam generator tube failures are understood and repaired, reliability is assured, and radioactive releases are prevented.

Friends of the Earth retained FairewindsAssociates (fairewinds.com) to conduct this review and issue this report.

Arnie Gundersen, MSNE, and chief engineer for Fairewinds authored this report. FairewindsAssociates is a paralegal services and expert witness firm specializing in nuclear engineering and nuclear safety analysis in the US, Canada and overseas. Mr. Gundersen, who has 40-years of nuclearpower engineering experience, is a former nuclearindustry senior vice president who earned his Bachelor and Master Degrees in nuclearengineering from RPI, holds a nuclear safety patent, and was a licensed reactoroperator. During his industry-centered career,Mr. Gundersen managed and coordinated projects at 70-nuclearpower plants in the US. As part of the education mission of the 501(c) 3 non-profit organization Fairewinds Energy Education, Mr. Gundersen speaks to the public about the lack of adherence throughout the world to nuclear safety regulations.

Page 12 Fairewinds Associates I Steam Generator Failure at San Onofre

Attachment 3 San Onofre CascadingSteam Generator Failures Createdby Edison: Imprudent Design andFabricationDecisions CausedLeaks Fairewinds Associates April 12, 2012

Fairewinds Associates, Inc Burhngton, VT o54o8 Phone 802-865-9955 contact@fairewinds.com San Onofre Cascading Steam Generator Failures Created By Edison Imprudent Design And FabricationDecisions Caused Leaks This report was completed on April 101h 2012 for Friends of the Earth. On April 1 1 th it was reported that Southern California Edison (Edison) had identified the same type of wear in the tubes of the Unit 2 Replacement Steam Generator as those originally discovered in Unit 3.1 Edison's findings confirm the original analysis released by Fairewinds in their March 2 6th report and is confirmed by the evidence reviewed for the report that follows below.

Executive Summary In Fairewinds Associates March 26, 2012 report entitled Steam GeneratorFailuresAt San Onofre: The Need For A Thorough Root Cause Analysis Requires No Early Restart, Fairewinds recommended that both San Onofre Unit 2 and Unit 3 remain shut down until the 'root causes' of these twin nuclear power plants' rapid tube failures are understood and repaired, reliability is assured, and radioactive releases are prevented. In its March 23, 2012 letter to the Nuclear Regulatory Commission (NRC), Edison suggests that there are no similarities in the unique and simultaneous problems with each San Onofre reactor's leaking steam generator, therefore restarting Unit 2 is a decision that is separate and apart from any restart considerations for Unit 3.

Additionally, in its March 27, 2012 Confirmatory Action Letter, the NRC appears to agree with Edison and develops two separate review and restart procedures for Unit 2 and Unit 3.

Due to the common elements connecting the tube failures in both Units 2 and 3, Fairewinds believes that Units 2 and 3 must be analyzed concurrently. Moreover, in Fairewinds opinion, the most likely common element or 'root cause' of the simultaneous steam generator tube failures at each Unit may be traced to Edison's unwarranted steam generator design changes. San Onofre's original steam generator tubes lasted almost 30-years, and it is probable that replacement steam generators meeting the original design criteria would have lasted for another 30 years.

San Onofre Has Always Been Unique Originally designed and built by Combustion Engineering (CE), San Onofre's nuclear steam generators are a very unique design that is radically different from all other Pressurized Water Reactor (PWR) designs. The CE design at San Onofre has only two steam generators while all other PWR's of comparable size to San Onofre have four U-Tube Steam Generators. In order to produce as much power as other PWR's from only two steam generators, each steam generator at San Onofre is twice as large as those at similar PWR's with comparable power output.

I http://sciencedude.ocregister.com/2012/04/11 /odd-tube-wear-seen-in-both-onofre-reactors/170292/

Page 2 of 7 Therefore, both the original and the replacement steam generators at San Onofre are some of the largest steam generators that have been designed or manufactured anywhere in the world. NRC spokesperson Victor Dricks agreed when he said, "...this problem is specific to the steam generators at San 2Onofre... They are virtually a unique design and no other plant in the United States has them."

Extensive Unreviewed Design Chan2es Were Made To San Onofre's Replacement Steam Generators Edison decided to replace each San Onofre steam generator due to tube deterioration and degradation that slowly evolved during each Unit's 25-years of operation. Documents reviewed show that the four replacement steam generator specifications are identical to each other and they were purchased together under a single contract with Mitsubishi Heavy Industries (MHI).

However, rather than simply rebuild the steam generators to their original design specifications, Edison decided to extensively modify the original San Onofre steam generator design.

Furthermore, none of the design modifications were necessary for operation of either San Onofre Unit 2 or 3, and in fact were added by Edison without adequate consideration for their impact upon the reliability or safety of steam generator and the reactor.

A joint report3 prepared by Edison and (MHI) describes in great detail numerous changes Edison made to the original steam generator design. The evidence reviewed in the Edison/MHI report shows that extensive modifications were made to the original design without the requisite total engineering analysis. Additionally, MHI has only constructed one other replacement steam generator for a CE designed PWR, and that reactor produced one-half the power output created by the San Onofre Units.

Cascadin2 Design Changes By Edison Created Steam Generator Failure The evidence provided in the Edison/MHI report shows that a cascadingseries of deliberate design changes likely caused the tube failures and tube degradationthat has shut down San Onofre Units 2 and 3 since January of 2012.

