ML13144A824

Comment (304) of Jessica Olson on Behalf of Ayres Law Group Opposing the Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, U

ML13144A824
Person / Time
Site:	San Onofre
Issue date:	05/16/2013
From:	Olson J, Ayres R, Gladd K H Ayres Law Group, Friends of the Earth, Natural Resources Defense Council
To:	Rules, Announcements, and Directives Branch
References
78FR22576 00304, NRC-2013-0070
Download: ML13144A824 (181)

Text

Paae 1 of 1 RULES-AWD-DIRECTIVES BRANCH As of: May 17, 2013 Received:

May 16, 2013 PUBLIC SUBM ISSIONMAY I7 AM 10: 53 StatusPendingPost Tracking No. ljx-85d7-jdby Comments Due: May 16, 2013 Submission Type: Web Docket: NRC-2013-0070 RECEIVED Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination Comment On: NRC-2013-0070-0001 Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, Unit 2 Document:

NRC-201 3-0070-DRAFT-0212 Comment on FR Doc # 2013-08888 Submitter Information Name: Jessica Olson Address: 1707 L St, NW Suite 850 Washington, DC, 20036 Organization:

Ayres Law Group General Comment Please see attached 3 files.Attachments Part 1 of 3-FoE & NRDC Comments on Proposed NSHC Det for LAR 263 Part 2 of 3-FoE & NRDC Comments on Proposed NSHC Det for LAR 263 Part 3 of 3-FoE & NRDC Comments on Proposed NSHC Det for LAR 263 Part 2 SUNSI Review Complete Template = ADM -013 E-RIDS= ADM-03 Add= B. Benney (bjb)https://www.fdms.gov/fdms-web-agency/component/contentstreamer?objectld=09000064812e838b&for...

05/17/2013 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE NRC STAFF In the Matter of )) Docket ID NRC-2013-0070

)SOUTHERN CALIFORNIA EDISON CO. ))(San Onofre Nuclear Generating Station, )Units 2 and 3) ) May 16, 2013 FRIENDS OF THE EARTH'S AND NATURAL RESOURCES DEFENSE COUNCIL'S COMMENTS IN OPPOSITION TO PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION I. INTRODUCTION Friends of the Earth (FoE) and Natural Resources Defense Council (NRDC) submit the following comments in opposition to the recently proposed no significant hazards consideration determination regarding a license amendment request that would modify the terms of San Onofre Nuclear Generating Station Unit 2's operating license. FoE and NRDC assert that there should be a hearing prior to any Nuclear Regulatory Commission (NRC) decision on the proposed license amendment.

The proposed amendment would allow operation at no more than 70%Rated Thermal Power (RTP) (or 2406.6 megawatts thermal) for the duration of Cycle 17.1 2 To analyze the proposed no significant hazards consideration determination, Friends of the Earth has enlisted the assistance of four experts with substantial and relevant experience related to the issues presented by the proposed no significant hazards consideration finding: 1 Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generation Station, Unit 2, 78 Fed. Reg. 22576 (April 16, 2013)("Notice of License Amendment Proposal").

2 Because Edison plans to operate Unit 2 intermittently during Cycle 17, this operational period could last between 22 and 24 months.

• Nuclear engineer and former NRC Staff, Dr. Joram Hopenfeld:

During his time at the NRC, Dr. Hopenfeld's work led to the creation of a Steam Generator Action Plan to address safety issues in steam generators.

He has extensive experience with steam generator tube failure." The Honorable Victor Gilinsky, former Commissioner of the United States Nuclear Regulatory Commission:

During Dr. Gilinsky's NRC tenure Congress passed the Sholly amendment 3 and the Commission first interpreted and applied the amendment.

• Nuclear engineer John Large, of Large & Associates:

Mr. Large is a Consulting Engineer and Chartered Engineer, who was a full-time member of the Academic Staff at Brunel University for over 25 years. Mr. Large frequently provides expert evidence on nuclear systems failures and other technical issues in the U.K.Crown and Civil Courts.* Mr. Arnold Gundersen, a nuclear engineer:

Mr. Gundersen is a former licensed nuclear reactor operator and Chief Engineer at Fairewinds Associates.

Public comment on the proposed no significant hazards consideration determination is made difficult by the lack of information associated with both the specific license amendment request and the Staff's proposal.

In order to respond to the proposed determination, Friends of the Earth's experts reviewed studies submitted by SCE in the parallel Confirmatory Action Letter (CAL) proceeding supporting a proposal to restart Unit 2 at 70% of power.4 Incorporated into the Atomic Energy Act at 42 U.S.C. § 2239(a)(2)(A).

4 SCE submitted the operational assessments reviewed here in response to a March 27, 2012 CAL. See Letter from Elmo E. Collins, Regional Administrator, Region IV, Nuclear Regulatory Commission, to Peter T. Dietrich, Senior Vice President

& Chief Nuclear Officer, Southern California Edison, Confirmatory Action Letter -San Onofre Nuclear Generating Station, Units 2 and 3, Commitments to Address Steam Generator Tube Degradation, CAL 4 001 (Mar. 27, 2012), available at ADAMS Accession No. ML12087A323.

2 Having reviewed these submissions of SCE in support of the proposal to allow operation of Unit 2 at 70% of power, and the analyses of Mr. Large, Dr. Hopenfeld, Dr. Gilinsky, Mr.Gundersen, and consistent with LBP-07-13, the May 13, 2013 opinion of the Atomic Safety &Licensing Board (ASLB), 5 discussed below, FoE and NRDC request that the proposed no significant hazards consideration determination should be withdrawn because (1) the Staff s proposal exceeds the authority granted to it by the Sholly amendment; (2) the licensee's application of the criteria under 10 C.F.R. § 50.92, as adopted by the NRC Staff, does not justify a finding of no significant hazards consideration; and (3) the Staff have not performed an environmental review of the proposed finding and license amendment as required by the National Environmental Policy Act (NEPA), and the proposed actions do not satisfy criteria for a categorical exemption from NEPA review, provided at 10 C.F.R. § 51.22(c)(9)(i).

FoE's and NRDC's analysis is supported by the ASLB's May 13, 2013 decision holding that the licensee's restart plan, which proposes operation in conformance with the proposed license amendment, constitutes a defacto license amendment proceeding.

The decision is based not only on the need to revise technical specification 5.5.2.1 1.b.1 through the license amendment proposed in the present action, but also on the need to revise the Updated Final Safety Analysis Report (UFSAR), which currently fails to account for in-plane fluid elastic instability (FEI)-one of the main defects present in the replacement steam generators.

Approving the requested license amendment-temporary operation at 70% of power-would authorize the licensee to operate the plant with an outdated and insufficient UFSAR; another reason why a finding of no significant hazards consideration is inappropriate in this case.5 Southern California Edison Co. (San Onofre Nuclear Generating Station, Units 2 and 3), LBP-13-07 (May 13, 2013) ("ASLB Order"), appended to these comments as Attachment 6.3 The Staff has ignored the language and legislative history of the Sholly Amendment as recited by the Ninth Circuit in the Mothers for Peace case, discussed below. As the court held, the amendment gives the NRC Staff the right to screen out trivial changes that could not possibly affect safety. If a safety issue is identified, however, then the Staff must legally conclude that a significant hazards consideration exists, and must refer the issue(s) to an Atomic Safety and Licensing Board for a hearing prior to the decision on the proposed license amendment.

The court ruled that the NRC Staff should "not resolve doubtful cases with a finding of no significant hazards consideration." The court added, from the legislative history, that the NRC Staff should not "prejudge the merits" of the issues by a proposed license amendment.

The Staff does not determine whether a significant hazard exists; that is for the ASLB to determine.

Thus the Staff's proper role under the Sholly amendment is essentially ministerial:

it determines whether a significant hazard consideration exists, and if so, asks the ASLB to determine whether the proposed amendment creates a significant hazard. Consistent with the Mothers for Peace case, the Staff must refer SCE's proposed license amendment to an ASLB: A probability, not just a possibility, of significant hazards permeates this case.It follows that the Staff has disregarded its appropriate role in proposing a no significant hazard consideration finding in this instance.

The safety issues presented here are real, not trivial, as acknowledged in the ASLB panel decision of May 13, 2013. Under applicable law, the Staff must withdraw the proposal and refer the proposed license amendment to the ASLB for an adjudicatory hearing, as requested by Friends of the Earth and NRDC in these comments, before a decision on the proposed amendment can be made.4 II. FACTUAL BACKGROUND

a. The Shut Down of Units 2 and 3 On January 31, 2012, San Onofre experienced a steam generator tube rupture in Unit 3 that resulted in the release of radioactive material into the environment.

The licensee, Southern California Edison Company (SCE or "Edison")

also discovered excessive wear in the Unit 2 replacement steam generators.

The unit was offline for a refueling outage. Subsequently, untimely degradation of the walls of many tubes was discovered in the replacement steam generators, which had been in operation for eleven months in Unit 3 and less than two years in Unit 2.On March 23, 2012, SCE submitted a description of the steam generator problems and its commitments to address the issues at Units 2 and 3, which were formalized in a CAL to SCE on March 27, 2012.None of the investigations conducted to date have determined the root cause of the premature and extensive tube degradation in the replacement steam generators.

Lacking such understanding, SCE has not proposed any action to actually fix the problems of either Unit 2 or Unit 3. Rather, SCE has proposed a restart plan based on substantially reduced operational limits that it has asserted is safe.b. Southern California Edison's Replacement of the Steam Generators at San Onofre Units 2 and 3 In 2010 and 2011, SCE replaced the original steam generators in Unit 2 and Unit 3, which had operated for 28 years, with ones constructed by a different manufacturer, Mitsubishi Heavy Industries (MHI). The new design differed from the original in significant ways. SCE requested MHI, for example, change the design by adding 377 more tubes, remove the stay cylinder supporting the tube sheet, and replace the "egg crate" tube support with a broached 5 design, among other alterations.

6 SCE convinced itself that the replacement steam generators were a "like for like" replacement for the old ones and did not seek a license amendment for these changes. Thus, the impact of these changes on safe operation of the plant has not previously been evaluated by the NRC.c. Extent of Tube Degradation in the Steam Generators in Units 2 and 3 Both units show indications of extensive tube wear after fewer than two years of operation.

The tube degradation in each unit is unlike, in both mechanism and extent, tube wear in other replacement steam generators in other U.S. plants at the same stage of their useful lives: 0 San Onofre Unit 2 has 1595 degraded tubes; Unit 3 has 1806;* Unit 2 has 4721 tubal wear indications; Unit 3 has 10,284;0 Unit 2 has 510 tubes plugged after one cycle of operation of the replacement steam generators; Unit 3 has 807;0 SCE and NRC have reported that 9% of the tubes in Unit 3 steam generators have greater than 10% through-wall wear indications; in Unit 2, 12% of the tubes show such wear.Tube wear of this magnitude after such an abbreviated period of operation is unprecedented

.d. SCE's Previous Assessments of Operation at 70% of Rated Thermal Power As part of its response in the CAL process, SCE submitted to NRC Staff numerous operational assessments by its consultants.

While agreeing that the proximate cause of wear of the replacement steam generator tubes was excessive vibration, SCE and its consultants have not identified a root cause of the excessive vibration causing the premature and extensive tube wear, 6 For a more detailed description of the changes, see Declaration of Arnold Gundersen (May 31, 2012) (Originally submitted to the NRC as an attachment to a June 18, 2012 Petition to Intervene by Friends of the Earth) at ¶¶ 22-23 and MHI Root Cause Analysis and Supplemental Technical Evaluation Report at pp. 47-48, appended to these comments as Attachment 4 and 5, respectively.

7 See ASLB Order at p. 25 (citing SCE's statement to that effect).6 as described in Table 6-1 of SCE's Unit 2 Return to Service Report. In fact, Edison's own consultants disagree with one another on the mechanistic cause of the tube wear. SCE's response to the CAL includes an analysis of tube-to-tube wear and argues that the cause of such wear is FEI. However, this response does not identify the root cause that produced the FEI or acknowledge other thermal hydraulic forces at work in the steam generators.

Without knowing the root cause, as declarant nuclear engineer John Large asserts, it is not possible to determine whether the steam generators can be safely operated in their current condition.

SCE's response to the CAL includes a proposal to restart Unit 2 at no more than 70%power for 150 cumulative days, at which time SCE promises to shut down the reactor and inspect the tube wear. The current proposed license amendment is required because Edison has failed to demonstrate to the NRC that it can meet the terms of the existing license requiring a demonstration of tube structural integrity at 100% of power. This point alone makes it impossible for the NRC to reach a determination that Edison's proposed license amendment presents no significant hazards consideration.

Edison's response to the CAL is nearly identical to the license amendment request in the present instance:

to modify Unit 2's license to limit maximum power for operation at 70% for Cycle 17. SCE hired AREVA NP, Westinghouse Electric Company LLC, and Intertek/APTECH to provide operational assessments (OAs) of this proposal.

MHI, the manufacturer of the replacement steam generators, also examined the unprecedented tube wear and present condition of the tubes.These assessments, which are included in SCE's response to the CAL, not only demonstrate clearly that there are significant hazards to be considered before ruling on the license amendment request, but they also suffer from important omissions.

The studies focus on 7 tube-to-tube wear as the threat to tube rupture, incorrectly assuming that this mode of wear will outpace all other wear modes. They do not analyze the potential safety effects of further degradation of the tubes in Unit 2 that are vibrating against the retainer bars and tube restraint structures; nor do the OAs address extent and impact of metal fatigue on the damaged tubes'structural integrity.

The OAs point to different mechanical interactions resulting from FEI and random fluid excitation sources as the causes of the tube degradation, but none determined the root cause of the in-plane tube motion excitation forces, which appear to be unique to San Onofre's replacement steam generators.

In addition to failing to identify the root cause of the tube degradation or to recognize the different modes of wear, SCE and its consultants also failed to agree on the projected length of time before a tube burst may occur, even by their own inadequate analysis.

Estimates vary from six months to sixteen months of operation at 70% RTP, indicating that the underlying risk analysis is fundamentally flawed.8 In other words, SCE cannot say with confidence that a tube burst is unlikely within the time frame of Cycle 17, which is 22-24 months, at 70% of power.What both assessments say is that the tubes will deteriorate at a pace that will cause steam generator failure, in the best-case scenario within 16 months, and the worst case 6 months-a mere month more than the period SCE proposes to run the plant, were its license amendment to be approved.The consultants' estimates are remarkable for two reasons. First, neither projects the unit can be run safely, even at reduced power, for more than 16 months-even though the original expectation of SCE and the designers was that they would last for three decades or more.Second, the two estimates differ by a factor of nearly three. The fact that each of the consultants 8 Declaration of Mr. John Large, May 16, 2013 ("Large Decl.") at ¶ 8.5.14, appended to these comments as Attachment

2.8 relied

upon by SCE project significantly different periods of time before reaching and surpassing this safety threshold "shows that the underlying data and methodology of the predictions is fundamentally flawed." 9 In view of this "uncertainty and unreliability" Mr. Large concludes that"little assurance can be placed with SCE's confidence that its Cycle 17 ... will pass without encountering a significant increase in the risk of tube failure."'1 0 Moreover, the estimates by SCE's consultants do not account for the fact that "a certain percentage of steam generator tubes have used up their entire or a large fraction of their allowable fatigue life during cycle 16."'"1 Fatigued tubes present a more significant risk than tubes degraded by stress corrosion cracking because tube failure caused by fretting fatigue will result a sudden burst and "proceed rapidly to its maximum as it happened a North Anna (NRC Bulletin 88-02)."'12 These facts demonstrate the absurdity of the Staff's proposal to conclude that operation of San Onofre as described in the proposed license amendment presents no significant hazards consideration.

SCE has received assessments on the issue of various tube wear modes by AREVA, the other consultants, and MHI, but SCE did not include important aspects of these assessments in its response to the CAL. For example, SCE have chosen not to emphasize or explain an analysis by MHI, which found that tube wear from contact between the tubes and anti-vibrations bars in Unit 2's replacement steam generators arose in areas of the tube bundle where FEI was inactive, suggesting that the wear was caused by turbulent flow forces that may persist even at the proposed power level of 70% intended to suppress the FEI. In light of these facts, the NRC cannot properly find that no significant hazards consideration is raised by 9 Large Decl. at ¶ 8.5.14.10 Large Decl. at¶ 8.5.15.I Declaration of Dr. Joram Hopenfeld, May 16, 2013 ("Hopenfeld Deci.") at p. 7.12 Hopenfeld Decl. at p. 8.9 amending the license to allow the unit to be restarted.

What is clear is that the proposed amendment does present significant hazards consideration that require airing in a public adjudicatory hearing bejbre the license amendment can be granted and the unit can be allowed to operate again.III. COMMENTS We note at the outset that the record for this proposed no significant hazards consideration determination is perplexingly thin. The docket contains only the Federal Register notice of the proposed finding (along with a few comments from citizens).

To address the significant hazards consideration involved in the proposal to operate the damaged replacement steam generators at 70% of power, FoE's experts were required to review the technical analyses in the public record in another proceeding (i.e., the operational assessments provided by SCE in response to the March 27, 2012 Confirmatory Action Letter).The Staff has not placed any analysis of the § 50.92 factors into the record.1 3 They have ignored the fact that these steam generators are so badly damaged that the licensee has not proposed restarting Unit 3 and concedes that the damaging forces will continue to degrade the steam generator tubes in Unit 2 to the point of failure.In reality, the Staff's proposal ignores the fact that what is at stake is the licensing of badly damaged steam generators that Edison concedes will continue to be further damaged by operation.

It fails even to attempt to explain how operating Unit 2 can pass the rigid standards of 10 C.F.R. § 50.92 and it ignores the Sholly amendment by treating as routine the safe operation of a plant that even the Staff admits is not safe to run at full power.13 See Docket ID No. NRC-2013-0070.10 In a related proceeding, an ASLB convened by the Commission recently determined that SCE's proposal to restart San Onofre Unit 2 at 70% ofpower on an experimental basis is a de facto license amendment proceeding, which requires "rigorous NRC Staff review appropriate for a licensing action."'1 4 The ASLB found that SCE's proposal would allow Unit 2 to operate outside the current licensing basis of the plant, not only because a maximum operating level of 70% of power does not comply with Technical Specification 5.5.2.1 i.b. I, but also because restarting the steam generators in their current degraded condition is outside the bounds of the safety analyses that form the licensing basis for the plant (the UFSAR). Having found that"there is a dearth of applicable experiential data available for in-plane vibrational motion, because, as conceded by SCE, 'tube-to-tube wear due to in-plane [fluid elastic instability]

ha[s]not been previously experienced in U-tube steam generators,'

,,5 the Board held that prior to restart SCE is required to submit a license amendment that properly updates the FSAR to include a full assessment of the effects of in-plane fluid elastic instability:'

6 We conclude that until the tube degradation mechanism is fully understood, until reasonable assurance of safe operation of the replacement steam generators is demonstrated, and until there has been a rigorous NRC Staff review appropriate for a licensing action, the operation of Unit 2 would be outside the scope of its operating license because the replacement steam generator design must be considered to be inconsistent with the steam generator design specifications assumed in the FSAR and supporting analysis.

In short, the start-up of Unit 2 pursuant to the CAL process would transform that process into a de facto license amendment proceeding by allowing steam generator operation with a tube degradation mechanism not considered in the FSAR -i.e., in-plane vibrations due to fluid elastic instability.'

7 The Staff's proposal to find that the license amendment request presents no significant hazards consideration would authorize SCE to restart Unit 2 at 70% of power without updating 14 ASLB Order at 32.15 ASLB Order at 34, n. 54.16 ASLB Order at 32.17 ASLB Order at 32, 33 (internal citations omitted).11 the FSAR. The proposal, if made final, would thus contravene the ASLB's order. More fundamentally, the three impartial experts who constitute the panel have confirmed the views of FoE's experts that the proposed license amendment and restart plan is an experiment that raises significant safety issues in all three of the regulatory criteria that must be satisfied in order to make a finding of no significant hazards consideration.

For these reasons, the proposed no significant hazards consideration determination should be withdrawn.

Moreover, (1) the ASLB's conclusions mirror the analyses by FoE's experts that the proposed license amendment fails to meet any of the criteria required for a no significant hazards consideration finding, and (2) the Staff's proposed determination violates the terms of the Sholly Amendment.

a. The Proposed Finding of No Significant Hazards Consideration Exceeds the Authority of the NRC Staff Section 189a of the Atomic Energy Act (AEA) requires that, if requested, a public hearing must be held prior to the issuance of any license or license amendment before an Atomic Safety and Licensing Board.1 8 The "Sholly" amendment, 42 U.S.C. § 2239(a)(2)(A), provides a limited exception to this general rule. The NRC staff may issue a license amendment before a hearing only if it finds the license amendment'raises no significant hazards consideration.

The relevant regulations are found at 10 C.F.R. § 50.92(c), which we describe in detail below. The legislative history of the Sholly amendment makes clear that it is limited to only the most routine license amendments, which may be granted prior to the hearing guaranteed by the AEA.19 Under NRC regulation 10 C.F.R. § 50.92, the NRC Staff may not determine that a proposed license amendment raises no significant hazards consideration, and thus must refer the 18 42 U.S.C. § 2239(a)(1)(A).