Approximately seven years ago Southern California Edison began the process of replacing San Onofre's original steam generators. Fairewinds believes that two key management decisions created the cascading failure sequence that now exists in San Onofre's steam generators:

2 http://www.kpbs.org/news/2012/apr/06/nrc-chair-visit-troubled-san-onofre-nuclear-power-/ Friday, April 6, 2012, KPPS Radio.

3 Improving Like-For-Like Replacement Steam Generatorsby Boguslaw Olech of Southern California Edison and Tomouki Inoue of Mitsubishi Heavy Industries, NuclearEngineeringInternational,January 2012, page 36-38.

http://edition.2agesuite-professional.co.uk/launch.aspx?referral=other&pnum-=36&refresh=K0s3a21GRg61 &EID=af75ecb l-5b23-49be-9dd6-d806f2e9b7b5&skip=&p=36

Page 3 of 7

1. First, a decision was made to change the tube alloy from Inconel 6004 to Inconel 6905.

Inconel 600 was the tube alloy in use during the construction of San Onofre and other PWR's. However, many replacement steam generators have been constructed with Inconel 690 that to date appears to have allowed most replacement steam generators to operate longer with less deterioration. Fairewinds believes that if the only fabrication change by EdisoniMHI to the San Onofre steam generator had consisted of a tube alloy composition change from Inconel 600 to 6906, it was most likely that the San Onofre steam generators would have continued to operate successfully well beyond the 40-year license for either Unit.

Furthermore, a document review of nuclear industry reports reviewing PWR reactor replacement steam generator installations shows that the single composition change from the tube alloy Inconel 600 to 690 appears to be a well established PWR industry fabrication standard, while the addition of almost 400 new tubes to the replacement steam generators created unanalyzed flow and stress design changes that have severely compromised reactor operability for Units 2 and 3.

2. The second key fabrication change supplanted to the San Onofre steam generators by the Edison/MiHI team increased the total number of tubes in each steam generator by almost 400 tubes to more than 104% of each generator's original design. Each Original Steam Generator contained 9350 tubes 7 while the Replacement Steam Generators each contain 4 Inconel 600 is a nickel-chromium super alloy. It is readily weldable and non-magnetic in nature and used in various corrosion resisting applications. The chromium content of the Inconel 600 provides resistance to weaker oxidizing environments while the high nickel content provides exceptional resistance to chloride stress corrosion cracking and a level of resistance to reducing environments. http://sbecpl.com/products/nickel-alloys/inconel-600/

5 INCONEL alloy 690 is a high-chromium nickel alloy having excellent resistance to many corrosive aqueous media and high-temperature atmospheres. The alloy's high chromium content gives it excellent resistance to aqueous corrosion by oxidizing acids (especially nitric acid) and salts, and to sulfidation at high-temperatures. In addition to its corrosion resistance, alloy 690 has high strength, good metallurgical stability, and favorable fabrication characteristics. http://www.specialmetals.com/products/inconelalloy690.php 6 FrictionInduced Vibrations, R. Ibrahim, Encyclopedia of Vibration, 2001, Pages 589-596 http://www.sciencedirect.com/science/article/pii/S0029549303001870 "In the steam generators of nuclear power plants, the flow of cooling water can cause the tubes to vibrate, resulting in fretting wear damage due to contacts between these tubes and their supports. The tubes are made of Inconel 690 and Inconel 600 and the supports are made of STS 304. In this paper, fretting wear tests in water were performed using the materials Inconel 690 and Inconel 600 in contact with STS 304. Fretting tests using a cross-cylinder type set up were conducted under various vibrating amplitudes and applied normal loads in order to measure friction forces and wear volumes. Also, conventional sliding tests using a pin-on-disk type set up were carried out to compare these test results.

In the fretting tests, friction force was found to be strongly dependent on normal load and vibrating amplitude.

Coefficients of friction decreased with an increase in the normal load and a decrease in the vibrating amplitude applied. Also, the wear of Inconel 600 and Inconel 690 was predicted using a work rate model. Depending on the normal load and vibrating amplitude applied, distinctively different wear mechanisms and often drastically different wear rates occurred. It was found that the fretting wear coefficients for Inconel 600 and Inconel 690 were 9.3xl0-t5 and 16.2x 10- 5 Pa', respectively. This study shows that Inconel 690 can result in lesser friction forces and exhibits less wear resistance than Inconel 600 in room temperature water."

7 http://www.cpuc.cajov/Environment/info/aspen/sanonofre/feir/vl/figs/Fig%20B-4.pdf

Page 4 of 7 9727 tubes 8. Fairewinds believes it was this management decision to increase the number of tubes that lead in turn to a series of cascading design changes that created the serious problems San Onofre is experiencing in 2012.

Fairewinds believes that Edison's decision to cram an additional 377 tubes into the replacement steam generators was the root cause to the steam generator leaks encountered in 2012.

Moreover, Edison's decision to add 377 more tubes to the existing 9,350 tubes rather than perform the like-for like steam generator replacement for which it applied at San Onofre caused a cascading series of replacement steam generator failures.

Cascading Fabrication Changes Following Edison Decision To Dramatically Increase The Number of Tubes Similar, but smaller, Tubesheet image - prior to drilling for tube insertion The original San Onofre steam generator contained a tubesheet9 , which is a metal disc approximately 13-feet in diameter and slightly less than two feet thick, located near the bottom of the steam generator. Due to the already extremely large size of the CE steam generators, this tubesheet'° is one of the largest tubesheets ever fabricated after which 18,700 holes (9,350 in-8 http://www.nrc.gov/reading-rm/doc-collections/news/2012/12-01l.iv.pdf.

9 Ibid.

10 A 'tubesheet' divides the steam generator into the primary side and a second side, which 'tubesheet' has an array of holes having U-shaped heat transfer tubes inserted therein, which communicate between the inlet section and outlet section of the primary side of the steam generator. In operation, the heat pressurized fluid passes through the U-shaped heat transfer tubes and is discharged from the outlet section of the primary side of the steam generator to a line, containing a primary coolant pump, back to the reactor in a continuous closed loop.

http://www.patentstorm.us/patents/4728486/fulltext.html

Page 5 of 7 hot/9,350 out-cold) were then drilled. This metallic disk [see image] serves as an anchor into which both sides of the U-tubes are inserted. Not only is the tubesheet extraordinarily heavy, but also there can be a pressure difference of approximately 2,000 pounds per square inch (psi) between the radioactive water on one side and non-radioactive water on the other.

In order to support the enormous tubesheet metallic disk, the original steam generator design at San Onofre contained a 'stay cylinder' in the center of the tubesheet that is a support pillar designed to relieve the weight in the middle of the tubesheet.

1. When Edison decided to cram in additional steam generator tubes, the fabrication technique created by Edison/MiHI for the San Onofre steam generators necessitated the removal of the 'stay cylinder' so that more tube holes could be drilled through the tubesheet. The Edison/MHI decision to add additional tubes and replace this key support pillar was part of the cascading fabrication changes that caused additional stresses and steam generator failure.

2. Removing the stay cylinder required additional cascading fabrication changes. Because the tubesheet was no longer supported in the center by the stay cylinder, Edison/MHI required the fabrication of a thicker tubesheet so that it could bear the additional stress without a stay cylinder. This change in the tubesheet thickness meant yet another design change by reducing the volume of water in the steam generator and changing the flow pattern and also reducing the inspection access area beneath the tubesheet that is required to fit personnel and equipment for tube inspection.

3. Changing the structural loads on the tubesheet have not only affected the reliability of the steam generators but also should have raised a serious safety concern because the tubesheet is the key barrier to keeping radiation inside the containment. Should the tubesheet fail, radiation within the reactor would bypass the containment and pass directly into the environment. Due to the installation of the 'stay cylinder' in the original San Onofre steam generator configuration, a tubesheet failure and subsequent radiation release is considered to be beyond the calculations for a design basis accident at San Onofre. Yet Edison chose to challenge this critical safety barrier and licensing parameter by removing the "stay cylinder" in order to install more, unnecessary tubes.

4. Fabricating more tubes increased nuclear reactor core flow, which was unacceptable because it changed the original design basis safety calculations for cooling the reactor.

For that reason Edison welded a flow-restricting ring into the steam generator nozzle in order to reduce the flow of cooling water back into the reactor to the original design parameters, which also changes the flow distribution to the tubes. Thus significant operational changes were also made to the radioactive side of the steam generator as a result of Edison's addition of more steam generator tubes.

Page 6 of 7 Typical Combustion Engineering Steam Generator STEAM 0OJTLET DEFLECTOR STEAM 175 STEAM DRLM DRYERS 165STEAM 32 STEAM SEPARATORS\ DRYER DRAINS SECONDARY MANWAY (2) -

• ~I NSTRUMENT NOZZLE NORMAL RECIRCULATION RSER WATER SUMP LEVEL RECIRCULAT ION SLAIP DRAINS AUXIL ARY FEEDWATER -MAIN FEEDWATER NOZZLE _-

NOZZLE MAJN FEED RING NSTRLMENT WRAPPER \

NOZZLE BATWING EVAPORATOR (TUBE B4.INDLE)

EGG CRATE SUPPORTS

. VERTICLE LU-TkoBES SECONDARY HANDHOLE :2)

BOTTOM BLOWDOWN

& DRAIN NOZZLE TUBESHEET HOT LEG COLD LEG OUTLET (2)

INLET

5. All of these changes necessitated even more fabrication changes within the steam generator. For example, more tubes meant that the tube supports had to be modified in an attempt to avoid the increased vibration caused by the flow changes induced by the Edison/MHI fabrication changes. The feedwater distribution ring inside the steam generator was also dramatically modified in order to avoid a serious flow induced water hammer.

Page 7 of 7 Conclusions In Fairewinds' opinion, the vibration between the tubes caused the steam generator leaks and degradation uncovered in January 2012 and was due to the simultaneous implementation of numerous unreviewed fabrication and design changes to the replacement steam generators by Edison/MHI. Almost all of these changes were avoidable if San Onofre's management had not made the decision to cram an additional 377 unnecessary tubes into each steam generator. This Edison decision made seven years earlier was not in the best interests the San Onofre Units and costs to implement these unreviewed changes were not prudent.