"9 See Declaration of Victor Gilinsky, May 16, 2013 ("Gilinsky Decl.") at ¶ 5, appended to these comments as Attachment 3 ("Congress permitted

[no significant hazards consideration]

determinations in routine cases that obviously had no or essentially no safety significance, but not otherwise.").

12 matter to an ASLB for resolution before the license amendment may be issued, whenever a proposed license amendment will: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.The proposal to approve SCE's license amendment application is fundamentally inconsistent with the purpose of the Sholly amendment to the Atomic Energy Act, which is the authority for 10 C.F.R. § 50.92. The Sholly exception to the rule that proposed license amendments should not be approved prior to a public hearing before an ASLB was intended to be a narrow one, to be used to avoid delay for routine amendments no one would suggest posed significant hazards considerations, such as replacing a gauge.2 0 The three criteria of 10 C.F.R. § 50.92 are therefore to be read narrowly.

Unless the three conditions are met unequivocally, the NRC should grant a hearing before an ASLB prior to deciding whether to approve a license amendment.

In short, if a proposed license amendment presents a significant hazards consideration-that is, if a comment identifies an issue involving significant hazards-then the matter must be referred to an ASLB for resolution before the proposed license amendment may be considered.

SCE's proposed license amendment could not be further from the kind of change Congress sought to exempt from a prior hearing through the Sholly amendment:

as demonstrated by FoE's experts and the recent opinion of the ASLB, the prospect of restarting San Onofre Unit 2 with damaged and unrepaired steam generators presents significant new and ill-understood safety risks, not routine changes to technical specifications such as updated inspection routines or new gaskets or gauges.20 Gilinksy Decl. at ¶ 3.13 The proposed no significant hazards consideration determination in this case must be withdrawn because, rather than determining whether a significant hazards consideration was present requiring referral to an ASLB for a hearing prior to a decision on the license amendment, the NRC Staff preempted the Atomic Safety and Licensing Board, eschewing its proper role as regulator and instead assuming the judicial role allocated to the ASLB under the AEA, including the Sholly amendment.

Rather than identify and refer the obvious significant hazards considerations involved in SCE's application for a license amendment, the NRC Staff simply adopted the licensee's evaluation of the merits of the license amendment request, apparently without question.This is exactly what the Ninth Circuit held invalid in San Luis Obispo Mothers for Peace.21 Under the Mothersfor Peace ruling, the purpose of the no significant hazards consideration determination by the Staff is simply to identify whether there are new or increased risk considerations that should be reviewed by an ASLB before the proposed license amendment can be issued. Whether the new or increased risks are acceptable is a decision for the ASLB, to be decided in a hearing held prior to deciding whether to approve the proposed license amendment.

Moreover, the ASLB's recent decision, noted above, on the risks presented by SCE's proposed temporary operation at 70% of power found that, The unprecedented extent of tube wear and failures that SCE experienced in the SONGS Unit 3 replacement steam generators reveal that these steam generators have serious design and operational issues, placing them beyond the envelope of experience with U-tube steam generators.. .Although the Unit 2 steam generators did not experience the accelerated and extensive tube-to-tube wear suffered in the Unit 3 steam generators, they nevertheless are the identical design as those in Unit 3 and they operate under similar conditions.

2 2 21 San Luis Obispo Mothers for Peace v. U.S Nuclear Regulatory Commission, 799 F.2d 1268 (9th Cir. 1986).22 ASLB Order at p. 25.14 In these circumstances, a finding of no significant hazards consideration is wholly inappropriate and exceeds the authority of the NRC under the AEA.Thus, the NRC should convene an ASLB prior to making a decision whether to issue the license amendment to examine the significant safety issues posed by SCE's proposed license amendment to allow operation of the damaged replacement steam generators at 70% power for Cycle 17 (22-24 months). Such an adjudicatory hearing would provide reassurance to the people of Southern California and would be consistent with the Commission's announced policy of transparency.

The current attempts to misuse 10 C.F.R. § 50.92 to exclude public participation can only exacerbate public distrust for the NRC and of the safety of the San Onofre plant, whatever decision is ultimately made.b. The Proposed License Amendment Presents New and/or Increased Risks That Endanger Public Health and Safety If it actually considered the criteria of 10 C.F.R. § 50.92, the Staff could not determine that the proposed license amendment for San Onofre entails no significant hazards consideration.

If the proposed change in the license fails to meet any one of the three criteria in 10 C.F.R. § 50.92, the NRC must withdraw the proposed no significant hazards consideration determination.

As demonstrated in the technical analyses appended to these comments, and by the ASLB's recent decision on San Onofre, the proposed amendment does not-satisfy any of the three criteria.To assess whether the change proposed by SCE creates a significant hazards consideration, the appropriate comparison is between the operation of the unit with undamaged steam generators as assumed in SCE's current license, on the one hand, and the operation at 70%of power with damaged steam generators that Edison now proposes.

SCE's rationale for concluding that no significant hazards consideration is presented is apparently based on 15 comparing operation with undamaged tubes at 100% and 70%, completely ignoring the current highly-damaged state of the steam generators in Unit 2. As Dr. Hopenfeld states, SCE's evaluation of the § 50.92 criteria "is based on the presumption that change in power level can be discussed without giving any considerations to the physical conditions of the tubes before and after the change." 2 3 The NRC cannot, despite its best efforts, ignore the events of the past 16.5 months.Major defects causing unprecedented tube wear have been discovered in the replacement steam generators at San Onofre, and while the mechanical force that inflicted the wear has been identified as primarily in-plane FEI, neither the NRC nor the licensee has yet determined the root cause of the FEI, let alone a remedy for it. Instead, SCE and the NRC propose to simply restart Unit 2 and operate it at reduced power for one cycle as an experiment to see whether the plant can be run longer at that reduced rate. One could not possibly conclude that such a proposal does not at least raise "significant hazards" considerations that require further scrutiny in a hearing to decide whether the additional risk of exposing Californians to radiation are acceptable.

The Staff simply ignores the fact that Unit 2's replacement steam generators have already demonstrated design flaws in components and systems critical to the safety of San Onofre Unit 2.i. The Proposed Finding of No Significant Hazards Consideration Should Be Withdrawn Because the Proposed License Amendment Would Involve a Significant Increase in the Probability or Consequence of an Accident Previously Evaluated.

Staff addresses the first criterion of 10 C.F.R. § 50.92 by simply restating SCE's analysis, which concludes that the proposed license amendment would not involve a significant increase in the probability or consequence of an accident previously evaluated "because there is no 23 Hopenfeld Decl. at p. 6.16 adverse effect on plant operations or plant conditions." 2 4 SCE relies on its response to Requests for Additional Information (RAIs) 11-14 as its basis for this assertion.

2 5 SCE, however, fails to make the appropriate comparison when applying this first criterion.

SCE's apparent position is that the relevant comparison is between operation of fully functional undamaged steam generators originally licensed to run 100% of power and operation of those same steam generators at 70% of power. That characterization is incorrect.

The proper consideration is whether operating at 70% of power with defective, damaged, and unrepaired steam generators involves a significant increase in the probability or consequence of an accident previously evaluated, as compared to the risk of operating at 100% of power with fully functional, undamaged steam generators.

In this context, operating the steam generators in their present condition at 70% of power creates a significant increase in the probability of a release of radioactivity and in the consequences

-exposure of potentially millions of people to increased radioactivity.

The three impartial experts who wrote the ASLB's recent decision on San Onofre found that operating the replacement steam generators at 70% would significantly increase the probability and consequences of a previously analyzed accident.26 For example, the replacement steam generators can no longer meet 10 C.F.R. Part 50, App. A -General Design Criterion (GDC) 14 (Reactor Coolant Pressure Boundary), which requires "an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture." SCE's own tube-to-tube wear assessment, as the ASLB order notes, shows that "one unstable tube can drive its neighbor into instability through repeated impact events." 2 7 Given this condition, there is no 24 Notice of License Amendment Proposal at p. 22577.25 Notice of License Amendment Proposal at p. 22577.26 ASLB Order at p. 27.27 ASLB Order at p. 27.17 longer "an extremely low probability" of the kind of tube failure GDC 14 is meant to guard against.Nuclear engineers Mr. Large and Dr. Hopenfeld show in the attached declarations that the proposed amendment would involve a significant increase in the probability or consequence of an accident previously evaluated.

Mr. Large explains that the excitation forces present in the steam generators exist due to pressure and temperature conditions that will not be affected by reducing the power from 100% to 70%.28 Thus, contrary to the assertions of SCE, operating Unit 2 at 70% of power during Cycle 17 would not reduce the forces exerted on the tubes during Cycle 16 that caused the unprecedented rapid tube wear and deterioration.

2 9 Both of SCE's operational assessments agree that the damage will continue at an unprecedented pace, differing only between 6 months and 16 months as the remaining life-expectancy of the Unit. Even at 70% of power, large numbers of tubes in the replacement steam generators will continue to wear and degrade and, as a consequence, significantly increase the probability of tube rupture.Dr. Hopenfeld asserts that the probability and consequences of a previously considered accident are significantly increased because, in addition to the fact that operating at 70% of power will not reduce the excitation forces that cause tube wear, SCE also failed to take into account metal fatigue caused by fretting, which is brought on by the FEI-induced vibration.

Tubes in Unit 2's steam generators used up a large fraction, if not all, of their allowable "fatigue life" during the last cycle of operation, Cycle 16.30 Dr. Hopenfeld asserts: The number of tubes which are susceptible to rupture by fatigue during a given accident scenario must be known if one is required to predict accident consequences.

Until this is done the present pressure based burst performance criteria cannot be used as a reliable 28 Large Dec. at ¶8.5.3.29 Large Decl. at ¶ 8.5.5.30 Hopenfeld Dec. at p. 7.18 indicator of risk. As a result it must be conservatively concluded that allowing Unit 2 to operate at any power level would significantly increase the consequences of the accidents, which were evaluated by SCE and were described in the UFSAR.3 a SCE and its consultants have inspected the steam generators for tube surface wear and tube wall thickness but have failed to account for metal fatigue, which cannot be discerned by inspection.

Technical Specification Task Force (TSTF) 449 requires SCE to evaluate additional loads on the tubes that could contribute to burst or collapse, even if they cannot be physically measured.3 2 SCE's analysis ignores the increased probability or consequences of an accident contributed to by metal fatigue in the tubes of the steam generators.

Tube fatigue increases the probability of an accident.

It also increases the consequences, because tube failure owing to metal fatigue happens more suddenly than failure owing to stress corrosion cracking (SCC). A tube failure from fatigue, such as that experienced at the North Anna Generating Station Unit 1 on July 15, 1987, occurs suddenly and quickly.3 3 In the event of a main steam line break, for example, accompanied by the rupture of five or more fatigue-weakened tubes, the operator's inability to control the loss of coolant rapidly enough would lead to a significant increase in the probability of uncovering the core, with major increases in the consequences of a previously evaluated accident, including the exposure of millions of Californians to radiation.

3 4 Dr. Hopenfeld therefore concludes that restarting the plant for another cycle would place Unit 2 outside of the bounds of accidents evaluated in the updated final safety analysis (UFSAR)report by significantly increasing the probability and consequences of a main steam line break 3 Hopenfeld Decl. at p. 8.3 Hopenfeld Decl. at p. 7-8.33 Hopenfeld Decd. at p. 8.34 Hopenfeld Decd. at p. 9, 33.19 (MSLB) accident.3 5 Similarly, Mr. Large found that a single tube burst caused by an MSLB that damages the fuel core could result in severe consequences beyond those considered in the UFSAR.3 6 The NRC's proposed findihg of no significant hazards consideration addresses none of the issues identified by Friends of the Earth's experts, as summarized above. Thus, the proposed finding must be withdrawn and a hearing on the proposed license amendment held before a decision is made on the proposal.ii. The Proposed Finding of No Significant Hazards Consideration Should Be Withdrawn Because the Proposed License Amendment Would Involve the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated.

Significantly, the UFSAR for the original steam generators for SONGS Units 2 and 3 excluded the possibility of in-plane vibrations caused byfluid elastic instability when evaluating the conditions necessary to maintain steam generator tube integrity[,]...

[an assumption that is] demonstrably unjustifiedfor the replacement steam generators.

3 7---ASLB Opinion, May 13, 2013 The NRC's regulations do not allow the Staff to make a no significant hazards consideration determination if it finds that the proposed license amendment would create the possibility of a new or different kind of accident not previously evaluated.

The Staff restates in the Federal Register notice proposing the license amendment the licensee's position that "the proposed changes do not require a change in any plant systems, structures, or components or the method of operating the plant other than to reduce power for the duration of Cycle 17.Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated." 3 8'5 Hopenfeld Deci. at p. 32.36 Large Decl. at ¶ 8.5.17.37 ASLB Order at pp. 31-32.38 Notice of License Amendment Proposal at p. 22577.20 Edison's "therefore" is misplaced:

the conclusion of the second sentence does not follow from the statement in the first. The premise of"no.change" that SCE relies on for this conclusion, however, is erroneous because it ignores the change that shut the plant down more than a year ago: that an abnormally high amount of tube wear has occurred in the replacement steam generators, and, in particular, the unprecedented fretting fatigue caused by massive FEI and its impact on the steam generator tubes.First, the UFSAR does not consider the possibility of accidents caused by tube wear from in-plane FEI because it is based on an assumption that in-plane FEI will not occur. UFSAR section 5.4.2.3.1.3, which analyzes steam generator tube integrity, is therefore inadequate and demonstrates that operation at 70% of power presents new and different kinds of accidents from those previously evaluated.

The ASLB agrees. In its recent opinion on SCE's proposed restart plan under the CAL, the ASLB found that operating the replacement steam generators in their current degraded condition is a test or experiment as described under 10 C.F.R. § 50.59(C)(2).

3 9 By definition then, the proposed license amendment cannot possibly meet the second criterion for a no significant hazards consideration determination.

Operating at 70% for any length of time with the replacement steam generators in their current condition is an experiment, the outcome of which has not been analyzed in the UFSAR.Second, the UFSAR currently considers only accidents resulting from excessive pressure loads, not fretting fatigue. During Cycle 16, the tubes in Unit 2's steam generators experienced fretting previously not experienced in the history of any U.S. steam generator.

4 0 To allow the operation of the steam generators without repairs would, because of this unanalyzed fatigue, 39 ASLB Order at p. 33.40 ASLB Order at p. 25.21 create the possibility of a new or different kind of accident from ally accident previously evaluated.

4 Accidents caused by fretting fatigue are different from accidents caused by stress corrosion cracking (SCC). As described above, unlike SCC, metal fatigue is difficult to detect through in-service inspections, and near or at the end of a tube's fatigue life cracking propagates much more quickly than SCC.4 2 There is no available data correlating field measurements to leakage from fatigued tubes during a design-basis accident.4 3 Thus, any safety analysis that is based on fatigue failures relates to a new and previously unanalyzed accident.4 4 SCE has yet to perform an analysis of probable accidents owing to fretting fatigue failures, which it must do before the proposed license amendment could possibly satisfy the second criterion of 10 C.F.R. §50.92.Specifically, Dr. Hopenfeld discusses five possible accident scenarios owing to fretting fatigue not considered by the existing UFSAR. In other words, the risk of these accidents arises from the fact that the tubes have already been substantially fatigued and will experience further fatigue at 70% operation:

1. Fretting fatigue rupture of a tube in the free span with a relief valve stuck open or a broken header;2. Unplanned closure of an isolation valve, increasing steam pressure abruptly, causing rupture of tubes on the border of exhausting their fatigue life;3. Seismically-induced ruptures of both plugged and unplugged tubes near the end of their fatigue life;41 Large Decl. at ¶ 8.6.26; Hopenfeld Deci. at p. 27-34.42 Hopenfeld Deci. at p. 29.43 Hopenfeld Deci. at p. 4."4 Hopenfeld Decl. at p. 27-34.45 Hopenfeld Deci. at p. 30-32.22

4. Severe accident causing rupture of tubes near the end of their fatigue life; and 5. Main steam line break accident:

in situ tests of tube integrity show only the tendency of tubes to leak on the basis of loss-of-wall-material or weakening by stress corrosion cracks. Fatigue failure would cause propagating circumferential cracks.4 6 Little data is available to assess the safety risks presented by these accidents due to the unprecedented and unique nature and extent of the damage to the tubes in Unit 2's steam generators.

4 7 Dr. Hopenfeld calculates that of the nearly 1100 tubes susceptible to fatigue failure, the probability of only 5 tubes rupturing during Cycle 17 exceeds the NRC's safety goals by a factor of 5.48 Thus, the proposed license amendment involves serious risks that SCE and the NRC have not considered, precluding a finding of no significant safety hazards consideration.

The risks associated with fretting fatigue are serious, and must be evaluated under the TSTF 449.Mr. Large also raises a number of considerations not taken into account by the Staff in its no significant hazards consideration determination.

While Mr. Large's technical analysis is presented in detail at section 8.6 of his attached declaration, the key points are summarized here.Foremost, Mr. Large emphasizes a critical omission in SCE's analysis:

SCE did not adequately consider-despite the evidence of extensive damage to literally hundreds of tubes-the possibility of a multiple tube failure, which would greatly exceed the design basis accident of a single tube burst. When evaluated against the current condition of the steam generators in Unit 2, Mr. Large details a number of situations with the potential for multiple tube failure that were ignored by SCE.The first of these situations is a scenario in which one of the restraining structures (the anti-vibration bars, or "AVBs"), some of which are already significantly worn, physically detach, 46 Hopenfeld Deci. at pp. 30-32.4' Hopenfeld Decl. at p. 33.48 Hopenfeld Decl. at p. 33.23 damaging tubes in the surrounding area.4 9 Since the conditions for such a potential AVB "break up" are possible (including a scenario in which seismically induced loading on the tube bundle could detach a worn-through AVB component), 5° SCE is required to consider the possibility of a worn section of an AVB detaching under various accident scenarios, thereby leading to a multiple tube failure.51 SCE, however, failed to do so.5 2 Notably, this includes SCE's failure to evaluate the seismic loading of the overall tube bundle, taking into account the degraded and defective tubes and components.

5 3 Mr. Large also describes a second accident scenario ignored by SCE in which both pressurized and plugged tubes failed locally, dislodging shrapnel into the tube bundle and thereby creating a pathway for a multiple tube failure.5 4 Mr. Large notes that various mechanisms exist that could lead to this result, including tube surface damage and flaws-scarring that is already present in the tubes but which SCE has not taken into consideration in the UFSAR.5 5 In short, SCE has not accounted for the effect of known mechanisms, such as this scarring, in its analysis of whether the proposed amendment would exceed the allowable stress limits in place to ensure tube integrity.

5 6 As Mr. Large explains, when a tube is subject to certain stresses such as exist here, it is subject to two types of fatigue 5 7 (one of which, fretting fatigue, is discussed at length in the Hopenfeld Declaration).

A situation in which a number of tubes have high levels of fatigue is 49 Large Decl. at TT 8.6.14-8.6.17 (describing a number of situations that could detach portions of a worn AVB and the potential effects of an unrestrained object within the tube bundle).50 Large Decl. at 8.6.16 (stating "[t]here are a number of situations that could challenge and possibly physically detach sections of such a worn down AVB, including seismically induced loading on the tube bundle, the immediate aftermath of a LOCA, and, quite possibly, the dynamic fluid forces triggered by a MSLB").5' Large Decl. at¶ 8.6.17.52 Large Decl. at ¶ 8.6.33.53 Large Decl. at 7 8.6.34 (noting that, moreover, SCE may in fact be required to undertake a seismic response evaluation for the entire replacement steam generator assembly).

54 Large Decl. at T 8.6.18.55 Large Decl. at 77 8.6.19-8.6.22.

56 Large Decl. at ¶ 8.6.22.57 Large Decl. at ¶ 8.6.24.24 more likely to result in multiple tube failure, particularly in the event that fatigue-weakened tubes come into contact with either shrapnel from a single burst tube or the severed tube itself.5 8 Having failed to address even the issue offatigue, SCE could not have evaluated, as it must, the effect of fatigue on a new or different type of accident involving multiple tube failures.Last, and significant for the purpose of evaluating the proposed license amendment, fatigue can run its course to failure within a single operation cycle, 5 9 underscoring the importance of taking this factor into account in accident scenarios.

At base, the fundamental point here is that the damage to the tubes and tube restraint components that occurred during the previous operating cycle at San Onofre Unit 2 was so substantial that the response of these structural components to both normal-as well as possibly adverse-operating conditions have not been accounted for, either in the original design accident cases, nor in the analyses SCE relies upon to justify restarting Unit 2 at 70% of power.Accordingly, SCE's analysis cannot purport to demonstrate that running the plant at 70% power will not involve the possibility of a new or different kind of accident from the types considered previously.

Soberingly, it is precisely this type of accident, such as, for example, a multiple tube failure, that would result in the most severe consequences for public health and safety.6°5' Large Decl. at ¶ 8.6.25.59 Large Decl. at ¶ 8.6.26.60 Large Decl. at ¶ 9.1 (stating that "it is quite feasible that failure of a few defective tubes could trigger a major nuclear plant malfunction that, in itself, provokes the bursting of more degraded or defective tubes creating a very significant radiological release via a primary containment bypass. Also, there is the possibility that a major plant malfunction, such as a N4SLB, could rapidly result in failure of multiple tubes already weakened in a degraded or defective condition").