Fairewinds believes that if the original steam generators had been replaced with duplicates (like-for-like) with the only significant change that of the tube alloy composition from Inconel 600 to 690, it is most likely that the San Onofre steam generators would have continued to operate successfully well beyond the 40-year license for either Unit. However, the extensive unreviewed design and fabrication changes implemented by EdisoniMHI to the new San Onofre steam generators are hardly a like-for-like change.

The evidence reviewed by Fairewinds shows that the replacement steam generators for Units 2 and 3 have identical specifications and were subject to identical operating conditions. While Unit 2 has experienced more tube degradation and Unit 3 has experienced deeper cracking in fewer tubes, both Units require a concurrent root cause analysis due to their identical specifications and operating conditions.

Friends of the Earth retainedFairewindsAssociates (fairewinds.com), a paralegalservices and expert witness firm specializing in nuclear engineeringand nuclear safety analysis, to conduct this review and issue this report. Arnie Gundersen, MSNE, and chiefengineerfor Fairewinds Associates authoredthis report.

Attachment 4 San Onofre's Steam Generator Failures Could Have Been Prevented Fairewinds Associates May 14, 2012

SAN OOFRES FIIJ S COUL I T

- 4 S InI

~ARNIEGUND.RSEN

• 0

2 of 13 San Onofre's Steam Generator Failures Could Have Been Prevented Summary Southern California Edison's four replacement steam generators at their San Onofre Nuclear Generating Station failed in less than two years of operation, while the original equipment operated for 28 years. Fairewinds has been analyzing the data in order to determine how such an expensive investment could fail so quickly.

In June of 2006 Edison informed the NRC that the replacement steam generators to be manufactured by Mitsubishi would be fabricated to the same design specifications as the original San Onofre Combustion Engineering (CE) steam generators. According to Nuclear Engineering International, Edison has admitted that this was a strategic decision to avoid a more thorough license amendment and review process.'

At SONGS, the major premise of the steam generator replacement project was that it would be implemented under the 10CFR50.59 rule, that is, without prior approval by the US Nuclear Regulatory Commission (USNRC). To achieve this goal, the RSGs were to be designed as 'in-kind' replacement for the OSGs in terms of form, fit and function.2 Fairewinds finds that there are numerous changes to the San Onofre steam generators that are not like-for-like or "in-kind".

Furthermore, the facts reviewed by Fairewinds makes it clear that if Edison had informed the NRC that the new steam generators were not like-for-like, the more thorough NRC licensing review process would have likely identified the design problems before the steam generators were manufactured.

Finally, Fairewinds finds that tube plugging is not the solution to the vibration problem 3 and that the damaged steam generators will still require major modifications with repair and outage time that could last more than 18 months if Edison and Mitsubishi are even able to repair these faulty designed steam generators. However, Fairewinds finds that the safest long-term action is the replacement of the San Onofre steam generators.

3 of 13 Analysis The requirements for the process by which nuclear power plant operators and licensees may make changes to their facilities and procedures as delineated in the safety analysis report and without prior NRC approval are limited by specific regulations detailed in The Nuclear Regulatory Commission's 10 CFR Part50, Domestic Licensing of Productionand Utilization Facilities,Section 50.59, Changes, Tests and Experiments.

The implementing procedures for the 10 CFR 50.59 regulations have eight criteria that are important for nuclear power plant safety. (These eight criteria are provided in Table 1, footnote A below.)

These implementing procedures created for 10 CFR. 50.59 require that the license be amended unless none of these eight criteria are triggered by any change made by Edison at San Onofre. If a single criterion is met, then the regulation requires that the licensee pursue a license amendment process.

By claiming that the steam generator replacements were a like-for-like design and fabrication, Edison avoided the more rigorous license amendment process. From the evidence reviewed, it appears that the NRC accepted Edison's statement and documents without further independent analysis. In the analysis detailed below, Fairewinds identified 39 separate safety issues that failed to meet the NRC 50.59 criteria. Any one of these 39 separate safety issues should have triggered the license amendment review process by which the NRC would have been notified of the proposed significant design and fabrication changes.

As the NRC guidelines state:

"(c)(1) A licensee may make changes in the facility as described in the final safety analysis report (as updated), make changes in the procedures as described in the final safety analysis report (as 1.187-A-1 updated), and conduct tests or experiments not described in the final safety analysis report (as updated) without obtaining a license amendment pursuant to § 50.90 only if: (i)A change to the technical specifications incorporated in the license is not required, and (ii) The change, test, or experiment does not meet any of the criteria in paragraph (c)(2) of this section." 4 [Emphasis Added]

In its previous reports, Fairewinds identified at least eight modifications to the original steam generators at San Onofre.

Table 1 below was designed to compare the eight major design modifications that Fairewinds identified in its analysis with the eight criteria the NRC applies to the license review process in order to determine whether or not a new license amendment process is required. The major design changes are located at the top of the table, and the NRC Criteria are listed in the left hand column of table. The term SSC stands for Systems, Structures and Components. A green No means that the like-for-like criteria were indeed met and that no license amendment was required.

A red Yes means that Edison should have applied for a license amendment.

Table 1 shows that 7 out of 8 of the major design changes to the original steam generators meet a total of 39 of the NRC's 50.59 criteria requiring amendment to the license.