25 iii. The Proposed Finding of No SignfiLcant Hazards Consideration Should Be Withdrawn Because the Proposed License Amendment Would Involve a Si2nificant Reduction in a Margin of Safety.The assessment in this [Hopenfeld's]

report does not support SCE's position that operation of Unit 2 for five months at 70% power will not affect safely. It is shown that SCE conclusions are not conservative.

Operation of Unit 2 even for one month at any power level would present a safety risk.--- Dr. Joram Hopenfeld NRC's regulations at 10 C.F.R. § 50.92 prevent the Staff from making a finding of no significant hazards consideration where the proposed amendment would involve a significant reduction in a margin of safety. As an initial matter, the ASLB's decision raises a number of serious safety considerations that are evidence that the Staffs position on the no significant hazards consideration is indefensible.

SCE's optimistic Operational Assessment estimates of the margins of safety of operation at 70% of power are not justified by experience, as the ASLB pointed out: SCE's prediction that accelerated tube wear will be precluded by plant operations limited to 70% power is grounded on theory that is not yet supported by actual experience

.... [T]here is a dearth of applicable experiential data available for in-plane vibrational motion, because, as conceded by SCE, "tube-to-tube wear due to in-plant [fluid elastic instability]

ha[s] not been previously experienced in U-tube steam generators."'

6 2 The ASLB further held that the in-plane vibrations caused by FEI were never considered in the UFSAR.6 3 The analyses in the UFSAR provide the basis for operating the plant within an acceptable margin of safety. Restarting a reactor unit with known defects caused by mechanisms (e.g., in-plane FEI) that were not analyzed in the UFSAR thus significantly decreases the margin of safety provided for by the UFSAR.6 Hopenfeld Decl. at p. 10.62 ASLB Order at p. 34, n.54, quoting Edison Answering Brief at 10.63 ASLB opinion at p. 31.26 FoE's experts agree that SCE and the Staff cannot show that SCE's license amendment proposal would maintain the required margin of safety in the current license. Dr. Hopenfeld, for example, concludes that operating Unit 2 at 70% of power for Cycle 17 would not be in compliance with ASME code, as required by 10 C.F.R. § 50.55(a), because many of the tubes in Unit 2's steam generators have exhausted their fatigue life.6 4 An increased risk of a MSLB accident is an obvious example of the significant reduction in the margin of safety posed by the license amendment request, since such an accident would cause the largest leakage from the fatigued tubes.6 5 According to Dr. Hopenfeld's analysis, the proposed license amendment would increase the Large Early Release Frequency (LERF) of radiation escaping to the environment to a level five times greater than the Commission's stated safety goals.6 6 A five-fold increase in risk with potential for large-scale human exposure and the evacuation of southern California is undoubtedly a "significant reduction in the margin of safety." Mr. Large similarly rejects SCE's conclusion that the proposed amendment would not involve a significant reduction in a margin of safety on the grounds that when it was originally determined, the safety margin 6 7 required by the NRC assumed that the functionality of the replacement steam generators complied with the design specifications.

6 8 The fact that they do not is now evident. Critically, the import of this is that "any detriment arising from a design omission or design shortcoming," such as those discussed above, "would not have been included 64 Hopenfeld Decl. at p. 9.65 Hopenfeld Del. at p. 9.66 Hopenfeld Deci. at p. 9.67 Large Decl. at ¶ 8.7.2.6" Large Deci. at ¶ 8.7.4.27 for in the safety margin"69-meaning that the safety margin that exists now has been substantially eroded by the defective tube conditions.

This deficiency, which reduces the safety margin by an unknown degree, is further exacerbated by any additional processes created by the design defects, such as, for example, the fretting fatigue discussed by Dr. Hopenfeld.

Thus, as Mr. Large states, the "particular processes arising from such a omission or shortfall, in this case the occurrence of fretting fatigue at the AVB-to-tube contact point and its potential to substantially reduce the plain fatigue life of individual tubes, would also not have been included for in the safety margin." 7°In sum, the safety margin critically does not take into account the current condition of the plant, specifically, the effect that operating with numerous, severely damaged tubes has on the margin of safety assumed to be in place. In other words, the safety margin is not nearly conservative enough, given the condition of the plant. The second critical point the Staff missed is that the safety margin-overly optimistic to begin with-is now being further reduced, according to FoE's expert, "in ways and to an extent that cannot be precisely defined,

as operating the plant at 70% versus 100% will not reduce the forces acting to degrade the tubes.7 2 Last, regarding stress analyses, MHI's analysis, performed for SCE, of stress on the tubes in the replacement steam generators is deficient in a number of ways that significantly reduce the margin of safety of the proposed change. For example, MHI used a finite element model to calculate the stress to which the tubes were subjected and concluded based on this model that the 69 Large Decl. at ¶ 8.7.4.70 Large Decl. at T 8.7.5.71 Large Decl. at ¶ 8.7.8.72 Large Decl. at ¶ 8.5.3 (stating "The driving force, so to speak, for single tube failure is the differential pressure acting across the tube wall at the operating temperature.

Operating at the proposed 70% RTP will not result in any significant change in the tube differential pressure and the peak tube wall temperature, so the tubes will be subject to the much same forces (radial stress) and tube material strength response (ie the yield stress weighted in account of temperature) as experienced at 100% RTP.").28 tubes would not fail from fatigue.7 3 MHI's analysis was based on erroneous assumptions, however, When corrected, MHI's model would predict tube failure from fatigue because the stress on the tubes exceeds the ASME Endurance Limit.7 4 Taken together, these analyses by FoE's experts show that the proposed amendment would involve a significant reduction in the margin of safety of Unit 2.iv. Summary In order to issue a finding of no significant hazards considerations, the NRC Staff bears the burden of showing that the hazards considerations raised by Friends of the Earth's experts in these comments and by the ASLB's recent decision in the CAL proceeding are insignificant.

The Staff cannot make that showing, and consequently the proposed finding must be withdrawn and a hearing on the proposed license amendment held by an ASLB before the amendment may be approved by the NRC.c. National Environmental Policy Act The proposed license amendment should not be considered prior to a public hearing because the proposal presents a significant hazards consideration.

The National Environmental Policy Act of 1969 (NEPA), 42 U.S.C. § 4321 et seq., requires NRC Staff in such circumstances to at least prepare an Environmental Assessment (EA), which the Staff has not yet done.NEPA requires federal agencies such as the NRC to examine and report on the environmental consequences of their actions. NEPA is an "essentially procedural" statute intended to ensure "fully informed and well considered" decisionmaking.

7 5 Under NEPA, each 73 Hopenfeld Decl. at p. II.74 Hopenfeld Decl. at p. I 1-14; 20, Figure 7.75 Vermont Yankee Nuclear Power Corp. v. NRDC, 435 U.S. 519, 558 (1978).29 federal agency must prepare an Environmental Impact Statement

("EIS") before taking a "major Federal action[] significantly affecting the quality of the human environment." 7 6 An agency can avoid preparing an EIS, however, if it conducts an Environmental Assessment

("EA") and makes a Finding of No Significant Impact ("FONSI").

7 7 Specifically, no EIS is required if the agency conducts an EA and issues a FONSI sufficiently explaining why the proposed action will not have a significant environmental impact.7 8 However, in deciding whether to prepare an EIS, the agency must 1) "accurately identifly]

the relevant environmental concern," 2) take a "hard look at the problem in preparing its EA," 3) make a "convincing case for its finding of no significant impact," and 4) show that even if a significant impact will occur,"changes or safeguards in the project sufficiently reduce the impact to a minimum." 7 9 An agency's decision not to prepare an EIS must be set aside if it is "arbitrary, capricious, an abuse of discretion, or otherwise not in accordance with law.", 8 0 The Federal Register notice is silent as to the application of NEPA to this case. One can only conclude that the Staff is relying on the categorical exemption from the procedural requirements of the NEPA, as described in NRC's regulations at 10 C.F.R. § 51.22(c)(9), available when the Staff makes a finding of no significant hazards consideration.

However, as FoE and NRDC demonstrate in these comments, the Staff cannot make such a finding in this instance.At the very least, an EA and subsequent FONSI must be completed because the proposed amendment would allow steam generators with a severe and dangerous level of wear to operate 76 42 U.S.C. § 4332(2)(C).

77 See Sierra Club v. Dep't of Transp., 753 F.2d 120, 127 (D.C. Cir. 1985); see also Theodore Roosevelt Conservation P'ship v. Salazar, 616 F.3d 497, 503-04 (D.C. Cir. 2010) (explaining NEPA procedures).

78 Dept. of Transportation

v. Public Citizen, 541 U.S. 752, 757-58 (2004).79 Taxpayers of Michigan Against Casinos v. Norton, 433 F.3d 852, 861 (D.C. Cir. 2006) (internal quotation omitted).80 Public Citizen, 541 U.S. at 763 (quoting 5 U.S.C. § 706(2)(A)).

30 without repair. Since the leak of radioactive steam in January 2012 resulting from rapid wear in the steam generator tubes, the licensee has proposed no actions to prevent the conditions that caused the leak. The proposed license amendment therefore poses great potential risk to the environment, as shown by the analyses of FoE's experts and the recent ASLB decision, and thus requires the NRC to follow the procedures under NEPA to address that risk.IV. CONCLUSION For the foregoing reasons, the Staff's proposed finding of no significant hazards consideration should be withdrawn and the significant hazards consideration instead referred to an ASLB, with an attendant public adjudicatory hearing held prior to a decision on SCE's proposed license amendment.

As the ASLB recently held with respect to San Onofre Unit 2: We conclude that until the tube degradation mechanism is fully understood, until reasonable assurance of safe operation of the replacement steam generators is demonstrated, and until there has been a rigorous NRC Staff review appropriate for a licensing action, the operation of Unit 2 would be outside the scope of its operating license because the replacement steam generator design must be considered to be inconsistent with the steam generator design specifications assumed in the FSAR and supporting analysis.8 1 There is simply no basis for a no significant hazards consideration determination in the case of the proposed license amendment for San Onofre Unit 2.Respectfully submitted,/Signed (electronically) by Richard Ayres/Richard Ayres Jessica Olson Kristin Gladd Counsel for Friends of the Earth Ayres Law Group 1707 L St, N.W., Suite 850 Washington, D.C. 20036 Telephone:

(202) 452-9300 E-mail: ayresr@ayreslawgroup.com 81 ASLB Order at p. 32 (emphasis supplied).

31

/Si2ned (electronically) by Geoffrey H. Fettus/Geoffrey H. Fettus Counsel for NRDC Natural Resources Defense Council 1152 1 5 th St. N.W. Suite 300 Washington, D.C. 20005 Telephone:

(202) 289-2371 E-mail: gfettus@nrdc.org Dated in Washington, D.C.this 1 6 th day of May 2013 Attachments

1. Declaration of Dr. Joram Hoppenfeld

2. Declaration of John Large 3. Declaration of Dr. Victor Gilinksy 4. Declaration of Arnold Gundersen, in Support of the June 18, 2012 Petition to Intervene by Friends of the Earth Regarding the Ongoing Failure of the Steam Generators at the San Onofre Nuclear Generating Station 5. MHI Root Cause Analysis and Supplemental Technical Evaluation Report (Selected Excerpts)6. Southern California Edison Co. (San Onofre Nuclear Generating Station, Units 2 and 3), LBP-13-07 (May 13, 2013)32 ATTACHMENT 1 Declaration of Dr. Joram Hopenfeld UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE NRC STAFF In the Matter of )))SOUTHERN CALIFORNIA EDISON CO. ))(San Onofre Nuclear Generating Station, )Units 2 and 3) )Docket ID NRC-2013-0070 May 16, 2013 DECLARATION OF DR. JORAM HOPENFELD I CONTENTS Qualifications of Dr. Hopenfeld

-3 Summary -4 1. No Hazards Change considerations

-6 a. Question 1 b. Question 2 c. Question 3 Introduction

-NO HAZARDS CHANGE CONSIDERATIONS

-IOCFR 50.92 -6 2. APPENDIX A -FATIGUE ANALYSIS -11 Part 1.a. Stress Concentration

b. Loss of Wall Thickness c. Surface Finish d. Correction of MHI stress Part 2.Rebuttal of SCE/MHI Fatigue Statements

.3. APPENDIX B -Discussion of Accident Scenarios

-25 A. SGTR -stuck open Relief Valve B. SGTR -initiated by Isolation Valve closure C. SGTR -initiated by seismic events D. Station Blackout E. MSLB 4. REFERENCES

-32 2 Qualification of Dr. Hopenfeld to Assess the Southern California Edison Response to 10 CFR 50.92 While employed by the Nuclear Regulatory Commission, NRC, Dr. Hopenfeld's research included a focus on steam generator tube degradation.

Consequently the NRC launched a Steam Generator Action Plan, SGAP, to address the various safety issues raised by Hopenfeld in a series of documents from 1992, known as the DPO and GSI 163. On September 2007 the NRC. issued a new performance technical requirement specifications, TS, to reduce the risk from accident induced and normal operations tube ruptures.

This action essentially closed the DPO and GSI 163, as discussed at the May 7, 2009 Advisory Committee on Reactor Safeguards (ACRS) meeting. During the fifteen year review Dr.Hopenfeld made numerous presentations to the Atomic Safety Licensing Board (ASLB)and the ACRS on various steam generator related issues.* Steam Generator Degradation Monitoring.

• Erosion/Corrosion, FAC (relevant to the feed ring failure at SONGS (1992)* Safety Consequences of Steam Generator Tube Failures,* Iodine transport and Spiking,* POD of crack detection by Eddy Current,* Metal Fatigue from Thermal Transients (PWRs and BWRS)* Vibrations in BWR dryers.* Managed a major International program, MB-2 (US, UK, EPRI ) on steam generator performance during design basis accidents.

• Conducted sensitivity studies with the RELAP computer code on operator's ability to keep the SG inventory at mid level as a function of the number ruptured tubes." Conducted studies on jet erosion as a potential for leakage increase during SG accidents.

• Conducted numerical studies on SG tube ruptures during severe accidents" Designed, fabricated and field-tested instrumentation for a very harsh vibration environment.

• Holds several patents on methods for monitoring.wall thinning* Managed the development of acoustic leak detection system for LMFBR steam generators.

• Testified before Congressman DeFazio regarding steam generator degradation at the Trojan Nuclear reactor.3

SUMMARY

Southern California Edison (SCE) requested the approval of the Nuclear Regulatory Commission (NRC) for a change in Technical Specification (TS) 5.5.2.1 1.bl to allow operation of San Onofre reactor Unit 2 during Cycle 17 at power levels up to 70% of Rated Thermal Power. To obtain approval, SCE claimed that it has demonstrated that the change would not involve any significant hazards, as required by 10CFR 50.92. The assessment in this declaration for Friends of the Earth demonstrates that that SCE has in fact not met the standards prescribed in 10CFR 50.92 which require a "no" answer to three questions.

The NRC I0CFR 50.92 states, 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

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

3. Does the proposed change involve a significant reduction in the margin of safety?SCE's justification for providing the three negative answers is based solely on a brief fatigue assessment by the replacement steam generator manufacture Mitsubishi Heavy Industries (MHI Ref 1), which showed that the vibration-produced stresses were too low to cause fatigue failures.

SCE endorsed these findings despite the fact that MHI, (a)relied on data which was inconsistent with the visual observations of tube degradation and (b) disregarded American Society of Mechanical Engineers (ASME) code requirement to account for variation in code data and field conditions.

Equally important is the fact that in their effort in trying to justify the restart of San Onofre reactor Unit 2 there is no indication that SCE utilized the large amount of data generated by the "lessons learned" from the vibration fatigue tube failures at North Anna (1987), Mihama (1991)and Indian Point (2001).SCE answered "no" to the three 50.92 questions but only by disregarding fatigue damage to existing tubes and industry guidelines of how to evaluate tube integrity under multiple loads. The SCE analysis is based on showing that San Onofre reactor Unit 2 will operate safety because tube rupture is only controlled by tube wall thickness and the tube differential pressure, AP. This declaration shows that the controlling factors of tube rupture are more complex when a significant fraction of tube fatigue life has already been incurred and in addition to AP loads the tube is subjected to cyclic loads from flow-induced vibration.

Under these conditions, the determination of the margin of safety, solely on the basis of AP, is invalid and significantly non-conservative.

4 The assessment herein includes a discussion of potential radiation release from tube ruptures for five design basis accidents and one severe accident.

Because of the unprecedented and unforeseen damage to 1806 tubes during one cycle of operation, there is no data that one can use to reliably calculate the consequences of tube failure risks in such accidents.

This declaration demonstrates the high degree of technical uncertainties and lack of robustness in the "no" answers provided by SCE.The analysis in this declaration indicates that a Main Steam Line Break (MSLB) would result in the most significant large early radiation release (LERF) because of the potential for many tubes to rupture and the high probability for human errors. Events which occur more frequently than MSLB exposing the tubes to relatively lower stress such as unplanned valve opening or closing or earthquakes have a lower probability for human error but are more difficult to analyze. Considerable effort would be required to ensure that the safety risk from such events is significantly lower than the safety risk from MSLBs.If as few as 1% of the degraded tubes in one steam generator, operating for six months, fail during an MSLB, the result is an LERF of 5x10-5 /yr which exceeds the Commission safety goals by a factor of 5.My assessment leads me to the conclusion that the proposed SCE TS change:Represents a new accident with high risk significance , Would create a new accident previously not evaluated and,-A Would involve a significant reduction in the margin of safety.Therefore my answer to each of the three questions is yes.5 NO HAZARD CONSIDERATIONS

-10 CFR 50.92 Introduction A determination of No Significant Hazard must provide assurance that the San Onofre Nuclear Generating Station (SONGS) licensing base (CLB) will be maintained between Steam Generator (SG) inspections during future operation over 18 months so-called cycle 17. However, SCE has failed to demonstrate that the modification of SONGS Technical Specification (TS) which will allow a change from 100% power to 70% power represents an added assurance of the functionality and integrity of SG tubes. As discussed below such a change entails a significant reduction in the margin of safety.SCE answers no to all three 10 CFR 50.92 questions.

Their answers are based on the presumption that a change in power level can be discussed without giving any considerations to the physical conditions of the tubes before and after the change. SCE is mistaken in believing that tube integrity is a function of the power level alone and independent of the actual degree of tube degradation.

As discussed in Appendix A, large numbers of both plugged and unplugged tubes have exceeded their allowable fatigue life.This loss of tube integrity significantly affects primary to secondary leakage during design basis accidents and consequently increases the Large Early Release Frequency (LERF).SCE disregarded the affects of fatigue damage on tube degradation by claiming that the stresses were too low to cause tube fatigue. The analysis in this declaration leads to a different conclusion:

the vibration during cycle 16 resulted in sufficiently large cyclic stresses to cause fatigue damage to a significant number of tubes.Another important factor that must be considered in comparing the change of operating Unit 2 from 100% to 70% power is the unknown behaviour of the tubes at the lower power level. Even if vibrations due to fluid elastic instability were significantly reduced at the beginning of the cycle it is uncertain that this will remain so through the five months of operations.

Tubes with low natural frequencies may continue to wear due to fluid turbulence.

The resultant increase in clearance between the AVB support and the tube could lead to an increase in the intensity of the impacts between these two components.

This could lead to an abrupt failure even for those tubes whose fatigue life has not been used up during cycle 16, i.e their cumulative usage factor was less than one (CUF < 1).It is for these reasons that my answers are in the affirmative to all three 1OCFR 50.92 questions as discussed below.2. Answers to 1OCFR 50.92 Questions 6

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

Response:

Yes In comparing the change proposed by SCE, one must compare the change in San Onofre reactor. Unit 2 at the beginning of cycle 16 to its proposed cycle 17 operations.

Consideration must be given to both power levels and the degree of tube degradation, not just the power level as SCE have done. The operation of Unit 2 which entered service in in 2011 (cycle 16) at 100% power must be compared to how it would operate if permitted to restart in cycle 17 at 70% power level with a large number of defective tubes.The proposed change would significantly affect the probability of accident initiators because a certain percentage of steam generator tubes have used up their entire or a large fraction of their allowable fatigue life during cycle 16. For this reason the operation of San Onofre reactor Unit 2 during cycle 17 will fall outside the bounds of the accidents that were evaluated in the existing SCE Updated Final Safety Analysis Report (UFSAR).While the proposed change does not affect the design of SG or its method of operation, it does increase adversely the consequences of Design Basis Accidents (DBAs), i.e., main steam line break (MSLB) and tube rupture (SGTR). The SONGS Technical Specifications, TS 5.5.2.11, require that SONGS provide the NRC during every outage an assessment, CM, with respect to tube structural integrity, accident induced leakage, and operational leakage. As discussed by SCE UFSAR Sections 3.2 and 5.2.9, the entire CM assessment is based on "operating experience with SG tube degradation mechanism that result in tube leakage".

Likewise, SCE's determination of Core Damage Frequencies (CDF) is also based on leakage methodology which was derived from tubes that were degraded by stress corrosion cracking.

As discussed below, these results are not applicable to the 70% power operation with fatigued tubes. A comparison of operation at 70% power with fatigued damaged tubes versus operation at 100% power with undamaged tube must consider fatigue damage. SCE is wrong in claiming that the change from 100% power to 70% only changes the power level without any potential adverse safety consequences.

In discussing its "no" response to 1 OCFR 50.92 question 1, SCE did not explain why industry guidance on how to ensure tube integrity was not included in its submission.

These guidelines, issued by the Technical Specification Task Force, TSTF 449, specify that primary/secondary pressure differential AP loads alone are not sufficient to ensure integrity when other loads are also present. Specifically,"additional loading conditions associated with the design basis accident or combination of accidents in accordance with the design and licensing base shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse." 7 In some accidents cyclic loads may be controlling tube rupture in others, AP loads may act in tandem depending on the degree to which the tube wall thickness was already reduced by wear. Therefore, the mean stress level of any tube must be considered together with superimposed cyclic stresses.

SCE treated cyclic loads as if they have never occurred at San Onofre reactor Unit 2.As discussed in Appendix A, the predominant degradation mechanism at San Onofre reactor Unit 2 during cycle 16 for an unquantified fraction of the tubes is fretting fatigue.Fretting fatigue would result in a larger and faster leakage rate from a tube rupture than the leakage from a tube that was degraded by cracks due to Stress Corrosion Cracking, (SCC) or wall thinning by erosion alone. The existing leakage performance criteria are based on the latter.For those tubes in San Onofre reactor Unit 2 where the Stability Ratio (SR) was relatively low, (less than 0.4,) tube rupture is expected to be controlled by burst pressure.

In this case present performance criteria are applicable.

During operational (non-Loss-Coolant Accidents, LOCAs) and accident transients (LOCAs) cracked tube can be expected to result in a slow progressing leakage, in contrast when fretting fatigue is the cause of tube failure the leakage would occur suddenly and proceed rapidly to its maximum as happened at North Anna (Ref 2).To evaluate the effect of existing defects in San Onofre reactor Unit 2 on the consequences of a given accident one must identify first the fraction of the tubes that were damaged predominantly by fatigue and the fraction of tubes that were damaged by wall thinning alone. This must take into account that high cycle vibration fatigue does not lend itself to in-service detection.

Tube fatigue life is almost entirely spent in the incubation period and once the crack is formed failure would follow quickly.To comply with industry guidelines TST- 449, Rev 4, degradation of each tube must be assessed simultaneously in terms of both its existing fretting damage (wall thinning) and its local SR. The number of tubes which are susceptible to rupture by fatigue during a given accident scenario must be known if one is required to predict accident consequences.

Until this is done the present pressure based burst performance criteria cannot be used as a reliable indicator of risk. As a result, it must be conservatively concluded that allowing San Onofre reactor Unit 2 to operate at any power level would significantly increase the consequences of the accidents, which were evaluated by SCE and were described in the UFSAR.2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response:

Yes 8 The proposed change will introduce significant changes to postulated accidents resulting from tube degradation.

Appendix B discusses how primary to secondary leakage from fatigue-ruptured tubes differs from leakage that resulted from tubes that failed from excessive loads. Since existing safety studies are based solely on changes in AP this represents a new type of accident.The analysis in Appendix A demonstrates that some tubes will enter service in Cycle 17 with no fatigue life left. The leakage from these tubes will not be affected by changes in AP. The rupture of these tubes would depend on the intensity of cyclic stresses that varies with the stability ratio, SR. The SR and the AP are independent variables:

different driving forces govern their respective changes during a given accident.

For this reason the operation of San Onofre reactor Unit 2 with fatigued tubes creates a new and different kind of an accident.3. Does the proposed change involve a significant reduction in the margin of safety?Yes Because many tubes have exhausted their fatigue life the proposed change would not be in compliance with the ASME code as required by 10 C.F.R. 50.55a, Codes and Standards.

Since the SG tubes form a barrier between the radioactive fission products in the primary water and the secondary system, loss of fatigue life reduces the safety function of the SG.Initial assessment in Appendix B suggests that the MSLB accident would be the most damaging accident from the standpoint of causing the largest primary/secondary leakage from fatigued tubes in comparison to more frequently accidents with a lesser damage potential.

Based on a probability of E-4/yr (one in every 10,000 years) that a steam a line break would occur outside containment, the Large Early Release Frequency, LERF of radiation escaping to the environment due to the reactor core becoming exposed is also E-4 /year because, as discussed in Appendix B, no credit can be given to the operators that they would terminate the accident before depleting the reactor water storage tank, RWST. This represents an LERF of 5 E-5 /yr (one in every 20,000 years) for six months operation.

This is an increase by a factor of 5 over NRC goals as considered in NRC Probability Risk Assessments.(Ref 11)Assuming that the AP will not be large enough to rupture tubes, and no leakage from fatigued tubes, SCE calculated a change in LERF of 4E-6/yr (NRC AIT Report July 18 2012)9 SCE results are not realistic and therefore the answer to question 3 must be yes.Conclusions SCE analysis is based on the assumption that any increase of radioactive primary water during hypothetical accidents would be controlled by burst pressure, AP. This assumption is flawed because some tubes at San Onofre reactor Unit 2 have already used up their fatigue life. In this case leakage increase would be controlled by the intensity of vibration-induced stresses, not AP. The large and sudden fatigue tube ruptures at North Anna, Mihama and Indian Point occurred when the All was essentially constant but tubes failed because they exhausted their fatigue life due to intense vibration similar to those that occurred at SONGS.In spite of its 170,000 inspections to understand the tube wear problem, SCE has not even mentioned the possibility that a fraction of the tubes had sustained fatigue damage. Nor did SCE discuss the uncertainties and errors in the MHI fatigue analysis and how they could affect their "no" answers.The assessment in this report does not support SCE's position that operation of San Onofre reactor Unit 2 for five months at 70% power will not affect safety. It is shown, that SCE conclusions are not founded on science and equally important are not conservative.

Operation of San Onofre reactor Unit 2 even for one month at any power level would present a significant safety risk.1 Joram Hopenfeld declare, under penalty of perjury, that the foregoing information and facts are true and correct to the best of my knowledge and belief, and that the opinions expressed herein are based on my independent and best professional and personal judgment..LI Joram Flopenfeld 10 APPENDIX A- FATIGUE ANALYSIS Introduction When an SG tube is in contact with its support either the Anti Vibration Bar (AVB) or the plate (TSP), the two contacting surfaces are damaged by fretting.

As discussed in detail by the Electric Power Research Institute (EPRI Ref 4) and Volchock (Ref 5)fretting damage occurs due to a combination of the sliding motion between the surfaces and the impacts when an external cyclic load is superimposed on the sliding motion. The sliding motion alone produces fretting wear while the cyclic impact produces fretting fatigue, the synergy of the two significantly reduces fatigue life and enhances wear. This synergy is especially important when wear predictions are mostly based on empirical parameters from laboratory tests. Even though the EPRI report indicates that wear and fatigue are controlled by different mechanisms and their respective equations are different, SCE, in calculating tube wall thinning, applied equations for wear without considering the differences between wear fatigue and fretting wear. Numerous considerations must be given in extrapolating lab data on fretting wear to SONGS conditions.

Since SCE did not perform any similitude studies on wear rates, SCE projection of tube wear for cycle 17 are unreliable.

This appendix was divided into two parts, the first part shows why SCE analysis lead them to believe that the tubes at SONGS did not suffer fatigue damage. The second part, Part 2, is a rebuttal to SCE and MHI's contention that tube fatigue damage during cycle 16 can be ignored.PART 1 -MHI ANALYSIS.As describe in Attachment 4 MHI Document L5-04GA564, MHI used a finite element model (FE), to calculate that the tubes were subjected to a stress of 4.2 Ksi (P 16-2), which is smaller than the endurance limit stress of 13.6 ksi (P 16-2). Consequently, MHI concluded that the tubes would not fail from fatigue even if they were subjected to infinite number of stress cycles, (P 16-13).The MHI results are based on two erroneous assumptions.

When these assumptions are corrected the opposite conclusion is reached. The issue is not with the FE, rather it is how the FE results were adjusted to account for high stresses at surface discontinuities.

a. Stress Concentration It is a well-established fact that geometrical discontinuities such as sharp corners introduce high local stresses, which act as a site for crack initiation.

A common engineering practice is to fillet or chamfer sharp corners to reduce stress concentrations and increase fatigue life. Any conceivable discontinuity has been considered and the!I results have been published in numerous publications to guide designers in selecting the particular fillet for a given application.

MHI used a design chart, Figure 2, for a tube in pure tension to determine the stress concentration factor Kt. Assuming an undisclosed value for the fillet radius and the value of the parameter (t) MHI concluded that Kt was less than 1.5 when t/r =1.33. These numbers indicate that MHI used a value oft/h that exceed unity. Had MHI assumed smaller values for t/h, and a smaller radius, Kt would have exceeded 1.5 because Kt is sensitive to the assumed geometry of the fillet. MHI selected an arbitrary geometry, which is not valid, and for this reason they only obtained an unrealistically low value for Kt.Fig 2 is intended for applications when one is trying to minimize stress concentration.

Visual examination of the contact between the AVB plates and the tubes do not suggest that the relative motion resulted in geometry with minimum stress concentration.

On the contrary, as shown in Figures 3 and 4, the method in which the AVB interacted with the tubes allows for a formation of sharp corners at the intersection of the plate with the tube.MHI's own discussion is not consistent with their application of Chart 5 in Fig. 2. The observation that the "tube and the AVB are worn into each other" and the fact that the AVB plate has sharp corners suggest that Chart 3.5 does not apply to observed wear pattern.The model shown in Figure 5, represents more closely the wall thinned geometry than the one used by MHI in selecting the stress concentration factor. Since Figure 2 does not provide data for fillets with very small radius, it is necessary to consider a similar geometry giving Kt values for small radii. In Figure 6 (a special case of Fig 2, di = 0), Kt is plotted for very small radiuses for bending -Kt values in tension are similar).Using a reported wear of 35%TW. Kt is calculated as follows: t = (1 -%TW) T = (1-0.35) 0.043 = 0.028in d/D = D/D-2t = 0.750 /0.750 -0.056 = 0.750/0.694

= 1.08 Kt = 5 when r= 0.0014 for Kt =5, r = 0.002 (0 .694) = 0.00 14 (Theoretically the chamfer radius of a sharp corner is zero, and therefore Kt will tend to be very large for a finite but small radius of 0.00 15 which is close to describing a sharp corner, Kt exceeds 5.)b. Loss of Wall Thickness (wall thinning)The effective wall thickness Teff of the geometry in Figure 4 can be expressed as: 12 Teff =tx (20)/360 +Tx(360-20)360 0 = 2 Cos -1 (d/2 +t )/(d/2 +T)For a 35% and 70 % tube wear, 0 equals 44.6 and 44.1 degrees respectively, the corresponding effective tube thickness equals 0.0360 and 0.0345 and respectively.

c. Surface Finish Fatigue life, and therefore the endurance limit, is strongly affected by surface finish.Figure 9 show that fatigue life can be considerably reduced by abrasions.

The data (Edison Attachment 6- Appendix D pages 130 -131) indicates that the fretted tube surfaces do not maintain their original surface finish instead they are severely scarred. Such scars are sites for the formation of micro cracks.Bounding calculations would require that the ASME design stress used by MHI (13.6 ksi)be lowered to account for surface finish. It is not clear however that the introduction of both a stress concentration and surface finish correction simultaneously would not be overly conservative.

Since no data was found in the literature where both a sharp comer and adjacent rough surface, a surface finish correction was not included in the present assessment.

In that sense, the application of a concentration factor of 5 together with curve C of Figure 1 may not be conservative.

d. Corrected MHI stress.Corrected stress = MHI stress multiplied by concentration correction factor K, multiplied by thickness correction factor Tc, = 4.2Kx(Tc)K = actual stress concentration factor/ MHI concentration factor = 5/1.5 = 3.33 l/Tc =- Decrease in wall thickness

/original wall thickness

= 0.036/0.043 for beginning of cycle 0.0345/0.043 at the end of cycle assuming the same wear rate.Tc =1.19tol.25 Increase in stress= 4.2x 3.33 x 1.19 to 4.2x 3.33 x 1.25 = 16.7 to 17.5 Actual increase over the endurance limit = 16.7/13.6 to 17.5/13.6

= 1.22 to 1.29 e. Conclusions 13 The impact of correcting the MHI calculations is demonstrated in Figure 7: it is self-explanatory.

It should be noted that the stresses that ruptured the tubes at Mihama and North Anna (Ref 5), about 7.4 to 8.7ksi and 4 to 9ksi respectively was within the error band indicated in the Figure. The ruptures at Mihama and North Anna occurred due to support plate/ tube interaction not due to AVB/tube interaction.

14 APPENDIX A GRAPHS/CHARTS

1. Purpose The purpose of this document is to show that the stress of the tube in SONGS RSG due to in-plane vibration is under the fafigue limit.2. Conclusions The stress on the tube due to in-plane vibration is 4.2ksi and is under fatigue limit (13.6ksi).

The tube has structural integrity for the stress due to in-plane vibration from the view point of fatigue evaluation.

3. Assumptions and Open Items The tube deforms in-plane until contacting with the outer next tube in Row direction due to in-plane vibration.

The stress due to in-plane vibration is high cycle latigue 4. Acceptance Criteria The fafigue limit is 13.6ksi according to the folkYving design fatigue curve ft 1-9.2U2 199R WAiýTict iLu, DrIvISJoN I -AIpENDJCRs "6 I 74 Cum* A 14 -M ;a=C I 106j 10,tO toI Figure 4-1 Design Fatigue Cunre for Tube I-Figure-I Fatigue data used by MHI to determine tube fatigue life. The cycle independent line represents the endurance limit, MHI used an endurance limit of 13.6ksi. Attachment 4, P 16-2. data for smooth specimen.15 CHARTS 157 t/r Charl SIress concenlralion facitor K, fil'r .1 Lulh- in ltnsion with fillet (L1e an.d Ades 1956: FSDU 1981).Figure 2 -Stress concentration factors used by MHI for calculating maximum tube stress, Attachment 4 P.16-2. Source: W.D. Pilkey, Peterson's Stress Concentrations Factors, John Wiley and Sons 1997.16 3.2 Wear Pattern-2 (Local Wear on Tube Surface)Characteristics (1) Local wear occurs on the tube but the wear surface is not exposed (cannot be seen)(2) Unable to determine if wear occurs on tube or AVB or both a) Unable to determine the direction of motion or vibration 4 An extreme Interpretation Is that both tube and AVB are worn Into each other.t-ibe tuM AVS V 1. th. mealic sheen =M ca.Unable Io see the metaf h in extreme case, both the tube due to narrow wear area & AVB wear into each other (a) Case 1 (b) Case 2 Fig-2 Wear Pattern 2 Fig. 3 -Wear due to AVB/tube Interaction

-Attachment

4. It should be noted that both the impact and the sliding motions play a part in the tube/AVB interaction.

These factors reduce tube strength because of material loss but also because of loss of fatigue strength.17 Non-proprietary Version I) (P.10-20)DocumeXL5-04GA564(9)

V h Fig. 6-4 Wear shape of tube at the contact point with AVB Figure 4 -MHI description of wear shape at tube/ AVB contact point. Attachment

4. P. 10-20 T -originaIwall thicknesl t = remaining wall tikness d = inner dla v Figure 5 -Schematic for determining a stress concentration factor Kt and reduced wall thiclkness of a tube due to double sided wear to a thicklaess (t) over an arc defined by 0 18 166 SHOULDER FILLETS 4.5 r d- 0:00M.. ... j. ..... .I*...- -d= .2 T ...T M: ... .' "'- -ii L jLKj 351 S tf~~ .I,A Kr rr. 0. 01.5., .,..,.30D d.- id0 02 2.5 pit t' I ' -V : : q 10.' .. ... ... ..r...... ......... .f ..7 1 -. ....1011 1i5 20 2.5 Ch"art .II Stiss Lvf,',Lr.IIj1 i~n faCIOTs K, br bcndin; of a stcpOrcd barf ciraivLir cross, w,'ii'h a filh I (h;Lw on pholr .lr;sfij t-,ls of Leven wind I lartmran 19511; Wilson LrId Whil 1973). This char wFcs'r Ito supplemcnt Chart 3.'10.Figure 6 -Stress concentration used in the present analysis.

W.D. Pilkey, Peterson's Stress Concentrations Factors, John Wiley and Sons 1997. (Similar Kt values in tension -Peterson's Chart 3.4, and for internally pressurized vessel Chart 3.6 for a small radius)19 COMPARING THE MHI CALCULATED STRESS WITH THE STRESS CALCULATED IN THE CPUC REPORT Stress U N SAFE TO 0 P E RAT E (ksi) CORRECTED STRESS (CPUC Report)17.5 --13.6 ASME Endurance Limit -UNCERTAINTY PIHI CALCULATED BANJ 4.2 STRESS SAFE TO OPERATE _Number of vibration cycles to failure Figure 7 -Effect on potential fatigue damage by correcting the MHI calculation PART 2 -Rebuttal to SCE/MHI Assessment of Hopenfeld Fatigue Calculations Appendix A was attached to Dr. Hopenfeld Testimony to the CPUC and has been in the public domain since March 29, 2013. In reply to questions from ABC Channel 10 in San Diego, SCE and MHI responded on April 2 5 th as follows: SCE"Hopenfeld's fatigue analysis concerning in-plane tube vibration is significantly flawed in that it applies an unreasonably high stress concentration factor based on solid body geometry rather than the more realistic stress concentration factors for a cylindrical geometry applicable to the SONGS steam generator tubes." MHi MHI did analyze the potential for fatigue failure of the RSG tubes under operating conditions and determined that fatigue was not a credible tube failure mechanism because the stresses sustained by the tubes due to in-plane vibration are well below the stresses that would cause fatigue failure. The analysis that supports this conclusion is contained in Appendix 16 to the "Tube wear of Unit-3 RSG -Technical Evaluation Report." It should be noted that the technical reviews and analysis, both by the NRC and industry experts, have not mentioned fatigue failure of the tubing." 20 Since SCE approved MHI fatigue analysis as specified in the original SCE design document, the reply to the above separate statements will be consolidated.

REPLY SCE/MHI calculations are based on ASME data that has not been corrected for the conditions that represent the tube surface following fretting after 18 months of operations.

The S-N fatigue data was obtained by testing a number of polished solid specimens and the lines represent mean stress limits. It is a common practice of a user of the ASME code to make conservative correction when that data is applied to field conditions which are drastically different than those in the code. When applying the data to tube surfaces that have experienced fretting, Curve C, which was used by MHI, must be lowered to reflect the increase in surface roughness due to fretting.

As shown in Figure 9 a change of surface roughness from 0.05 microns to 2.67 microns reduces the fatigue life by a factor of 8.8.In their report (Appendix 16), SCE/MHI stated that because the AVB and the tubes are imbedded in each other the condition of the surface cannot be seen. Given that the respective surfaces have been sliding and impacting each other it is difficult to imagine how SCE/MHI concluded that such motion would produce polished surfaces.

One must conclude that SCE/MHI disregarded the intent of the ASME code by not adjusting the Curve C stress to account for surface roughness.

Comparison of the actual ASME curve with those that were reproduced by SCE/MHI shows that SCE/MHI incorrectly labeled the data to indicate that it was generated for tubes and was limited to operation below 800 F. The data would not be applicable to severe accidents, which were discussed above. It is not clear why SCE/MHI mislabeled the ASME figure to indicate that it was generated for tubes.21 ORNL-DWG84-6108 ETD I .. ..CURVEA 11I l l~24 1 ---u20_j 118 E l --<.6 I --I I I CURVE lll 16 7 1e ... .. -.. V: VE C wwlllt I I II 14 -M I. E REC 1 2 ..... I 106 lo,10 109 1010 loll N, NUMBER OF CYCLES NOTE: E* 28.3 X 106 psi Fig. 1. ASME Code Sect. III high-cycle design fatigue curves for austenitic steels, nickel-chromium-iron alloy, nickel-tron-chromium alloy, and nickel-copper alloy for temperatures not exceeding 800°F (from Ref.22).Figure 8 -Showing the same ASME data that was shown in Figure 1 but with the original correct caption. This the data was not for SONGS steam generator tubes.22 Table 12-3 Fatigue life of SAE 3130 steel specimens tested under completely reversed stress at 655 MiPat Median fatigue li.Type of finish Surface roughness, pm cycles Lathe-formed 2.67 24,000 Partly hand-polished 0.15 91,000 Hand-polished 0.13 137,000 Ground 0.18 217,000 Ground and polished 0.05 234.000 Superfinished 0.18 21 2000 t P. G. Fluck. Am. Soc. Tcsr. ,fater. Proc., vol. 51, pp- 584-592, 1951.Figure 9- Effect of surface roughness on Fatigue life 2. Incorrect selection of the Stress Intensity Factor The ASME curves are used only to calculate average stresses only. At least 100 years of experience has been accumulated to show that sharp surface discontinuities introduce high local stress concentrations where crack are initiated.