4 of 13 Table 1 Steam Generator Design Changes Identified By Fairewinds Compared With The NRC's Like-For-Like Criteria 50:59 (B) Remove Change Tube alloy Add tubes Change Add flow Additional Feed water Criteria stay tube sheet change tube restrictor water distribution (A) cylinder support volume ring i - Accident Yes (1) Yes (1) No Yes (3,4) Yes (3,4,8) No No No Frequency Increase ii - Increase Yes (1) Yes (1) No Yes (3,4) Yes (3,4,8) No No No in SSC Malfunction occurrence iii - Accident Yes (1) Yes (1) No Yes (3,4) Yes (3,4,8) Yes (2) Yes (2,5,6) No consequent increase iv - Increase Yes (I) Yes (1) No Yes (3,4) Yes (3,4,8) Yes (2) Yes (2,5,6) No in SSC consequence of malfunction v - Create Yes (1) Yes (1) No No No Yes (2) Yes (2,5,6) Yes (3,7,8) unanalysed accident vi - Create Yes (1) Yes (1) No No Yes (3,8) Yes (2) No Yes (3,7,8) new malfunction vii -Alter Yes (1) Yes (1) No Yes (3) No No No No fission product barrier viii - Change Yes (2) Yes (2) No Yes (2) Yes (2,8) Yes (2) Yes (2,5,6) No design basis evaluation method Table Footnotes A - The criteria listed in the left column in the table above refers to the criteria as laid out in the NRC Guidelines 5 which states as follows:

"(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);

(ii) Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the final safety analysis report (as updated);

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

(iv) Result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the final safety analysis report (as updated);

(v) Create a possibility for an accident of a different type than any previously evaluated in the final safety analysis report (as updated);

(vi) Create a possibility for a malfunction of an SSC important to safety with a different result than

5 of 13 any previously evaluated in the final safety analysis report (as updated);

(vii)Result in a design basis limit for a fission product barrier as described in the FSAR (as updated) being exceeded or altered; or (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."

B - The horizontal axis contains a list of design changes made by Edison and whether they meet or have not met the criteria as set out in 10 CFR 50.59.

1 - The Steam Generator Replacement Project modified the tube sheets and stay cylinder that are a containment barrier - The NRC was not informed nor did it specifically approve these changes to the containment barrier as they were apparently not addressed under Edison's analysis for the 10 CFR 50.59 process; 2 - The Mitsubishi thermo hydraulic code is inadequate to assess flow inside the Steam Generators that dramatically affect the ability to cool the nuclear reactor core in the event of an accident; 3 - The Steam Generator Replacement Project increases the consequences of a steam line break accident; 4 - The Steam Generator Replacement Project has already proven to increase the frequency of tube failure; 5 - The Steam Generator Replacement Project changed the volume of primary coolant because more tubes were added, which changes the Final Safety Analysis Report; 6 - The Steam Generator Replacement Project changed the flow rate of primary coolant, which changes the Final Safety Analysis Report; 7 - The Steam Generator Replacement Project changed the potential for water hammer. Given that the Mitsubishi thermo hydraulic code is inadequate, the potential for water hammer is increased; 8 - The Steam Generator Replacement Project created steam binding at top of steam generator.

The steam generator is designed to remove heat in the event of an accident and its role has been compromised.

Ramifications Of Edison's Decision To Avoid The License Amendment Process Edison's strategic goal was to avoid the process of license amendment according to the January 2012 article in Nuclear Engineering International NEI Magazine 6. Had Edison notified the NRC that the new steam generators at San Onofre were not a like-for-like replacement, a more thorough review through the license amendment process would have been required. Given that scenario, it is likely that the requisite and thorough NRC review would have identified the design and fabrication inadequacies that appear to have caused the San Onofre steam generator tube failure.

More specifically, Fairewinds believes that the NRC would have identified the inadequacy of the Mitsubishi Heavy Industry computer code applied to validate the tube design and vibration pattern prior to fabrication. Mitsubishi's computer code was simply not capable of analyzing Combustion Engineering (CE) designs like San Onofre and was only qualified for Westinghouse designs that are not similar to the original CE steam generator design. In NRC licensing jargon, the Mitsubishi design codes were not benchmarked for the CE Design7.

While Mitsubishi Heavy Industry has been supplying steam generators for many years in Japan, it did so under a specific license from Westinghouse for Westinghouse nuclear reactors.

Although Mitsubishi made several incremental changes to the Westinghouse design, such as switching to alloy 690 tubing and the use of stainless steel broached plate tube supports, Mitsubishi has had very little experience with the tight tube pitch and the egg crate design used in the original CE design for San Onofre.

6 of 13 Figure 1: Broached Tube Si ed To Keep Tubes From Rattling Broached (quatrefoil and trefoil) tube support plates (TSPs)8 Figure 2: Eggcrate Tube Support Plate - Designed To Keep Tubes From Rattling Horizontal Tube Supports (Eggcrate) 9

7 of 13 The original steam generators designed and manufactured by CE for San Onofre were successfully operated 28 years. Moreover the original steam generators had a triangular tube pitch pattern, very closely packed U-tubes, and unique egg-crate tube supports that kept the tubes from vibrating and colliding. The pitch to diameter ratio of tubes in the original CE generators is dramatically different from any of the Westinghouse generators fabricated by Mitsubishi.