The ASME code requires that the average stress of a component be multiplied by the appropriate stress intensity factor.Because of the importance of local stresses on fatigue life, hundreds publication are available for smooth discontinuities and thereby reducing local stress. The concept 23 SCE/MHI Field observations Sharp corners = poor fatigue strength Decreasing Failure probability J.Av:C "'ur inio cacd /ixt sharp notch smooth notch better notch much better notch Figure 3 -IGNORING FIELD DATA SCE/MHI SELECTED GEOMETRY WITH GOOD FATIGUE LIFE FIGURE 10- This has been duplicated to show why SCE /MHI concluded that the "vibration are well below the stresses that would cause fatigue failure" as illustrated in Figure 7.Sharp corners lead to a poor fatigue strength while smooth comers or a gradual transition reduces stress concentrations thereby improving fatigue strength.

The most common source for stress concentration factors are the Peterson's charts which are available for numerous different geometries.

As shown in Fig. 2of Appendix A, SCE/MHI had to select a fillet radius in order to calculate the stress concentration factor. If one selects the radius arbitrarily, you can get any number he wishes. SCE/MHI used radius that they have redacted, however an examination of their calculated stress concentration factor,(K t) clearly indicates that they selected a relatively a smooth fillet (large radius) and that SCE/MHI did not select a sharp notch. Since SCE/ MHI stated that the interface between the AVB and the tube is not visible, and their interpretation of the contact surface geometry shows a 90 degree comer, it is impossible to conceive how could they justify using a large radius fillet. The Peterson charts were designed to minimize stress concentrations, when the AVBs impact the tubes they do not follow fracture mechanics guidance to avoid formation of sharp notches.Figure 10 above illustrates schematically how fatigue life is improved as the notch radius increases.

In the stress calculation, Par 1, I have selected a sharp notch because this is consistent with the observation that the AVB and the tube imbedded in each other through impacts.24 Fatigue damage by impact loads would lead to a brittle fracture because such loads do not mitigate slip. Selection of sharp notch geometry is appropriate because such notches can lead to a brittle fracture.

In contrast, a well-designed fillet would result in a ductile fracture.Another reason why it is incorrect to select an arbitrary fillet radius with smooth surface to calculate fatigue life is the synergy between surface roughness effects and cyclic loading effects. Such synergy leads to a significant reduction in fatigue life as has been clearly demonstrated in Reference

3. Therefore even if MHI had corrected their stress intensity factor (K t) of 1.5 to account for surface roughness (Fig 9) it still would leave a large uncertainty due to synergy. This only indicates that calculations which are solely based on a sharp notch (K t =5) may not be sufficiently conservative.

As a reality check on their fatigue model, one must wonder why SCI/MHI did not compare their calculated stress of 4.2ksi at San Onofre reactor Unit 2 with the stress (7.4-8.7 ksi and 4-10 ksi) that caused the rupture at Mihama and North Anna (Ref 6)respectively.

Such a comparison should be made for each affected tube on the basis of the local velocity, steam quality, tube stiffness, natural frequency, and temperature gradients across the tube wall and A P. SCE/MHI should show that the differences in conditions at Mihama and North Anna vs. conditions in San Onofre reactor Unit 2 account for the fact that Mihama and North Anna tube ruptures occurred at somewhat a higher stress.The SCE/MHI statement that stress concentrations at sharp discontinuities depend on whether the component is a hollow or solid, appears to be a new discovery in fracture mechanics.

It is well established that stress concentration gradients at sharp notches decrease rapidly with the distance from the notch. In other words, the crack would be initiated at the tip of the discontinuity and is practically independent of the geometry further away. As the comment to Figure 6 indicate, examination of Peterson's charts clearly demonstrates this point.In light of the many unstated assumptions that SCE/MHI used in applying Figures 1 and 2 to the SONGS tubes, the statement that it is unrealistic to apply "stress concentration factor based on solid body geometry rather than the more realistic stress concentration factors for a cylindrical geometry applicable to the SONGS steam generator tubes." Is not appropriate.

I used the solid geometry for convenience only. Extrapolation of the tube data in Figure 2 to sharp comers (r-"0) would have resulted in the same stress concentration factor.SCE/MHI appear to justify their position that fatigue failure would not occur at SONG by relying on the fact that the NRC did not raise this issue. In the light of the significant component failures in power plants from high cyclic fatigue due to thermal or hydraulic instabilities, it is puzzling that the NRC did not raise the fatigue issue. The suggestion that the fact that the NRC did not raise the fatigue issue is not a valid technical reason 25 that supports SCE/MHI fatigue analysis.

Nevertheless, ultimately it is SCE's responsibility to operate the plant safely. It is not the NRC's responsibility.

26 APPENDIX B FRETTING FATIGUE TUBE DAMAGE -NEW AND DIFFERENT FROM ANY ACCIDENT PREVIOUSLY EVALUATED AT SONGS 1. Introduction There are two main reasons why fretting fatigue introduces a new un-analyzed accident at SONGs. The massive fretting fatigue suffered by the SONGS steam generators is unique in the history of United States SG tube degradation.

Assessments of accidents, which could be induced by degraded SG tubes, were focused on the consequences of operations with tubes that were degraded by Stress Corrosion Cracking (SSC). With three exceptions, North Ana (1985), Mihama (1991) and IP B (2000) all other tube ruptures resulted from stress corrosion cracking and loose part wear as shown in Table 1 below.Fatigue failures at these three plants were limited to a single tube and unlike at SONGS the root cause was fairly well understood.

Given the fact that fatigue damage in the above three accidents was confined to one tube it is puzzling why SCE/MHI completely ignored the wide spread fatigue damage at SONGS. In comparison very extensive fatigue investigation was conducted in connection with the North Anna event, (Ref 6).Since SCE frequently quotes the existing performance criteria the understanding of these criteria is critical in assessing the SCE conclusion it would be appropriate to briefly review the basis for the present performance criteria.Since it became obvious in the late 1980s that steam generators would have to stay in service with ssc cracks all efforts were focused on attempting to define the safety consequences of such operations.

Starting in 1991 with a series of documents that became known as Differing Professional Opinion (DPO), efforts were made to cope with various aspect of the problem. In particularly the DPO focused on improving the voltage based methodology of predicting accident leakage from eddy current voltage measurements.

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• 2.s. 2065 M0C=w-"4 '51" &&% ___ ________ ________Table I -Tube Ruptures in US Plants Excluding Major Tube Leaks The DPO was resolved by the promulgation of new tube performance criteria in 2007.These criteria are strictly based on predicting the probability of tube rupture during various accidents and the related leakage from Bobbin voltage measurements.

Such predictions are not applicable when the mechanism of tube rupture is fretting fatigue.Voltage based methodology of leakage predictions does not bound fretting fatigue leakage because the latter results in an instantaneous circumferential tube rupture.The reason why the leakage from SCC cracks is fundamentally different than leakage from tubes that exceeded their allowable fatigue life can best be illustrated by considering the design basis MSLB. Thermal Hydraulic (T-H) analysis shows that the pressure differential AP across a tube initially increases slowly and therefore even if several tubes contained a very large number of cracks they would open slowly minimizing the primary to secondary leakage. It is only later during the event when the SG has been emptied and the emergency core cooling, ECCS kicks in that the AP starts increasing.

At this time however, its relative value is small. In contrast a rupture of a fretting fatigued tube does not depend on AP, the change in the sudden increase in stress intensity. (Increase in DP would be important only if wall thinning due to fretting was reduced to below the burst thickness).

During the MSLB event, vibrations triggered by forces from outside or inside the steam generator vessel would be an obvious source for increases in local stress intensities.

High cyclic stress from FEI during the MSLB event would cause a small crack to rapidly propagate circumferentially to failure when the tube is near or at its allowable fatigue life. Leakage increase from the propagation of circumferential fatigue cracks was not addressed in the DPO and therefore is not included in the 2007 tube performance criteria.28 The NRC AIT report states that SCE informed the NRC that "they are reviewing their calculations of the LERF (4E-6/yr) and believe that review that will likely indicate that the differential pressures generated by a steam line break would not be large enough to rupture the degraded tubes as long as operators successfully implemented their emergency procedures.

If this is confirmed, the risk associated with steam line breaks will be significantly reduced." Such a conclusion would be only valid if the tubes had not been damaged by fatigue. Since this is clearly not the case, SCE hope for lowering the LERF is unrealistic.

The stress intensity during the MSLB can best discussed in terms of the Stability Ratio (SR) which is an indicator of the FEI vibration intensity.

SCE calculated that the SR varies from 0.33 to 1.15 and 0.16 to 0.83 for 100% and 70% power respectively, depending on tube location.

Such reduction in the SR may be significant for steady state operation but is insignificantly small compared to the increase in SR on depressurization of the SG during the MSLB accident.

The corresponding increase in velocity and steam quality overshadows the reduction in these parameters by operating San Onofre reactor Unit 2 at 70 % power. Therefore, the reduction in SR has no relevance to accident analysis when the tubes entering cycle 16 have been damaged by fatigue.A second factor that distinguishes tube failures by SCC and high cycle fretting fatigue is the difficulty of detecting the latter during in-service inspections.

This is because the crack initiation phase constitutes a high fraction of the total fatigue life in high cycle fatigue, once an engineering crack has been initiated, fracture occurs abruptly when the intensity level is sufficiently high (Ref. 7, 8, 9). The DPO project invested considerable effort on improving the sensitivity of eddy current detection of SCC cracks for leakage predictions.

Since comparable data for predicting leakage from fatigue induced cracks does not have any safety analysis that is based on fatigue failures one cannot use the 2007 performance criteria to ensure safety. This is a reason why the SCE safety analysis is not valid and why it must be re-evaluated in terms of fretting fatigue induced leakage instead SCC induced leakage.Since it took more then 15 years to develop the SCC based leakage methodology and close-out the DPO (and the related GSI 163) it cannot be expected that the NRC will revise the existing performance criteria any time soon. Until that time, conservative assessments must be performed before nuclear plants with considerable fretting fatigue damage are allowed to remain in service. The SCE safety assessment is not conservative.

Therefore, before starting San Onofre reactor Unit 2 at any power level, SCE must formulate an approach that would assure that the public safety margins would not be decreased.

SCE can use any method for that purpose as long as it can defend it on a technically conservative basis. The following five accident scenarios are discussed to provide further insight why operation with pristine tubes at 100% power, current 29 licensing base (CLB), is drastically different than operation at 70% power with fatigue damage tubes.2. Accident Scenarios The Steam Generators in San Onofre reactor Unit 2 (SGE 88 and SGE 89) contain 482 and 563 tubes respectively, with AVB wear ranging from 10% to 34%. The two SGs also contain a total of 515-plugged tubes. These tubes act as multiple sources for leakage during normal operations and during accidents (Ref. 9). They must be considered as sources for causing accidents and sources for propagating the leakage intensity during the accident.

An assessment of operations with such degraded tubes must demonstrate that at any time during normal operations and during accidents their local gap velocities, the corresponding SR and the burst pressure, will remain at sufficiently low levels to prevent leakages from exceeding acceptable levels. The following accidents are examples of accidents which must be included in such assessments.

A. Spontaneous fretting fatigue rupture of a single steam generator tube in the free span with a stuck open relief valve or a broken header Steam Generator overfill occurs relatively frequently in PWRs, an assessment should consider that the DBA SGTR will cause the relief valve to be stuck open during this event. The resulting higher local gap velocities and the corresponding increase in the SR must not cause additional tubes, (both plugged and un-plugged) to rupture.B. Tube Ruptures from Unplanned closing of an isolation valve.Closing an isolation valve would lead to an increase in steam flow through the unaffected SG. The corresponding increase in gap velocity would increase the local SR causing tubes which are on the border to exhausting their fatigue life to rupture abruptly (Ref 7, 8). This accident is similar to case A above with the exception that the increase in SR is expected to take place at a slower rate.C. Seismically -Induced Tube Rupture Both plugged and unplugged tubes can potentially lead to large primary to secondary leakage. Plugged tubes would behave differently, firstly because they do not generate a failure signal at the steam ejectors, and secondly, because the natural frequency of a broken tube would be lower than that that of an in service tube.Reactor experience (Ref. 9) has demonstrated that tubes that have been plugged due to wear will continue to wear and eventually break to impact and damage adjacent tubes.Material loss by wear not the mode of failure at plants was studied by EPRI. In their studies combining tube swelling with Fluid Induced Vibration (FIV) led to 30 circumferential fatigue failure. The difference between the cases studied by EPRI and the plugged tubes at SONGS is that at SONGS some plugged tube have already suffered considerable fatigue damage prior to plugging and are prone to fatigue failure. In this regard, EPRI recommends that tubes with pre-existing circumferential cracks be evaluated using linear elastic fracture mechanics.

Because some tubes at SONGS used up a significant fraction of their fatigue life they may contain micro cracks of various size.Because such cracks have not been detected at SONGS there is no indication that they do not exist. SCE did not address this issue.EPRI did not assess the effectiveness of tube stabilization in preventing damage to adjacent tubes; neither did SCE provide any information on their criteria for selecting tubes for stabilization.

SCE conclusions that the combined forces of the differential pressure and the seismic loads would not cause any tube to burst cannot be justified when the tubes are also subjected to cyclic loads simultaneously.

SCE calculation are based on the tensile strength that would cause tube rupture, a much lower stress, less than half, would be sufficient to severe tubes with cumulative fatigue usage (CUF) near unity (Ref 8)SCE calculations are based on a non conservative model and therefore their conclusions in the FSAR (5.4.2.2.1.3) regarding the ability of degraded tubes to withstand seismic loads are not valid.D. Station Blackout, SBO Severe accidents are not considered design basis accidents, nevertheless when changes in system operations are contemplated those changes must not increase safety risk. The operation of San Onofre reactor Unit 2 with a large number of fatigued tubes 'represents a new accident that has never been previously analyzed.'

All the analysis to date was based on tube failure by creep at high temperature.

The fact that the tubes were fatigued damaged demonstrates they can fail earlier due to natural flow instabilities in the steam generator.

The SBO accident is briefly described below.In this accident the primary system remains pressurized following a core becoming uncovered.

In the station blackout, SBO, accident scenario after the core is uncovered the secondary sides of all four steam generators are dry while on the primary side, steam flow by natural convection from the core to the steam generators and back to the core.The high pressure, high temperature steam will cause the weakest component in the system to fail thereby depressurizing the primary side. In this regard the hot leg surge line and the SG tubes are the weakest components in the reactor coolant system. If the high hoop stress on the hot leg surge line causes it to fail, the release of the highly radioactive gases will be contained within the containment.

If on the other hand, the high pressure high temperature steam opens up existing cracks in the steam generator tubes or ruptures 31 the tubes the primary side will be depressurized, by-passing the containment and allowing the highly radioactive gases to escape directly to the environment through the SG relief valve. The above scenario, also known as the high/dry core damage sequence, represents an early containment failure, which significantly increases the large early release frequency (LERF). When the containment fails early, the release to the environment is several thousands times larger in comparison to the release when the containment is intact. Most importantly, this early release occurs prior to the evacuation of the close population and therefore may cause early health effects (prompt fatalities).

Conformance to 10 CFR 50, Appendix B Criterion 16 dictates that operation with fatigued tubes will not increase the probability that fatigued tubes will not fail before the surge line. Appendix B dictates that to maintain its licensing basis the licensees must provide measures to assure that conditions adverse to quality such as failures, malfunctions, deficiencies, deviations, defective materials and equipment, are promptly identified and corrected.

Fatigued tubes definitely represent conditions which are adverse to quality.E. Main Steam Line Break, MSLB The fact that San Onofre reactor Unit 2 can pass the existing performance criteria from the in-situ tests results of San Onofre reactor Unit 3 provides no assurance at all that during a spontaneous MSLB accident the leakage will not exceed the DBA leakage. The in-situ tests only show that if the tubes were only exposed to tested pressure they would not leak if they maintained their wall geometry as tested. The in-situ tests were intended to determine leakage on the basis of tube weakening by actual loss of material and inclusions of stress corrosion cracks. In contrast to static pressure tests, fatigue failure due to high cycle FIV would result in a fast propagating circumferential crack at relatively low stresses (Refs 1, 5). Leakage from degraded tubes must be assessed in terms of the mechanism that has the potential to cause the largest leakage.If SCE wants to base their calculations on a realistic accident scenario, it must first demonstrate that the wear equation that was developed for laboratory data would be applicable to a tube that experienced impact wear in the SONGS steam generators.

As discussed in Appendix A, the wear equation which was used by SCE to calculate wall thickness did not properly incorporate the effects of impact wear. Secondly and more importantly, SCE must demonstrate that their burst pressure mechanism of determining leakage is conservative in comparison to the leakage that would occur during the fast MSLB depressurization.

The fast depressurization of the secondary side following an MSLB will lead to rapid increases in local gap velocity steam quality, thereby significantly increasing the stability ratio SR. The higher SR would, in term, increase the stress on the tube leading to rapid circumferential crack propagation as occurred in North Ana (Ref 2 )32 F. Risk Considerations The unusually large tube damage exhibited in both steam generators at San Onofre reactor Unit 2 is unprecedented, therefore little data is available to assess the increase in safety risk that would be associated with the above five accident scenarios.

Consequently accident assessments must be based on conservative assumptions.

The main uncertainty that must be considered in arriving at a risk estimate is the ability of the operator to shut the reactor down in a safe manner before depleting the RW Storage Tank inventory.

Operator's success would depend primarily on the unpredictable increase in leakage in an environment experiencing violent vibrations due to secondary side depressurization.

Operators are not trained in simulators that can reproduce such environments.

In my judgment, based on computer calculations, an operator would not be able to prevent the reactor core from being uncovered if the number of tubes failures would exceed five.Given that steam generator 89 contain at least 500 AVB tubes which have used up a significant fraction of their fatigue life and another 86 TSP tubes (Ref 10) which also lost some fatigue life, a rupture of 5 tubes out of 600 susceptible tubes as result of fatigue failure during an MSLB event is not an overly conservative assumption.

Taken the probability of a steam a line break outside containment at E-4 per year the Large Early Release Frequency, LERF of radiation escaping the environment due to the reactor core being exposed becomes 1E-4 /year or 5 E-5 /yr for six month of operation.

Such an increase is by a factor of 5 higher than Commission goals as described in Reference

11. In contrast, SCE calculated a change in LERF of 4E-6/yr on the basis of that the AP will not be large enough to rupture tubes.The LERF is a measure of risk, the safety goal takes into consideration that the LERF must be by an order of magnitude lower than the core damage frequency (CDF) to account for a large and early radiation release due to containment bypass.As discussed above, when the controlling mechanism of tube rupture is cyclic stresses from FIV, tube rupture will be controlled by variations in the stability ratio SR and not by variations in A P. During the MSLB the SR will be drastically increased due to an increase in local velocities and steam quality.G. Summary The reason that SCE concluded that operation of San Onofre reactor Unit 2 at 70% power would not involve a new unanalysed accident was because SCE assumed that the tubes would enter service in cycle 17 in the same conditions as they were at the beginning of cycle 16. In addition, SCE implicitly assumed that the stability ratio would not increase during Design Basis Accidents and the burst pressure could be determined by ignoring 33 scaling effects in fretting wear by impacts. Based on ample reactor experience and laboratory data there is no basis to accept SCE proposed no statements to CFR 50.92.34 References I -SONGS Unit 2 Return to Service Report attachment 4 MHI Document L5-04GA564 Tube Wear of Unit-3 RSG Technical Evaluation Report [Proprietary Information Redacted], S023-617-1-MI538TREV.

0, submitted by Southern California Edison to the NRC, October 3rd 2012.2 -NRC Bulletin 88-02: Rapidly Propagating Fatigue Cracks in Steam Generator Tubes, Feb 5, 1988 3 -Application of Risk Assessment and Management to Nuclear Safety George Apostolakis Commissioner US Nuclear Regulatory Commission DOE Nuclear Safety Workshop September 20, 2012 4 -PWR Steam Generator Tube Fretting and Fatigue Wear, EPRI- 6341 April 1989 5 -Volchock et.al. "The effect of Surface regular micro-topography on fretting fatigue life. Wear 253 2002 509-515 6 -H. J. Connors et al. Watts Bar Unit I Evaluation For Tube Vibration Induced Fatigue, April 190 WCAP- 12547 7 -F.A. Simonen and S. R, Gossein "Life Prediction of and Monitoring of Nuclear Power Plant Components for Service Related Degradation" Trans ASME Vol 123 Feb.2001 8 -Case Study of the propagation of a small flaw under PWR loading conditions and comparison with the ASME code design life G.T. Yahr et al ORNL Conf 8607622 -12 9 -Three Mile Island Plugged Tube Severance, May 2003-EPRI 10 -SCE, " San Onofre Nuclear Operating Station Unit 2 Return to Service Report, Oct.3, 2012" 11 -USNRC G. Aposttolakis "Application of Risk Assessment an Management to Nuclear Safety " DOE Workshop, Sept 20, 2012 35 I declarc, under penalty oftpcrjury, that the foregoing infonnation is true, accurate, and correct. Executed on Ma3 15, 2013, in Rockville, MD.36 ATTACHMENT 2 Declaration of John Large UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE NRC STAFF In the Matter of Docket ID NRC-2013-0070 SOUTHERN CALIFORNIA EDISON COMPANY (San Onofre Nuclear Generating Station, Units 2 and 3) May 16,2013 A REVIEW OF THE NRC1 PROPOSED DETERMINATION OF No SIGNIFICANT HAZARD CONSIDERATION FOR SOUTHERN CALIFORNIA EDISON'S PROPOSED LICENSE AMENDMENT FOR SAN ONOFRE UNIT 2 DECLARATION OF JOHN LARGE I, John Large, being duly sworn, state: I QUALIFICATIONS AND EXPERIENCE 1.1 I am John H Large of the Gatehouse, 1 & 2 Repository Road, Ha Ha Road, Woolwich, London, United Kingdom, SE18 4BQ.1.2 1 am a citizen of the United Kingdom.1.3 1 am a Consulting Engineer, Chartered Engineer, Fellow of the Institution of Mechanical Engineers, Graduate Member of the Institution Civil Engineers, Learned Member of the Nuclear Institute and a Fellow of the Royal Society of Arts.1.4 I head the firm of Consulting Engineers, Large & Associates.