Moreover, an NRC licensing review would have identified the fact that the Mitsubishi computer design code, which is based upon Westinghouse models, was not appropriate for design changes to the San Onofre replacement steam generators originally designed by CE.

Another problem with the San Onofre steam generators is that Edison and Mitsubishi made a very significant design change that magnified the San Onofre steam generator stresses and vibrations by removing the main structural pillar called the stay cylinder in order to fit an additional 400 tubes into the unique and already tightly packed design. Furthermore, this design is also bigger than anything Mitsubishi Heavy Industries (MHI) had ever fabricated or designed.

The NRC license amendment review process would likely have identified these and other problems.

The Actual Steam Generator Problem As water moves vertically up in a steam generator, the water content reduces as more steam is created. When the volume of steam is much greater than water then the flow resistance of the water/steam mixture passing through the tube supports accounts for one third of the total resistance at the top of the steam generator. Therefore to avoid vibration at the top of the tubes, Mitsubishi needed to specifically analyze the type of tube support to use in this unique application.

The flow resistance of the Mitsubishi broached plate is much higher than that of the original Combustion Engineering egg crate design because the tubes are so tightly packed in the original CE San Onofre steam generators. By reviewing the documents thus far produced, it appears that due to Mitsubishi's fabrication experience with broached plates, both Edison and Mitsubishi missed this key difference in the design and fabrication of the new San Onofre steam generators.

Not only is Mitsubishi unfamiliar with the tightly packed CE design, but also Edison's engineers created so many untested variables to the new fabrication that this new design had a significantly increased risk of failure. As a result of the very tight pitch to diameter ratios used in the original CE steam generators, Mitsubishi fabricated a broached plate design that allows almost no water to reach the top of the steam generator.

The maximum quality of the water/steam mixture at the top of the steam generator in the U-Bend region should be approximately 40 to 50 percent, i.e. half water and half steam. With the Mitsubishi design the top of the U-tubes are almost dry in some regions.' 0 Without liquid in the mixture, there is no dampingl" against vibration, and therefore a severe fluid-elastic instability developed.

8 of 13 In response to the Edison/Mitsubishi steam generator changes, the top of the new steam generator is starved for water therefore making tube vibration inevitable. Furthermore, the problem appears to be exacerbated by Mitsubishi's three-dimensional thermal-hydraulic analysis determining how the steam and water mix at the top of the tubes that has been benchmarked against the Westinghouse but not the Combustion Engineering design.

The real problem in the replacement steam generators at San Onofre is that too much steam and too little water is causing the tubes to vibrate violently in the U-bend region. The tubes are quickly wearing themselves thin enough to completely fail pressure tests. Even if the new tubes are actively not leaking or have not ruptured, the tubes in the Mitsubishi fabrication are at risk of bursting in a main steam line accident scenario and spewing radiation into the air.

This Tube Damage Cannot Be Repaired Edison claims that the proximate cause of these U-tube failures at San Onofre is high vibration, and it has embarked upon a process of plugging some of these damaged tubes in hopes of quickly restarting one or both units. Fairewinds believes that this damage is occurring on the outside of the tubes where they collide with each other, while access to the tubes for repair and/or plugging can only be conducted from inside the tubes. Space limitations due to the tight fit of the 9,700 tubes (19,400 holes in the tube sheet) in each steam generator have made it impossible to access the outside of the U-tubes for inspection where the wear is actually occurring.

Presently, the Edison approach is to plug tubes in the most heavily damaged zone of each steam generator. Plugging the tubes only eliminates the radioactive water inside the tubes, but it does not eliminate the vibration, so the plugged tubes will continue to vibrate and damage adjacent tubes. More than 500 tubes have already been plugged in Unit 2 and more than 800 tubes have been plugged in Unit 3.12 The number of plugged tubes is still considerably smaller than the number of tubes already ascertained as damaged in both steam generators. To date, Edison has not provided adequate data to compare damaged tubes to plugged tubes.

Initially, in March 2012, Edison claimed that as part of the Electric Power Research Institute's (EPRI) criteria used in the in-situ pressure testing of the Steam Generators, it was required to plug about one dozen tubes in the San Onofre steam generators. However, in May 2012, Edison announced it had plugged 1300 tubes, more than one hundred times the number of tubes required by the EPRI criteria. According to the industry steam generator experts interviewed by Fairewinds, Edison did not plug these additional tubes because they had failed, but rather Edison needed to plug these particular tubes because they would likely fail during a main steam line break accident.

If a steam line break accident were to occur, the depressurization of the steam generator caused by the steam line break coupled with the lack of water at the top of the steam generators would cause cascading tube failures, involving hundreds of tubes. The cascading tube failures would pop like popcorn and the cascading failures would cause excessive offsite radiation exposures.

In an attempt to avoid a severe steam line break accident Edison prophylactically plugged additional tubes.

9 of 13 Fairewinds investigation has found that plugging the tubes is not a sure solution, because it fails to deal with the root causes of a failed design and it relies upon the incorrectly applied Mitsubishi 3-Dimensional steam analysis to determine which tubes should be plugged.

Realistically, the 3-D steam analysis is not accurate enough to apply to such important safety-related determinations. To make such mathematical risk 3-D analysis, a very large margin of error must be applied, and that has not been done. For example, if the 3-D steam analysis determines that plugging 100 tubes is a solution, then plugging ten times that number might be the appropriate solution due to the mathematical errors in the 3-D analysis being applied by Edison and Mitsubishi.

Fairewinds concludes that plugging the tubes will never solve the underlying problem because vibration is the result not the root cause of the steam generator problems at San Onofre. The actual problem is a variety of design changes that have caused too much steam and too little water at the top of the steam generators. Plugging tubes cannot repair these design changes created and that are causing the tubes to collide with each other.

The tubes that Edison has already plugged on the inside will continue to vibrate because they are being pushed by steam and water from the outside. Therefore Fairewinds concludes that Edison's solution of plugging the inside of the tubes will not lessen the risk of an accident or stop the ongoing vibrational damage that is occurring to the inaccessible outside of the San Onofre steam generator tubes.

Options For Continued Operation Of The San Onofre Reactors Complete Replacement The ongoing plugging of the tubes will not eliminate the vibrational failure mechanism causing tube failures. Over time, the damaged tubes that are plugged will in turn damage more tubes. Therefore, Fairewinds believes that the only sure solution to this significant safety issue is to once again cut open the reactor containment and install new steam generators that replicate the original CE design.

Due to the significant risk of a steam generator tube rupture accident in such a highly populated and vulnerable area, both San Onofre Unit 2 and Unit 3 should remain shut down until such a significant safety threat can be mitigated with the fabrication of new like-for-like steam generators adhering to the original CE design. If all the appropriate steps are taken in design and fabrication of new CE replica steam generators, and the proper procedures are taken to repair and reseal the San Onofre containment coupled with requisite NRC oversight, Fairewinds estimates that the entire process might take Edison approximately four years and cost in excess of $800,000,000,. not including replacement power while the Units remain shut down.

10 of 13 Repair In Place While technically this would be an extremely challenging repair process, it may be possible to cut the steam generators apart while still inside the containment. Such a process would take approximately 18 months to make repairs and then weld the steam generators back together again without cutting the containment open. Cutting the top off the steam generators would allow construction personnel access so that additional supports could be inserted into the U-tube region. Smaller replacement packages would fit through the existing equipment hatch and the containment would not be compromised another time.

The cost for these repairs would be less than completely redesigning and manufacturing new steam generators and replacement power costs would be less. However, it is still reasonable to estimate that this cost would exceed $400,000,000 without replacement power, not including replacement power while the Units remain shut down.

There are two possible alternatives, both of which would require an additional analysis of the overall steam generator flow patterns to ensure that no new problems are created in the process of attempting to mitigate the damage from these design flaws and fabrication errors. The two alternatives are:

1. Because too much steam and too little water in the U-bend region cause the vibration problems, it might be possible to reduce the steam/water mixture qualities in the U-bend area by changing the internal structures to divert some of the internal recirculating flow into the U-bend region.

2. Another possible solution would require replacing the steam-water separators. The Mitsubishi separators require a water level that is quite low in the steam drum, and cannot be raised. Changing the separators to a different design may allow more water to reach the top of the tubes and thereby stop the tube vibration and wear.

Power Reduction Reducing power does not provide a remedy for the underlying structural problems that are creating the vibration that has damaged and will continue to damage tubes deep inside the San Onofre steam generator. Edison has suggested that plugging tubes and operating at indeterminate reduced power levels for the remainder of the life of the plant may be a solution to the San Onofre tube vibration problem. Unfortunately this course of action would leave San Onofre operating with a significant safety risk if the NRC were to allow the reactors to restart.

The concept of reducing the power output from the San Onofre reactors will not change either the inside steam generator tube water temperature or the steam temperatures outside of the tubes. Reducing the power output will also not change the 2200-pound per square inch pressure within the tubes or the 1,000-pound pressure outside the tubes.

Operating at reduced power will not prevent previously damaged tube supports and plugged tubes from vibrating and damaging surrounding tubes and tube supports, and it will worsen the existing damage.

More importantly, Fairewinds concern is that operating the San Onofre reactors at a lower power and flow rate might actually create a resonate frequency within the steam generators at which some of the tubes will vibrate as bad or worse than they did originally.

11 of 13 Because the plugged tubes are now filled with air their weight has changed, and therefore the plugged tubes will vibrate with a different amplitude and frequency. The inaccuracies in the Edison and Mitsubishi computer code do not allow Edison and Mitsubishi to conduct a resonant frequency analysis proving that such a problem will not occur.

It is impossible to determine exactly what is happening inside an operating steam generator. For example, at Millstone 2, a smaller CE reactor, the steam generator tube supports began to disintegrate due to vibration, and there was no method to alert the operations staff that such deterioration was occurring. This challenging problem was finally detected when the Millstone 2 was shut down for a refueling, and small cameras meant to inspect the steam generator found rubble on the tube sheet at the base of the tubes.

Historical evidence from other operating nuclear reactors that have attempted to mitigate vibrational damage by using power reductions rather than solving the resonant frequency issues have in fact compromised other nuclear safety related components by operating at reduced power.