15 Based in London UK, Large & Associates provides engineering and analytical services relating to nuclear activities, systems failure and engineering defects.1.6 Prior to founding Large & Associates, from the 1960s through to the early 1990s I was a full time, tenured academic in the School of Engineering of Brunel University (London) where, as a Whereas I acknowledge that this declaration relates to NRC's finding on the no significant hazard consideration, I have channelled my comments through SCE because the request for the license amendment derives from SCE and the NRC proffers no statement of its own position.

Senior Research Fellow, I undertook applications research on behalf of the United Kingdom Atomic Energy Authority (UKAEA) and other UK government agencies.1.7 A r1su6rn of my academic and professional consulting careers is available at the Large &Associates website.2 EXPERIENCE OF THE SAN ONOFRE NUCLEAR GENERATING STATION 2.1 I have previously prepared and submitted evidence in the matter of the San Onofre Nuclear Generating Station (SONGS) to the Nuclear Regulatory Commission (NRC) Atomic Safety Licensing Board (ASLB).2.2 In my It Affidavit (January 2013) to the ASLB I provided opinion on the failings of the SONGS replacement steam generator (RSG) design, how this gave rise to unrestrained tube motion and excessive tube wear, and on the uncertainties of restarting Unit 2 at the proposed maximum limit of 70% rated thermal power (RTP). In my 21(1 Affidavit (February 2013) to the ASLB I examined the RSG steamside thermal-hydraulic flow regime and how this determined the types and rates of tube and tube restraint component wear, particularly at the Southern California Edison (SCE) proposal to operate Unit 2 at 70% rated thermal power (RTP).2.3 I have also prepared and submitted opinion (March 2013) to the NRC Petition Review Board in which I review the involvement of SCE and Mitsubishi Heavy Industries (MHI) in the specification and design of the RSGs.2 3 LICENSE AMENDMENT REQUEST AND NO SIGNIFICANT HAZARD CONSIDERATION (NSHC)3 3.1 SCE has submitted a license amendment request for a temporary change to the steam generator management program and license condition for maximum power, both being integral parts of the OL Technical Specification (TS). In short, the amendment applies for the duration of the fuel cycle (Cycle 17) in that power operation would be restricted to up to 70% rated thermal power rating (RTP) and that a tube inspection would be undertaken at 150 days of operation into Cycle 17. Other than the power reduction and tube inspection period, no other physical 2 Tubte Wear ldentified ini the San Onofre Replacemele Steam Generators Mii.nhi.'hi Reports ULS-20120254 Rev.) (3/64)and L5-04ga5,Mt(O)

Tonether with Other Relevant hilormnti, m, March 2013 -this supplementary report was placed before the Petition Review Board of the NRC as part of the §2.206 process.3 NSHC is required under 10CFR §50.91 and §50.92 and the Regulatory Issue Suinx3, (RSI)Attributes

,?f. t Prolposed No Stýlifiovlft Ha:tzda Conshideration D[lerininatli ,n (March 29 2012) provides the public an opportunity to comment or request a hearing on the proposed amendment request via the published NRC Notice of April 16 2013 Apllicatioit (1i IAineminent to Fttsilif' Open aing hI' h'ing Prrop nved No

(.th atCnstderatiic it Otelotcic itt W": &1110 l.fte/n Nut mewu Generaiiitg Station. Ufnt 2.

changes to the operation and/or detailed installation of the components of the plant were proposed.3.2 A No Significant Hazard Consideration (NSHC) determination requires that operation of the facility in accordance with the proposed amendment would not 3.3 (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or 3.4 (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3.5 (3) involve a significant reduction in a margin of safety.4 EXTANT CONDITION OF THE UNIT 2 STEAM GENERATOR TUBING 4.1 My 1 St Affidavit to the ASLB describes the processes that gave rise to the extensive degradation of the tubes in each of the two RSGs serving SONGS Unit 2 and, similarly, Unit 3 -the events, processes and the extent of tube wear degradation have been extensively reported by a number 4 of sources.4.2 A summary of the Units 2 and 3 RSG tube wear is given in APPENDIX I of my 1 st Affidavit being a true reproduction of Table 6-1 of the SCE response 5 to the CAL of October 2012 -detailed inspection data for Unit 2 tube wear is provided by the SCE Special Report.6 4.3 The OL TS sets out criteria stipulating the condition in which individual tubes have to be 7 withdrawn from pressurized service by plugging.

Essentially, these tubes are: 4.3.1 a) First, in those instances where the tube wall thinning is equal to or greater than 35% of the original tube wall thickness (TW Depth -Column I of Table 6-1) the tube has to be withdrawn from service (by plugging).

4 There are a number of chronological narratives of the events leading up to the withdrawal of all 4 RSGs at SONGS, for example United States Nuclear Regulatory Commission Region IV, San Onw Jre Noclar Generatitg Sittiott NRC Aogt,nented ho.pectioui Team Report 05001836212012007, July 18 2012, also the SCE. Enclosure 2, Songs Return to Service Report, October 3 2012 and Attachment 4: MHI Iocument L5-t4GA564

-Tubc.Wcar of Ulit-3 RSG Technical Evaluation Report. Mitsubishi Heavy Industries S023-617-1 -M 1538 Rev 0.5 SCE, Enclosure 2, SONGS Return to Service Report, October 3, 2012 6 SCE Special Report, Ihspection of Steam Generator Tubes, Cycle 17, San Onafre Nuclear Generating Station, Unit 2, Docket No 50-361, 10 CFR 50.4, April 10, 2013 7 Localised repair of degraded tubes by sleeving is not pemaitted by the TS.

4.3.2 b) Second, the TS requires that the operational assessment, made at the time of the tube inspection, has to provide assurance that no individual tube will wear beyond the 35% TW Depth limit in the following fuel cycle or until the next tube inspection.

This means that the OA has to project the tube wear rate(s) forward over the next in-service cycle -for Unit 2 this forward projection applies to a period of 150 days into Cycle 17 being the tube inspection interval nominated by SCE as a condition of the restart.4.3.3 c) In addition to those tubes that have experienced or are likely to develop excess levels of wall thinning (=>35%), SCE has also chosen to preventatively plug zones of tubes to reduce the risk and incidence of tube wear.4.4 These three groups of plugged tubes make up the numbers of defective1 7 tubes of the tube bundles of the two RSGs serving Unit 2. Those tubes that have sustained some degree of wear to a depth of less than 35% and which remain in pressurized service are referred to as degraded1 7 tubes.5 ADEQUACY OF THE OPERATING LICENSE TECHNICAL SPECIFICATION 5.1 The TS tube integrity criteria relate tube resilience (integrity) to the remaining cross sectional (wall) area of then thinned tube section and the radial stress active within the wall. The radial or membrane stress derives directly (and solely) from the pressure differential between the reactor circuit and lower pressure of the steamside of the RSG. The resilience of a degraded tube is evaluated against allowable stress limits (membrane plus some in-plane bending stress) by multiplying the stresses for a non-degraded tube by the ratio of the corresponding sectional properties (thinned) of the degraded tube.8 5.2 In my opinion this approach, adopted for the TS tube integrity criteria, is overly simplistic in that it provides little account for anomalies (aging, chemical deterioration, etc) of tube material and/or physical degradation of the tube geometry (in this case surface imperfections and flaws).5.3 In effect, the allowable stress limits based on a pressure bursting failure mode provide, so it is assumed, sufficient margin to accommodate all other (undefined) processes and conditions that 8 For the degraded tube case evaluation the minimum tube wall thickness required to meet the structural requirements of UNSCR R.G. 1.121 is calculated by considering (1) wall thickness loss over the entire tube length, (2) wall thickness loss at the tube intersections with tube support plates (TSPs), and (3) wall thickness loss at the tube intersections with the anti-vibration bars (AVBs) in the tube bundle U-bend region and the minimum wall thickness is calculated for (a) the fault condition, and (b) the normal operating condition.

The more limiting of these two loading conditions determines the minimum allowable tube wall thickness for the tube not to burst under the conditions specified in R.G. 1.121.

could contribute to and/or accelerate failure of the tube -this is the basis of the TS criteria underwriting tube integrity.

5.4 In other words, the underlying premise is that at 35% thinning (for whatever reason and by whichever means), the tube remains a sound structure, there being sufficient margin in hand to safeguard against all other circumstances and conditions that might quite independently progress to tube failure. Moreover, the failure mechanisms of such other 'independent' factors, for example plain fatigue cracking of a vibrating tube, might themselves be enhanced by the nature of the tube wear, either by the presence of surface flaws, abrasions, notches and/or areas of work hardening.

9 5.5 Other than the margin, the TS tube integrity criteria do not provide for quantitative cross linkage of possible separate failure mechanisms to the condition of the tube surface as generated by the degradation processes (tube rubbing, abrasion and impacting) experienced in the San Onofre RSGs.5.6 For example, the 35% tube wall thinning threshold at or over which individual tubes should be plugged and withdrawn from pressurized service, must be drawn from operational and bench-testing experience of past tube failures.

This is because the 35% threshold must provide a satisfactory margin to cover metallurgical and physical geometry features that serve to trigger various failure modes, such as stress corrosion cracking,'

0 mechanical damage, wastage (thinning), denting, and vibration induced cyclic plain fatigue cracking.5.7 However, the San Onofre tube degradation is acknowledged to be unique so, it follows, that the nature of the tube degradation scars are also likely to include unique features that are not replicated in the data bank of past tube failures at other nuclear plants and from bench-testing trials. If so, the use of the present single-failure mode TS tube integrity criteria (that relies upon 9 The impact behavior and fracture response of Inconel 690 has not attracted that much research, although indications are that impacting results in the formation of localized shear bands which can prompt catastrophic failure -see Woei-Shyan Lee and Tai-Nong Sun, Plaslic Flow Behaviour

(?!'lncone1 690 Super Alloy Under Compressive Impact Loading, Materials Trans, Vol. 45, No. 7, 2004 10 The are number of such modes of failure including, but the TS is heavily biased towards stress corrosion cracking which had become by the 1990s the principal degradation mechanism for SG tubing worldwide.

For example outside diameter stress corrosion cracking (ODSCC) where the probability of failure is determined from proprietary coefficients obtained by bench-testing -for example, EPRI suggest the failure function for ODSCC to be Ap 1 (a) = A + B.logl0(a)

+ E where A and B are proprietary coefficients and Apla) is the burst pressure for a given flaw of a dimension.

A common locality for ODSCC is at the TSP where debris, comprising corrosion sludge fills the TSP aperture providing conditions conducive to dry-out and adverse cation/anion being an accelerant to inter-granular SCC and crack linkage. Similar ODSCC is known to occur at dented TSP locations.

Alloy 690 tube material is generally more (about 10 times) resistant to SCC than the earlier Alloy 600.

past experience of tube failures at other nuclear plants, etc) at San Onofre is inappropriate and introduced uncertainty.

I 1 2 6 NATURE OF THE TUBE DEGRADATION AT SAN ONOFRE 6.1 In its reporting' 3 to SCE, the manufacturers of the RSGs, Mitsubishi Heavy Industries (MHI), describes distinctly different patterns of wear at the TTW, AVB and TSP locations.

The in-plane direction of the tube vibration, particularly at the AVB-to-tube fretting, localities, is generally acknowledged to be unique to the San Onofre RSGs in that this mode of degradation has not been experienced at other US nuclear power plants wherein tube wear is dominated by out-of-plane motion.6.2 These variations in types of tube wear are described by MHI.3 6.2.1 TTW: This wear pattern occurs on the free span portion of the tubes (between the remaining effective AVB restraint points) in the U-bend region of the tube bundle. TTW produces long scars running in the axial direction of the tube as a result of continuous contact fretting or clashing of impacting tubes.6,2.2 MHI conclude that the tube in-plane motion giving rise to T1W is caused either by random vibration and/or fluid elastic instability (FEI), favoring the latter on the basis that the amplitude of random vibration is small.6.2.3 It is also possible that out-of-plane TT'W occurs and that this lower frequency vibration mode is excited by low frequency flow induced forces from, for example, vortex shedding in the wake of tubes or groups of tubes. In this mode, even if the in- and out-of-plane FEI is suppressed in the 70% RTP restart of Unit 2, the tubes will remain vulnerable to excitation by flow induced fluid forces.I I The EPRI Steam Generator Eamnination Guidelines Revision 5 state that flaws in qualification data sets should produce signals similar to those observed in the field in terms of signal characteristics, signal amplitude, and signal-to-noise level.12 For example, MHI describe a zigzag pattern wear scar -see ¶63.1 -which might provide a stress raiser in the axial tube direction in which the pressure membrane stress acts, although insufficient description of this type of wear scar is publicly available.

13 Attachment 4: MHI Document L5-04GA564

-Tube Wear of Unit-3 RSG Technical Evahlation Report, Mitsubishi Heavy Industries S023-617-I-M1538 Rev 0.

6.3 AVB: Wear at the tube-to-AVB contact points wear occurs in three distinctive patterns: 6.3.1 In-Plane:

To generate this pattern of wear at the AVB the tube (shown right) moves relative to the AVB in the in-plane direction (up-and down). The resulting wear scar sits across the AV bar depth indicating relatively large in-plane motion Wear amplitude.

Scar 6.3.2 This wear arises because the 'zero tube-to-bar gap and zero-AV Bar preload design functionality of the AVB provides no tube restraint the in-plane direction leaving the tube free to respond Adapted from MH1 and slide (up and down) across the AV bar contact surface.6.3.3 This pattern of wear scar is much longer than the typical case adopted in the Updated Final Safety Analysis Report (UFSAR)1 4 for which the scar length is assumed to equal the AVB-to-tube contact length (ie the AV bar cross-section depth). In this case the UFSAR is overly optimistic in determining the permissible tube wall wear depth because it is generally accepted that tubes with shorter wear scar lengths exhibit higher burst pressures.

6.3.4 A variation of this in-plane motion is a zigzag or saw-tooth surface wear pattern suggesting a combination of in- and out-of-plane tube motion. This pattern of tube wear produces a distinctive line flaw orientated in the axis of the tube thereby presenting a weakness in the tensile direction of the principal tube stress arising from the pressure differential.

6.3.5 AVB Dig In: In this pattern of wear the misaligned or twisted AVB digs-in to the tube surface -the pattern is probably unique to Unit 2 because the Unit 3 AVBs were more effectively flattened by a modified manufacturing process -see T7.12.UNIT 2 UNIT 3 Twisted AVBI Flattened AVBI 6.3.6 The resulting wear scar is a sharp notch or 'stress raiser'in T Alatt d 11.....the surface of the tube.14 San Onofre Nuclear Generating Station Unit 2 & 3 Updated Final Safety Analysis Report Revised April 2011 -see Table 15.10.63.2-4 for the transient analysis summary results for a steam generator tube rupture.

6.3.7 For failure analysis, account of this stress raiser is taken by assuming a stress concentration factor (k 1) determined by the dimensional geometry of the notch. Since the detection of this wear is blind from within the tube it is impracticable to Twb'tdJ Notch determine the sharpness and depth of the notch so, it follows, AVT,, the appropriate value of k, cannot be chosen with absolute1,0,,o MNI 0.011*r certainty.

5 6.3.8 A variation of this wear pattern is where there is both in- and out-of-plane movement of the tube to produce the zigzag pattern described earlier (16.3.4), similarly producing an axial flaw that presents to the radial tensile stress in the pressurized tube wall.6.3.9 In-Plane AVB and Tube Wear: This wear pattern is where both tube and AVB bar have both worn simultaneously or when the wear between tube and AV bar cannot be distinguished because visual access to the wear interface is not possible, although the left-hand tube in the example shown right, the tube motion has worn substantially through the width of the AV bar ('-40%).6.4 All modes of AVB-to-tube wear are provoked by fluid flow random vibration (ie turbulence) and, thus, the AVB-to-tube contact locations remain vulnerable induced excitation and wear even if the 70% RTP eliminates FEI.1 6 6.5 These two latter wear patterns (T6.3.6 and T6.3.9 -as shown by photographs reproduced from the MHI inspection) 1 3 highlight the difficulty of accessing the wear scars to determine the extent of surface damage. Much the same applies to the locations of the TTW, where the close proximity of adjacent tubes practicably limits access for visual inspection.

6.6 In other words, although the eddy-current (El) in-service through-wall inspection results provide a generally reliable measure of overall tube wall thinning, the assessment of the nature of individual incidences of tube surface damage (imperfections, etc) is uncertain!.'

15 For example the range of stress concentration faictor kt t, given in Chart 3.5 of Walter D. Pilkey, Peterson's Stress Concentration Factors Second Edition, John Wiley, Sons, Inc., 1997 for a tube under axial tension (but not pressurized).

16 Although not discussed here, I consider it likely that the TSP-to-tube wear is also driven by random fluid processes

-this locality of tube wear is an important factor in considering the potential for the tubes that are effectively pinned at the top TSP but with successive AVB-to-tube restraint not active, to fail by high cycle fatigue.

6.7 So far I have considered the surface changes brought about by, for want of a better description,'gouging' of two adjacent parts (eg tube-to-tube, AVB-to-tube and TSP-to-tube) to form distinctive scars or stress raisers in the tube outer surface. Under tensile loading, deriving either from internal pressure, bending or plain cyclic fatigue, the stress concentration can develop cracking resulting in early tube failure from ductile tearing or brittle fracture.6.8 Thus the presence of surface flaws produced in TTW, AVB- and TSP-to-tube wear may bring forward tube failure before that predicted by the TS tube burst criterion (3xAP and 1.4xAP for SIPC and AILPC cases respectively).

Since the TS does not specifically refer to this and other types of tube failure, it vital to maintain the 3xAP and 1.4xAP margins to cover such contingencies.

6.9 There

is another strength of materials phenomenon, referred to as fretting fatigue, occurring at the contacting and sliding surfaces of two adjacent parts under load and subject to slight relative movement by vibration or some other force. At very low stress levels and often after only a few thousands of cycles, fretting fatigue may initiate micro cracking in the rubbing surfaces that then become available to propagate into ductile/brittle failure zones (as in 56.7).6.10 In plain fatigue (without fretting) the initiation and development of small cracks typically represents upwards of 80 to 90% of the total component life, but with a fatigue fretting contribution the plain fatigue strength or endurance limit can be reduced by as much as 50 to 70% during subsequent (or simultaneous) cyclic loading of the tube overall.8'19 6.11 As I previously noted in 56.6 through-wall Er may not have sufficient resolution to detect fine micro cracking between surfaces in contact (ie the tube and TSP or AVB or another tube) and, if so, the presence of established fretting fatigue may have passed unnoticed.

17 The ET inspection system must detect tube wall internal and surface flaws at an acceptable level of detection reliability and it must also size the significant flaws. For the San Onofre degraded tubes Er must have acceptable reliability to detect and size flaws which are not necessarily significant but which might require action to mitigate further tube damage. This grading of flaws determines whether the tube is degraded but fit for continued pressurized service or defective because it contains a flaw of such severity that it is unacceptable for continued pressurized service until the next tube inspection outage.18 ASPM Handbook V 19, Fatigue and Fracture, ASM International 19 Plain fatigue is where there is no direct contact, say where the pipe vibrates in a free span situation.

In fretting fatigue there is contact between two slightly moving parts -the contact point stress gradients are likely to be very high due to the localised stress concentration at the contact and the magnitude of these stress gradients is usually much higher than those associated with typical design features of components, such as notches and holes. Also, loading is likely to be non-proportional in the neighbourhood of the contact with this feature caused by the non-linear nature of the friction at the contact interface.

Localised surface damage at the asperity level may play a role in accelerating the initiation of cracks at the asperity scale.

6.12 Moreover, because surface imperfections contribute strongly to premature failure (ie failure within the TS allowable stress safety margin), it is absolutely essential that the wear surface flaws (either deriving from gouging, fretting fatigue, or just plain fatigue) be fully understood and taken into account when projecting the tube integrity for the Cycle 17 in-service period..7 PROJECTING FURTHER TUBE DEGRADATION INTO CYCLE 17 7.1 Whereas the SCE license amendment request specifically applies to Unit 2 it is, nevertheless, important to consider the extent and nature of the tube wear degradation in the identical Unit 3.7.2 This is because the tube bundle damage in both RSGs serving Unit 3 is universally acknowledged to be so severe and extensive that any level of return to powered operation of this nuclear plant would introduce further risk and lack assurance that Unit 3 could operate safely.This situation is confirmed by the requirement of the Confirmatory Action Letter (CAL)2 0 issued by the NRC following the RSG tube failure that provoked the forced shutdown of Unit 3 in January 2012.7.3 The CAL specifically requires SCE to undertake a number of Actions relating to any intention to restart Unit 3, including 7.4 "... 7. Prior to entry of Unit 3 into Mode 4, SCE will submit to the NRC in writing the results of your assessment of Unit 3 steam generators, the protocol of inspections and/or operational limits, including schedule dates Jbr a mid-cycle shutdownnfor firther inspections, and the basis for SCE's conclusion that there is a reasonable assurance, as required by NRC regulations, that the unit will operate safely." my added emplhsis 7.5 On its part SCE has chosen not to respond to this Action 7 (and also four other CAL Actions)and no preparations have been made to restart Unit 3.2t In my opinion, this failure of SCE to respond to Action 7 is tacit acknowledgement that it is not possible to provide a 'reasonable assurance'that Unit 3 with the present level of tube degradation will 'operate safely'.7.6 This brings me to consider Unit 2 which SCE proposes to restart, subject to the license amendment request being accepted and ratified by the NRC. The CAL also required SCE to undertake specific actions, these being: 20 Letter from Elmo E Collins (USNRC) to Peter T Dietrich (SCE), (.Confirnualyt Later 4-12-001.

San Onofre Nuclear Generating Station. Units 2 and 3, Commitments to Address Steam Generator Tube Degradation, March 27 2012.21 In fact, the nuclear fuel of the reactor core of Unit 3 has been removed and placed in water pool storage.

7.7 "... 1. Southern California Edison Company (SCE) will deternine the causes of the tube-to-tube interactions that resulted in steanm generator tube wear in Unit 3, and will implement actions to prevent loss of integrity due to these causes in the Unit 2 steam generator tubes. SCE will establish a protocol of inspections and/or operational limits for Unit 2, incl.ding plans for a mi-cycle shutdownforffurther inspections.

2. Prior to entry of Unit 2 into Mode 2, SCE will submit to the NRC in writing the results of your assessment of Unit 2 steam generators, the protocol of inspections and/or operational limits, including schedule dates for a mid-cycle shutdown for further inspections, and the basis for SCE's conclusion that there is reasonable assurance, as required by NRC regulations, that the unit will operate safely." my added emphasis 7.8 In responding to the CAL, SCE presented a series of Operational Assessments (OAs) that it claimed justified restarting and operating Unit 2 for a trial period. The OAs of interest here are those by AREVA22 and, independently, the latest revision of the OA by Intertek APTECH.23 7.9 First, it is of interest to note that each OA fails to 'determine the causes of the tube-to-rube interactions' as stipulated by the CAL. This is because all of the OAs (including a third OA by Westinghouse)24 skirt round and stop short of identifying the root cause, delving no further into the design features, peculiarities and processes of the RSGs that give rise, so it seems uniquely at San Onofre, to the tube motion in-plane excitation forces.2 5 7.10 However, both OAs recognized that 7.10.1 i) the tube wear degradation experienced in Unit 2 was less advanced than the tube wear in Unit 3; although, that said 7.10.2 ii) the wear locations (AVB and TSP) and number of incidences present in Unit 2 were very similar to those present in Unit 3, although Unit 3 had, in addition, a much greater number of TTW incidences; and that 22 SCE. Attachment 6 -Appendix B: SONGS U2C17 -Stean Generator Operational Assessment for Tube-to-Tube Wear, AREVA 23 SCE, Enclosure

1. Amendment I Operational Assessmentfor SONGS Unit 2 Steam Generators for Tube-to-Tube Wear Degradation 100% Power Operation Case, Intertek AES 13018304-2Q-1 March 2013, March 14,2013.24 Attachnent 6- Appendlix D: Operational As.es,,ment of Wear Iulications In the U-bend Region of San Unit 2 Reph.iceieni Steam Generators, Westinghouse Rev 3, October 2012.25 In fact, as I discuss in some detail in my I' ASLB affidavit, there is disagreement between the various OA consultants as to whether the tube motion excitation forces derive from random fluid processes (turbulence, downstream wake, etc.) and/or fluid elastic instability.

7.10.3 iii) the wear processes involved in these virtually identical RSGs related to the effectiveness of restraint provided by the anti-vibration bar (AVB) assemblies in the U-bend region of the tube bundle.7.11 It follows that the tube wear mechanism is a two-stage process whereby, first, the AVB contact with the individual tubes is worn away by vibration of individual tubes excited into in-plane motion by local fluid flow forces. Second, loss of the AVB restraint, and successive points of AVB restraints, results in a lengthening of the unsupported or free-span sections of tube to the extent, again by fluid forces, that the tube vibrates at low frequency and at sufficient amplitude to enable tube-to-tube clashing and, hence, accelerated tube-to-tube wear (17W)26 to occur.7.12 The TTW in the Unit 2 RSGs was less advanced than that of Unit <(--, , '3 because of the omission in the manufacturing process of the "goo' 0 0P Unit 2 AVB components that, quite fortuitously, resulted in the presence of an unintentional clanmping or preload force across the individual tubes -this preload force delayed the loss of the in- Ai plane (IP) AVB restraint and the onset of second phase TTW process in Unit 2.27 7.13 In other words, the tube wear extant in Unit 2 is representative of the first stage of the two stage mechanism that leads to accelerated TTW.7.14 The disadvantages arising from the installation of the distorted (twisted)

AV bars include i)notching and the formation of a stress raiser as shown at ¶6.3.7, and ii) the opportunity for fretting fatigue in the locality of the notch of i), generated whilst the AVB restraint was active with the preload force. Both of these features could contribute to localized ductile/brittle failure driven either by internal pressure and/or exceeding the modified plain fatigue cyclic loading endurance limit.7.15 SCE's proposed new operating regime for Unit 2 at 70% RTP claims that the 1st and 2 nd stages of wear will be slowed but not totally eliminated.

26 For a fuller description of the tube wear processes see my I ' Affidavit to ASLB.27 The intended design function of the AVBs was to provide a 'zero-gap-zero-contact-force

', that is no preload, across the individual tubes thereby offering no effective restraint in the in-plane direction.

Unit 3 achieved this design functionality following modifications in the manufacturing of the AVBs, whereas the unmodified AVBs of Unit 2 remained distorted (twisted) so much so that certain of the AVB locations applied a tube clamping preload force.

7.16 In fact, both AREVA and Intertek OAs agree that even at the 70% RTP reduction, the Ist stage process (AVB restraint loosening) will progress to the threshold at which the 2"d stage TTW commences in earnest, thereafter putting individual tubes at risk of structural failure (bursting) as a result of tube wall thinning.7.17 Intertek's projection for rate of TTW 1 "R"* l 70% RTP for the proposed Unit 2 restart (Cycle 17) is summarized by the following

, T.23~~~ _u I95%graphic:23 7.18 The 1 st stage AVB-to-tube restraint 005 opcR,,T, , .... l s .... ,00o5,. , .........loosening proceeds from the time of SCE PROPOSED 5 INTEERoETW o MONTH INSPECTION COMMENCES restart over a period of 12 months at PERIOD 0.4 1.0 1.6 which time a group of successive TIME INTO CYCLE17 -YEAR AVBs have little or no preload force in ATE, I..E.TE. FIG 1.5-1 the in-plane direction.

FIGURE 1 INTERTEK PROJECTION OF CYCLE 17 TUBE DEGRADATION 7,19 From this point forward the 2 1d stage wear process (TTW) commences following the characteristic

(-).7,20 The TFW wear rate is such that the tube passes the 95% tube burst criterion

(-)2 at about 0.35 years thereafter

-this means that the tube is considered to have failed at about 1.35 years into Cycle 17.7.21 SCE's argument is that the proposed Unit 2 shut down and RSG tube inspection at 5 months ()safeguards against tube failure by providing a sufficient time buffer ahead of the Intertek 1.35 year (16 months) period to exceed the 95% threshold.

7,22 However, Intertek's projections do not compare at all favorably with the timings evaluated by the AREVA OA 2 9 because: 28 The 95% Probability, 50% Confidence criterion for an individual tube burst is specified in the Operating License Technical Specification for Unit 2 of the San Onofre nuclear plant.29 Attachment 6 -Appendix B: SONGS U2C 17 -Steam Generator Operational Assessment for Tube-to-Tube Wear, AREVA -the data presented here relates to Figure 8-3 but this has been declared proprietary information and thus cannot be reproduced here -instead the set points of the AREVA AVB and TTW wear phases have been taken from the same but non-proprietary information available in the text of Appendix B -see Large & Associates Affidavit Response to 4tomnic SqaRnt and Licensing Board's 1-Factual Issues I January 22, 2013.

7.23 for AREVA (shown thus), the AVB-to-tube restraint loosening period until the 2 nd stage TTW commences is about is-0.3 year compared to 1 year by Intertek; and similarly 7.24 the equivalent total time to tube burst 0 (95% probability) projected by AREVA 0.is in the range of 0.5 to 1.5 year compared to -1.35 year Intertek.I AREVA TOTAL TIME TO INTRTEK I 95% BursT" 70% RTP I PROBABILITY RANGEC 0 0O ZZ Document No.UES-20120254 Rev. No.0 1 11 1 1 I e Ie 2 MITSUBISHI HEAVY INDUSTRIES, LTD.

U ES-20120254Rev.O (2/64) FNon-Proprietary Revision History Date Approved Checked Prepared Rev. Summary of Changes issued By by by 0 Original Issue See See See See Cover Page Cover Page Cover Page Cover Page UES-20120254 Rev.0 (3/64) INon-Proprietar Root Cause Analysis Report for tube wear identified in the Unit 2 and Unit 3 Steam Generators of San Onofre Nuclear Generating Station Root Cause Analysis Report for tube wear identified in the Unit 2 and Unit 3 Steam Generators of San Onofre Nuclear Generating.Station Attachment-3 UES-20120254 Rev.0 (47/64) Non-Proprietarý Change analysis For the SONGS RSGs, a change analysis was performed in two stages. The first stage compared the SONGS SG design to previous MHI SG designs for the triangular tube configuration.

MHI had previously performed three steam generator designs using a triangular tube configuration.

The second stage compared the SONGS RSGs to the previous SONGS SG design (Combustion Engineering type design). Only the most significant changes are included in this analysis.The change analysis results are set out below.(1) Differences between SONGS RSGs and previous MHI SG triangular design.--The SONGS RSGs have:* ( ) circulation ratio* ( ) maximum flow velocity* ( 3 average flow velocity:F I ]P/D ratio out-of-plane FEI stability ratio Largest U bundle radius* Specified AVB twist[ 3 ( )*( ]range of G-value (tube diameter, out-of-plane)

Highest steam quality (void fraction)* Thinnest and longest retainer bar* ( 3 nominal tube-to-AVB gap (0.002" cold / 0.000" hot)* ( 3 variation in tube-to-AVB gap (3 sigma[ 3)(2) Differences between SONGS RSGs and the previous SONGS OSG desig§n. --* Increase in tube bundle heat transfer surface area (11%)Increase in number of tubes (5%)Root Cause Analysis Report for tube wear identified in the Unit 2 and Unit 3 Steam Generators of San Onofre Nuclear Generating Station Attachment-3 UES-20120254 Rev.0 (48/64) Non-Proprietar!

• 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 (3) Identification of the changes from previous SG designs led to the recognition that the RSG design deserved close scrutiny.

MHI considered the changes in the SONGS design from previous steam generator designs and compared the basic design parameters of the SONGs RSGs (e.g., heat transfer area, circulation ratio, steam pressure, etc.) with other steam generator designs. Further, as part of the development of the SONGS RSG design, MHI conducted a detailed comparison between its proposed AVB support for the tubes in the U-bend region and that of a comparison plant of similar design. A special AVB team was formed and included industry experts to conduct an extensive design review process in 2005 / 2006 to optimize the U-bend design and address the technical issues. The team concluded that the SONGS design was significantly more conservative than previous designs in addressing U-bend tube vibration and wear.Also MHI and SCE recognized that the SONGS RSG steam quality (void fraction) was high and MHI performed feasibility studies of different methods to decrease it.Several design adjustments were made to reduce the steam quality (void fraction)but the effects were small. Design measures to reduce the steam quality (void fraction) by a greater amount were considered, but these changes had unacceptable consequences and MHI and SCE agreed not to implement them. It was concluded that the final design was optimal based on the overall RSG design requirements and constraints.

These included physical and other constraints on the RSG design in order to assure compliance with the provisions of 10 C.F.R. §50.59.Thus, MHI did compare the SONGs RSG design with previous steam generator designs, and in particular did a detailed evaluation of different options of the AVB design taking into account other large steam generator designs.Root Cause Analysis Reportfor tube wear identified in the Unit 2 and Unit 3 Steam Generators of San Onofre Nuclear Generating Station ATTACHMENT 6 Southern California Edison Co. (San Onofre Nuclear Generating Station, Units 2 and 3), LBP-13-07 (May 13, 2013) ("ASLB Order")

LBP-13-07 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD Before Administrative Judges: E. Roy Hawkens, Chairman Dr. Anthony J. Baratta Dr. Gary S. Arnold In the Matter of Docket Nos. 50-361-CAL, 50-362-CAL SOUTHERN CALIFORNIA EDISON CO.ASLBP No. 13-924-01-CAL-BDO1 (San Onofre Nuclear Generating Station, Units 2 and 3)May 13, 2013 MEMORANDUM AND ORDER (Resolving Issues Referred by the Commission in CLI-12-20)

In its November 8, 2012 decision in CLI-12-20, the Commission referred to the Atomic Safety and Licensing Board Panel (ASLBP) a portion of the June 18, 2012 hearing request filed by Friends of the Earth (Petitioner) challenging aspects of a Confirmatory Action Letter (CAL)issued by the NRC to Southern California Edison Company (SCE) on March 27, 2012.1 In particular, the Commission directed a duly constituted Licensing Board to "consider whether: (1) the [CAL] issued to SCE constitutes a de facto license amendment that would be subject to a hearing opportunity under [s]ection 189a [of the Atomic Energy Act (AEA)]; and, if so, (2) whether the petition meets the standing and contention admissibility requirements of 10 C.F.R. § 2.309." CLI-12-20, 76 NRC at _ (slip op. at 5).For the reasons discussed below, we resolve the first issue in the affirmative, concluding that this CAL process constitutes a de facto license amendment proceeding that is subject to a hearing opportunity.

Because this resolution provides Petitioner with all the relief its contention 1 See Southern Cal. Edison Co. (San Onofre Nuclear Generating Station, Units 2 and 3), CLI-12-20, 76 NRC _, _ (slip op. at 5) (Nov. 8, 2012).

2 seeks, the second issue referred by the Commission is moot, and the proceeding before this Board is therefore terminated.

I. FACTUAL AND PROCEDURAL BACKGROUND A. Factual Background The San Onofre Nuclear Generating Station (SONGS) is located near San Clemente, California.

2 SONGS Units 2 and 3 are pressurized water nuclear reactors with two steam generators per unit.3 SCE is the licensee for SONGS Units 2 and 3. See Brabec Aft. at 3-4.SCE's steam generators are recirculating, vertical U-tube type heat exchangers in which primary coolant is circulated inside the tubes, with heat from the primary-side coolant transferred to the secondary-side feedwater that circulates outside the tubes. This converts the feedwater into saturated steam that is used to drive a turbine-generator to create electricity.

See Brabec Aff. at 4.Steam generator tubes serve critical safety functions.

For example, they are an integral part of the reactor coolant pressure boundary and thus are essential for maintaining primary system pressure and coolant inventory.

They also isolate the radioactive fission products in the primary coolant from the secondary system.4 In September 2009, SCE shut down Unit 2 for a scheduled refueling outage and the replacement of its steam generators to resolve corrosion and other degradation issues in the original steam generators, which had been in service for nearly thirty years.5 SCE completed 2 See [SCE's] Brief on Issues Referred by the Commission (Jan. 30, 2013) at 3[hereinafter SCE's Answering Brief].3 See id., Att. 1, Affidavit of Richard Brabec (Jan. 30, 2013) at 3-4 [hereinafter Brabec Aff.). SONGS Unit 1 ceased operation in 1992 and has since been decommissioned.

See SCE's Answering Brief at 3.4 See SCE's Answering Brief, Att. 8 [SONGS] Unit 2 Return to Service Report (Oct. 3, 2012) at 14 [hereinafter Unit 2 Return to Service Report].5 See Brabec Aft. at 4; Unit 2 Return to Service Report at 10, 17; Letter from Ryan E.Lantz, Chief, Project Branch D, Division of Reactor Projects, US NRC, to Ross T. Ridenoure, 3 the Unit 2 refueling and steam generator replacement outage in April 2010, and that unit returned to full power in May 2010.6* In October 2010, SCE shut down Unit 3 for a scheduled refueling outage and the replacement of its steam generators, which also had been in service for nearly thirty years.7 In February 2011, SCE completed the Unit 3 refueling and steam generator replacement outage, and that unit returned to full power in March 2011.8 The replacement steam generators for Units 2 and 3, which were manufactured by Mitsubishi Heavy Industries (MHI) (see Brabec Aff. at 4), differ in design from the original steam generators.

9 For example, each replacement steam generator (1) has 9,727 tubes, which is 377 Senior Vice President and Chief Nuclear Officer, SCE, NRC's [SONGS] -Unit 2 Steam Generator Replacement Project Inspection Report 05000361/2009007 (Mar. 4, 2010), Enclosure at 5 (ADAMS Accession No. ML100630838).

6 See Letter from Ryan E. Lantz, Chief, Project Branch D, Division of Reactor Projects, US NRC, to Ross T. Ridenoure, Senior Vice President and Chief Nuclear Officer, SCE, NRC's[SONGS] -Unit 2 Steam Generator Replacement Project Inspection Report 05000361/20010008 (June 30, 2010), Enclosure at 3 (ADAMS Accession No. ML101810506).

7 See Letter from Ryan E. Lantz, Chief, Project Branch D, Division of Reactor Projects, US NRC, to Peter Dietrich, Senior Vice President and Chief Nuclear Officer, SCE, NRC's [SONGS]-NRC Integrated Inspection Report 05000361/2010005 and 05000362/2010005 (Feb. 10, 2011), Enclosure at 7 (ADAMS Accession No. ML1 10420223).

8 See Letter from Ryan E. Lantz, Chief, Project Branch D, Division of Reactor Projects, US NRC, to Peter Dietrich, Senior Vice President and Chief Nuclear Officer, SCE, NRC's [SONGS]-Unit 3 Steam Generator Replacement Project Inspection Report No. 05000362/2010009 (May 10, 2011), Enclosure at 3 (ADAMS Accession No. ML111300448).

9 See SCE's Answering Brief, Att. 31, NRC Augmented Inspection Team [AIT] Report (July 18, 2012) at 36 [hereinafter July 18 AIT Report]; see also Opening Brief of Petitioner Friends of the Earth (Jan. 11, 2013) at 1, 3 [hereinafter Petitioner's Opening Brief]; Petitioner's Opening Brief, Att. 3, Far Outside the Norm: The San Onofre Nuclear Plant's Generator Problems in the Context of the National Experience with Replacement Steam Generators at 4[hereinafter Hirsch Report]; Petition to Intervene and Request for Hearing by Friends of the Earth (June 18, 2012), Exh. 1, Declaration of Arnold Gundersen Supporting the Petition to Intervene by Friends of the Earth Regarding the Ongoing Failure of the Steam Generators at[SONGS] at 3 [hereinafter May 31 Gundersen Decl.].SCE urges this Board to discount the Hirsch Report attached to Petitioner's Opening Brief because, in alleged disregard of the directive in this Board's December 7 Order, Petitioner"did not provide an affidavit to support the factual assertions in the Hirsch Report, which are 4 more tubes than are in the original; (2) does not have a stay cylinder supporting the tube sheet;and (3) has a broached tube design rather than an 'egg crate" tube support.1 0 As discussed infra Part ll.B.2, a licensee must obtain a license amendment from the NRC if a change to its facility triggers the safety standards described in 10 C.F.R. § 50.59.Despite the design differences mentioned above between the replacement and original steam generators, SCE concluded that the replacements were a like-for-like change that did not require a license amendment.

1 1 On January 9, 2012, SCE shut down Unit 2 for a scheduled refueling outage and steam generator inspection.

1 2 On January 31, 2012, while Unit 2 was still shut down, Unit 3 operators received secondary plant system radiation alarms, diagnosed a steam generator tube leak of approximately 82 gallons per day, and shut down Unit 3 as required by plant procedures.