In 2002 the Exelon Quad Cities Nuclear Power Plant in Illinois operated its Unit 2 reactor at reduced power in order to eliminate vibrationally induced damage causing high moisture carryover in its steam dryer. While the power reduction temporarily reduced moisture carryover, the problem reoccurred and a shutdown was ordered causing an extended unplanned outage. Vibrationally induced severe cracking was discovered in the steam dryer and repaired. Following an analysis and subsequent repairs, Exelon claimed to have rectified the Quad Cities Unit 2 problems only to be forced in 2003 to once again attempt operation at a reduced power level when vibrationally induced steam dryer moisture carryover became excessive. Following this second attempt to operate the reactor at a reduced power level, pieces of the dryer as large as a man broke off and damaged nuclear power safety related components, and a second unplanned extended outage ensued. Once again, vibration was determined to be the cause of the gross failure and another unplanned and forced outage. Finally, following years of analysis and two damaged steam dryers, Quad Cities made major piping modifications that are alleged to have eliminated harmonic frequencies, prevented further component damage, and 13 allowed Unit 2 to eventually return to full power production.

• A second example of a failed attempt to reduce power to solve vibrationally induced resonance frequency problems occurred at the Susquehanna nuclear plant in Pennsylvania. During the mid 1990s, a vibrationally induced failure in the jet pump sensing lines occurred at Susquehanna. This failure was attributed to the vane passing frequency from the recirculation pumps causing harmonic vibration of the lines. Like Quad Cities, Susquehanna attempted to implement a power reduction in order to minimize the harmonic vibrations. Unfortunately, the resonant vibration issues continued to damage systems after the power was reduced thereby forcing an unplanned outage and extensive modifications and repairs.14

12 of 13 Conclusion In conclusion, the NRC has stated that nuclear power plants like San Onofre cannot risk compromising critical safety systems and possible radiological contamination in an effort to return to operation before a thorough root cause analysis, modifications, and subsequent repairs are adequately reviewed by the NRC and implemented. Historical evidence has proven that power reductions do not solve underlying and serious degradation problems, resonance frequency issues. Rather, power reductions can significantly increase the risk of unplanned, forced outages during times of peak demand and can cause significant risk to public health in the event of a single tube rupture or a series of ruptures if the main steam line were to break.

Finally, if a steam-line accident were to occur, vibrationally induced tube damage at San Onofre could cause an inordinate amount of radioactivity to be released outside of the containment system compromising public health and safety in one of the most heavily populated areas in the entire United States.

Note.

This report represents the opinion of Fairewinds. Industir insiders, who have had lengthy careers in steam generatordesign, fabrication,and operation,and who have chosen to remain anonymous, have assistedFairewindswith researchfor this report, but are not responsiblefor its content.

13 of 13 Endnotes 1 Improving Like-For-Like Replacement Steam Generators by Boguslaw Olech of Southern California Edison and Tomouki Inoue of Mitsubishi Heavy Industries, Nuclear Engineering International, January 2012, page 36-38. http://edition.pagesuite-professional.co.uk/launch.aspx?referral=other&pnum=36&refresh=KOs3 a2 1GRq61%20&EID=af75ecb 1-5b23-49be-9dd6-d806f2e9b7b5&skip=&p=36 2 Ibid.

3 Vibration source: March 27, 2012, Confirmatory Action Letter - San Onofre Nuclear Generating Station, Units 2 And 3, Commitments To Address Steam Generator Tube Degradation 4 See, 1.187-A-1, ibid, ttp://pbadupws.nrc.gov/docs/ML0037/ML003759710.pdf 5 Ibid.

6 Improving Like-For-Like Replacement Steam Generators by Boguslaw Olech of Southern California Edison and Tomouki Inoue of Mitsubishi Heavy Industries, Nuclear Engineering International, January 2012, page 39. This article was based on a paper published at ICAPP 2011, 2-5 May 2011, Nice, France, paper 11330. Boguslaw Olech, P.E., South8fn California Edison Company, 14300 Mesa Rd., San Clemente, CA 92674, USA, Email: bob.olech@sce.com.

Tomoyuki Inoue, Mitsubishi Heavy Industries Ltd. (MHt), 1-1 Wadasaki-cho 1-Chome, HyogoKu, Kobe, Japan 652 8585, Email: tomoyukiJnoue@mhi.co.jp.

The authors wish to acknowledge all Edison and MHI personnel involved in the SONGS steam generator replacement project for their efforts to make this project a success.

7 This statement is based upon Fairewinds analysis and confirmed by two independent industry steam generator experts who wish to remain anonymous.

8 http://westinghousenuclear.com/Products & Services/docs/flysheets/NS-ES-0073.pdf 9 Figure 13-7 http://www.kntc.re.kr/openlec/nuc/NPRT/module2/module2 6/2 6.htm 10 With the Mitsubishi design the top of the U-tubes are almost dry in some regions. Fairewinds research and four independent industry experts, who wish to remain anonymous, substantiate this statement.

"Damping [dam-ping] noun Physics.

1. a decreasing of the amplitude of an electrical or mechanical wave.

2. an energy-absorbing mechanism or resistance circuit causing this decrease.

3. a reduction in the amplitude of an oscillation or vibration as a result of energy being dissipated as heat.

http://dictionary.reference.com/browse/damping?s=t 12 http://www.songscommunity.com/docs/SONGS 19_CALFactSheet_050712.pdf 13http://pbadupws.nrc.gov/docs/ML0609/ML060960338.pdf 14 http://www.nrc.gov/reading-rm/doc-collections/gen-comm!info-notices/1 995/in95016.html