See relied upon throughout

[Petitioner's]

Brief." SCE's Answering Brief at 14. Petitioner counters that an affidavit was not necessary to support the Hirsch Report because (1) it "uses data submitted to the NRC by utilities operating nuclear reactors with replacement steam generators to compare San Onofre to the experience of [replacement steam generators]

nationally";

(2) it was "commissioned by Senator Barbara Boxer, Chair of the Senate Environment and Public Works Committee, and admitted into the Senate record in a joint hearing on September 12, 2012"; and (3) the NRC Commissioners "placed the Hirsch Report into the record of the Commission briefing on steam generator problems held on February 7, 2013 .... at which Daniel Hirsch was invited to testify." Reply Brief of Petitioner Friends of the Earth (Feb. 13, 2013) at 27-28 [hereinafter Petitioner's Reply Brief]. In these circumstances, and given that SCE does not identify particular factual errors in the Hirsch Report, we decline SCE's suggestion to disregard that Report.10 See July 18 AIT Report at 36; see also May 31 Gundersen Decl. at 4-6; Petitioner's Opening Brief, Att. 2, Affidavit of Arnold Gundersen (Jan. 9, 2013) at 8-9 [hereinafter Gundersen Aff.]; Petitioner's Opening Brief, Att. 1, Corrected Affidavit of John H. Large (Jan. 22, 2013) at 11[hereinafter Jan. 22 Large Aff.]. For a full description of the replacement steam generators, including a diagram, see Brabec Aff. at 4-5.11 See May 31 Gundersen Decl. at 7; Gundersen Aff. at 8. Although SCE did not seek a license amendment relating to the design differences of the steam generators, it did obtain a license amendment in 2009 for changes to certain "SONGS Technical Specifications related to steam generator tube integrity." SCE's Answering Brief at 6.12 See NRC Staff's Answering Brief in the [SONGS] CAL Proceeding (Jan. 30, 2013)[hereinafter NRC's Answering Brief], Att. 1, NRC Integrated Inspection Report 05000361/2012002 and 05000362/2012002 (May 8, 2012) at 18-19 [hereinafter May 8, 2012 Inspection Report].

5 May 8, 2012 Inspection Report at 39.SCE's inspection of the Unit 3 steam generators revealed "extensive

[tube-to-tube wear]" (SCE's Answering Brief at 9) that SCE determined "was caused by in-plane fluid elastic instability from the combination of localized high steam velocity, high steam void fraction, and insufficient contact forces between the tubes and the [anti-vibration bars]." Id. SCE states that more than 150 tubes of the 9,727 tubes in each [of the Unit 3 replacement steam generators]

experienced

[tube-to-tube wear], including more than 100 tubes in each [replacement steam generator]

with wear equal to or greater than 35% of the width of the tube wall (which is the criterion in SONGS Technical Specification 5.5.2.11 for removal of the tube from service by plugging of the tube).Id. (footnote omitted).1 3 Significantly, SCE acknowledges that "[tube-to-tube wear] due to in-plane [fluid elastic instability]

had not been previously experienced in U-tube steam generators." SCE's Answering Brief at 10. SCE describes fluid elastic instability as a phenomenon in which the tubes vibrate with increasingly larger amplitudes due to the flow velocity exceeding the critical velocity for a tube, given its supporting conditions and thermal-hydraulic environment.

[Fluid elastic instability]

occurs when the amount of energy imparted on the tube by the fluid is greater than the amount of energy that the tube can dissipate back to the fluid and to the supports.

During in-plane [fluid elastic instability], tubes within the same column are excited by the fluid and move with the plane of the column, resulting in tube-to-tube contact and wear of the tubes.Id. at 9 (footnotes omitted).With regard to Unit 2, SCE states, "[i]n contrast to the extensive

[tube-to-tube wear] in Unit 3, [tube-to-tube wear in Unit 2] existed in only a single pair of tubes ... in one of the two 13 As characterized by Petitioner, each Unit 3 steam generator "exhibited approximately 5,000+ indications of wear localities, with many tubes having wear indications at more than one locality and of differing degrees of wear severity, with a total of about 900 individual tubes affected in each [replacement steam generator]." Jan. 22 Large Aff. at 10. A total of 193 tubes in one steam generator and a total of 188 in the other exceeded the wall thinning threshold of 35%, above which tube plugging is mandatory.

See id. "Because of the depth and length of certain of the tube wear scars, a number of tubes were subjected to in situ hydrostatic pressure testing in March 2012, [which] resulted in 8 individual tube failures, all located in one[replacement steam generator]." Id.; see also Hirsch Report at 4-5, 7-9.

6... [steam generators]." SCE's Answering Brief at 9. One of SCE's contractors "concluded that the [tube-to-tube wear] in Unit 2 was not due to [fluid elastic instability], but instead to proximity of the tubes in question and random vibration of those tubes." Id. at 10. But other SCE analyses "assumed that [fluid elastic instability]

could occur in Unit 2 at 100% power." Id. SCE attributes the difference in tube-to-tube wear between Units 2 and 3 to fabrication differences arising from allowable fabrication tolerances.

1 4 See id. at 10, 92; infra note 43.On March 23, 2012, SCE submitted to the NRC Staff a "Steam Generator Return-to-Service Action Plan" and described actions it committed to take before restarting Units 2 and 3.15 On March 26, 2012, the NRC Staff confirmed, by telephone, its understanding of the actions to which SCE had committed.

See NRC Staffs Answering Brief at 3. On March 27, 2012, the NRC Staff memorialized its understanding in a CAL that confirmed the actions SCE would take prior to restarting either unit.1 6 As discussed in greater detail infra Part II.A.1, the NRC Staff uses a CAL to commence an enforcement process in which (as relevant here) a licensee agrees "to take certain actions to remove significant concerns regarding health and safety, safeguards, or the environment." 1 7 In 14 The extent of the tube-to-tube wear is described in the SONGS Unit 2 Return to Service Report's Steam Generator Operational Assessment for Tube-to-Tube Wear. See SCE's Answering Brief, Att. 12, SONGS U2C17 Steam Generator Operational Assessment for Tube-to-Tube Wear [hereinafter Assessment for Tube-to-Tube Wear]; see also Jan. 22 Large Aff. at 10-11; Hirsch Report at 4-6, 8-10.15 See SCE's Answering Brief, Att. 7, Docket Nos. 50-361 and 50-362 Steam Generator Return-to-Service Action Plan [SONGS] (Mar. 23, 2012) [hereinafter Mar. 23, 2012 Return-to-Service Plan].16 See SCE's Answering Brief, Att. 3, [CAL] -- [SONGS], Units 2 and 3, Commitments to Address Steam Generator Tube Degradation (Mar. 27, 2012) [hereinafter CAL].17 SCE Answering Brief, Att. 13, NRC Enforcement Policy (June 7, 2012) at 68 [hereinafter NRC Enforcement Policy]. The NRC Enforcement Manual describes a CAL as follows:[Confirmatory Action Letters (CALs)] are flexible and valuable tools available to the staff to resolve licensee issues in a timely and efficient manner, emg., when an order is warranted to address a specific issue, a CAL is a suitable instrument to confirm initial, agreed upon, short-term actions covering the interval period prior 7 the instant case, the March 27, 2012 CAL provides, inter alia, that (1) SCE will take specified investigatory and corrective actions and provide information to the NRC Staff as prescribed in the CAL; and (2) SCE may not restart Units 2 and 3 until the NRC Staff has completed its review of SCE's Restart Reports and has authorized such restarts.

See CAL at 2.B. Procedural

Background

On June 18, 2012, Petitioner submitted a hearing request to the Commission arising out of the Staffs issuance of the CAL.1 8 Petitioner (1) requested that the Commission recognize that the CAL process for the start up of Units 2 and 3 is a de facto license amendment proceeding requiring an adjudicatory hearing (see Petition to Intervene at 2), and (2) proffered the following contention: "Petitioner contends that [SONGS] cannot be allowed to restart without a license amendment and attendant adjudicatory public hearing as required by 10 C.F.R. § 2.309, in which Petitioner and other members of the public may participate." Id. at 16.19 On July 13, 2012, SCE and the NRC Staff filed answers opposing Petitioner's hearing request.2 0 Petitioner filed a reply to those answers on July 20, 2012.21 to the actual issuance of the order.SCE's Answering Brief, Att. 14, [NRC] Enforcement Manual (rev. 7, Oct. 1, 2010) at 3-30 [hereinafter NRC Enforcement Manual].18 See Petition to Intervene and Request for Hearing by Friends of the Earth (June 18, 2012) [hereinafter Petition to Intervene].

19 Petitioner also advanced two other claims in its hearing request that are not relevant to this proceeding.

See infra note 24. In the meantime, on June 27, 2012, the National Resources Defense Council (NRDC) filed a response in support of Petitioner's hearing request. See NRDC's Response in Support of FOE Petition to Intervene, San Onofre Units 2 and 3 (June 27, 2012).20 See [SCE's] Answer Opposing Friends of the Earth Hearing Request and the [NRDC]Response Regarding

[SONGS] Unit 2 and 3 (July 13, 2012); NRC Staffs Answer to Petition to Intervene and Request for Hearing by Friends of the Earth on the Restart of the San Onofre Reactors (July 13, 2012).21 See Reply to SCE's and NRC Staffs Answer to Petition to Intervene and Request for 8 Meanwhile, consistent with its commitment in the CAL, on October 3, 2012, SCE submitted a CAL response to the NRC Staff entitled "Unit 2 Return to Service Report."'2 2 In that Report, SCE represented that it had taken the following corrective actions for Unit 2 and would impose the following operational limits to prevent loss of tube integrity in the steam generators due to tube-to-tube wear: SCE will administratively limit Unit 2 to 70% reactor power prior to a mid-cycle inspection outage....

This administrative limit is temporary and may change based upon the results of inspections, further analysis and long-term corrective actions.SCE has plugged the tubes adjacent to the retainer bars, plugged the two tubes with [tube-to-tube wear] in Unit 2, plugged the tubes with wear that exceeds the 35% through-wall criterion in SONGS Technical Specifications, and preventively plugged additional tubes in Unit 2 based on wear characteristics in Unit 3 tubes and actual wear patterns in Unit 2 (those tubes are in approximately the same region that experienced

[fluid elastic instability]

in Unit 3 at 100% power) .... [A]bout 3% of the total number of tubes in each of the [steam generators]

in Unit 2 have been plugged.SCE will shut down for a mid-cycle steam generator tube inspection outage within 150 cumulative days of operation at or above 15% power.SCE's Answering Brief at 10-11.23 On November 8, 2012, the Commission issued a decision on Petitioner's hearing request. As relevant here, the Commission referred to the ASLBP that portion of the request in which Petitioner argued that "the [CAL] issued to SCE, including the process for resolving the issues raised in the [CAL], constitutes a de facto license amendment proceeding." CLI-12-20, 76 NRC at _ (slip op. at 4). The Commission thus directed a duly constituted Licensing Board Hearing by Friends of the Earth (July 20, 2012).22 See SCE's Answering Brief, Att. 4, Docket No. 50-361, [CAL] -Actions to Address Steam Generator Tube Degradation

[SONGS], Unit 2 (Oct. 3, 2012) [hereinafter SCE's Unit 2 Restart Plan].23 SCE has not yet submitted a Unit 3 Return to Service Report (see SCE's Answering Brief at 11), and it represents that "its CAL response and restart actions for Unit 3 ... may be quite different than those for Unit 2 because the [tube-to-tube wear] in Unit 3 is far more extensive and severe than in Unit 2." Id. at 21.

9 to "consider whether: (1) the [CAL] issued to SCE constitutes a de facto license amendment that would be subject to a hearing opportunity under [s]ection 189a [of the Atomic Energy Act];and, if so, (2) whether the petition meets the standing and contention admissibility requirements of 10 C.F.R. § 2.309." Id. at 5.24 Following its establishment on November 19, 2012,25 this Licensing Board held a conference call on December 3, 2012 to discuss the procedural path forward, including a briefing schedule.2 6 Petitioner filed its opening brief with attachments on January 11, 2013 (see Petitioner's Opening Brief); SCE and the NRC Staff each filed an answering brief with attachments on January 30, 2013 (see SCE's Answering Brief; NRC Staffs Answering Brief);and Petitioner filed its reply brief on February 13, 2013. See Petitioner's Reply Brief.2 7 On March 22, 2013, this Board held an oral argument in the ASLBP's Rockville Hearing Room on the issues referred by the Commission.

2 8 24 As mentioned supra note 19, in its hearing request, Petitioner also advanced two additional claims, asserting that (1) SCE violated 10 C.F.R. § 50.59 insofar as it replaced the steam generators in Units 2 and 3 without seeking a license amendment; and (2) the Commission should exercise its inherent supervisory authority to initiate a discretionary adjudicatory hearing. See Petition to Intervene at 2. The Commission (1) referred Petitioner's section 50.59 claim to the NRC Executive Director for Operations for consideration as a petition under 10 C.F.R. § 2.206 (see CLI-12-20, 76 NRC at _ (slip op. at 4)); and (2) denied, without prejudice, Petitioner's request that the Commission initiate a discretionary adjudicatory hearing.See id. at 5.25 See Southern Cal. Edison Co., Establishment of Atomic Safety and Licensing Board, 77 Fed. Reg. 70,487 (Nov. 26, 2012).26 See Licensing Board Order (Scheduling Conference Call) (Nov. 26, 2012) (unpublished).

This Board's subsequent procedural directives are contained in the following orders: Licensing Board Order (Conference Call Summary and Directive Relating to Briefing) (Dec. 7, 2012)(unpublished);

Licensing Board Order (Granting in Part and Denying in Part Petitioner's Motion for Clarification and Extension) (Dec. 20, 2012) (unpublished).

27 Additionally, NRDC filed an amicus brief in support of Petitioner (see [NRDC's] Amicus Response in Support of Friends of the Earth (Jan. 18, 2013)), and Nuclear Energy Institute (NEI) filed an amicus brief in support of SCE and the NRC Staff. See Amicus Curiae Brief of[NEI] in Response to the NRC [ASLBP's]

Briefing Order (Jan. 30, 2013).28 See Official Transcript of Proceedings (Mar. 22, 2013) [hereinafter Tr.]. The oral argument was web streamed for the benefit of individuals who were unable to attend. See 10 II. ANALYSIS In Part II.A, we define the scope of the de facto license amendment issue referred by the Commission, concluding that -- based on the nature of the CAL process and the language in CLI-12-20

-- the Commission tasked us with determining whether any aspect of this CAL process, including a close-out of the CAL for Unit 2 that results in a plant start-up pursuant to SCE's Unit 2 Return to Service Plan, would constitute a de facto license amendment proceeding.

2 9 In Part II.B, we discuss the legal standards that will guide us in resolving this issue. In Part II.C, we apply the governing legal standards to the facts of this case, and we conclude that this CAL process constitutes a de facto license amendment proceeding that triggers the hearing requirements in section 189a of the AEA. Finally, in Part II.D, we consider the second issue referred by the Commission

-- i.e., whether Petitioner has standing and has submitted an admissible contention.

We conclude that, because our resolution of the first issue Licensing Board Order (Format for Oral Argument) (Mar. 12, 2013) at 2 (unpublished).

During oral argument, SCE announced that it was "considering filing a voluntary license amendment request with a no significan[t]

hazards consideration as the most expeditious method to resolve the issue raised by [Request for Additional Information]

32." See Tr. at 10.Subsequently, on April 8 and 9, 2013, respectively, SCE filed (1) a License Amendment Request for Unit 2; and (2) Supplement 1 to the License Amendment Request for Unit 2. See Docket No. 50-361, Amendment Application Number 263, Steam Generator Program,[SONGS], Unit 2 (Apr. 8, 2013); Docket No. 50-361, Supplement 1 to Amendment Application Number 263, Steam Generator Program, [SONGS], Unit 2 (Apr. 9, 2013). On April 11, 2013, the NRC Staff filed a copy of a "Notice of Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing," which it had forwarded the previous day to the Office of the Federal Register for publication.

See [SONGS], Unit 2 -Notice of Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination, and Opportunity for Hearing (TAC No. MF1 379) (Apr. 11, 2013). No party has filed a motion suggesting that this new development materially affects this proceeding, nor do we discern such an effect, because SCE's license amendment request for Unit 2 does not fully resolve the referred issue for Unit 2 (see infra note 48), much less for Unit 3.29 In this decision, we focus principally on Unit 2, because SCE has not yet submitted a"Unit 3 Return to Service Report." However, because SCE concedes that the tube-to-tube wear in Unit 3 is "far more extensive and severe" than in Unit 2 (see SCE's Answering Brief at 21), our conclusion on the first referred issue (infra Part II.C) would perforce apply to Unit 3 if SCE sought to restart it without a license amendment.

11 grants Petitioner all the relief that its contention seeks, the second issue referred by the Commission is now moot.A. The Scope of the De Facto License Amendment Issue Referred to this Board SCE and the Petitioner disagree sharply about the scope of the first issue referred to this Board. The Commission "direct[ed]

the Board to consider whether...

the [CAL] issued to SCE constitutes a de facto license amendment that would be subject to a hearing opportunity under section 189a [of the AEA]." 3° CLI-12-20, 76 NRC at _ (slip op. at 5). SCE argues that, consistent with the above language, this Board should cabin its review to "the provisions in the[March 27, 2012 letter] itself, without recourse to SCE's CAL response or its restart actions." SCE's Answering Brief at 20. The NRC Staff agrees with SCE's narrow view of the issue. See NRC Staff s Answering Brief at 48-49.On the other hand, Petitioner argues that the Commission referred a broader issue to this Board. Petitioner claims that the Commission viewed the CAL as a process, not as a discrete letter, and it therefore directed this Board to resolve whether any aspect of the CAL process, including a close-out of the CAL that results in a plant start-up pursuant to SCE's Unit 2 Return to Service Plan, would constitute a de facto license amendment proceeding.

See Petitioner's Opening Brief at 6. This conclusion, argues Petitioner, is compelled by (1) the nature of the CAL process; (2) the plain language in CLI-12-20; and (3) common sense. See Petitioner's Opening Brief at 9-10; Tr. at 23-24. We agree with Petitioner.

1. The Nature of the CAL Process Supports Petitioner's Interpretation Regarding the Scope of the Referred Issue SCE and the NRC Staff argue that the first issue requires us to limit our review to the four corners of the March 27, 2012 confirmatory action letter and determine whether that letter, viewed in isolation, constitutes a de facto license amendment.

This argument ignores that, 30 ll.B.1.The hearing opportunity mandated by section 189a of the AEA is discussed infra Part 12 although a "confirmatory action letter" can be referred to as a "CAL," the NRC Enforcement Manual also considers the term "CAL" to be a "process." See NRC Enforcement Manual at 3-32.As described in the NRC Enforcement Manual and as explained by the NRC Staff, the CAL process involves (1) the identification of a significant concern regarding health and safety, safeguards, or the environment; (2) the NRC Staffs issuance of a specific CAL; (3) a licensee responding by taking action and/or providing information as prescribed in the CAL; and (4) when the circumstances that prompted the NRC to issue the CAL have been addressed, the closing out of the CAL.3" See NRC Staffs Answering Brief at 31; NRC Enforcement Manual at 3-29 to 3-36; see also NRC Enforcement Policy at 68.In the instant case, the NRC Staffs use of the CAL process serves, inter alia, to confirm SCE's "[v]oluntary

... suspension of licensed activities" and its "agreement to NRC approval prior to resumption of licensed activities." NRC Enforcement Manual at 3-30. The March 27, 2012 letter thus states that the CAL will remain in effect until the NRC Staff (1) completes its review of SCE's tests, assessments and evaluations, corrective actions, and proposed protocol 31 The Enforcement Manual describes the process for closing out a CAL as follows: 3.5.7 Closing Out CALs A. A CAL may or may not require follow-up inspection to verify completion of the specified licensee actions. Whether the staff believes that an inspection is necessary to close a CAL will be determined on a case-by-case basis and will depend on the circumstances of the case.B. The issuing office (i.e., region, NRR, NMSS, FSME, NRO or NSIR) will issue documentation formally closing out the CAL.C. Correspondence closing out a CAL should be sent to the same person/address as the CAL; however, verbal notification, in advance of written correspondence, may be sufficient to permit plant restart or resumption of affected licensee activities.

NRC Enforcement Manual at 3-35 to 3-36.

13 of inspections and/or operational limits; and (2) concludes that the SONGS Units 2 and 3 can be operated without undue risk to public health and safety, and the environment.

See CAL at 2, 3.On October 3, 2012, SCE informed the NRC Staff that it had completed the actions prescribed in the March 27, 2012 letter for the restart of Unit 2, and it provided detailed information regarding fulfillment of those actions in a document entitled "Unit 2 Return to Service Report." See Unit 2 Return to Service Report.The NRC Staff has not yet closed out the CAL for Unit 2, because it continues to review SCE's "Unit 2 Return to Service Report." Incident to that review, to date, the NRC Staff has issued over 70 Requests for Additional Information (RAIs) to SCE, while SCE has submitted 8 voluminous responses.

3 2 In short, the CAL process for Units 2 and 3 is a protracted and evolving process. It will culminate in a close-out that will permit plant restart if the NRC Staff concludes such action can be accomplished without undue risk to public health and safety, and the environment.

This Board cannot determine whether that process constitutes a de facto license amendment proceeding by looking solely at the March 27, 2012 document that set this lengthy and complex process in motion. Rather, our resolution of that issue must be informed by considering the entire process and the documents generated incident to that process.We recognize that Licensing Boards are not empowered "to supervise or direct NRC Staff regulatory reviews." Duke Energy Corp. (Catawba Nuclear Station, Units 1 and 2), CLI-04-6, 59 NRC 62, 74 (2004). Our resolution of the referred issue will not violate that rule.We do not presume to supervise or to direct the NRC Staff in the performance of its CAL duties, including its review of the adequacy and safety of SCE's restart plan; rather, the scope of our authority is limited to adjudicating the issue referred by the Commission

-- i.e., whether this CAL process constitutes a de facto license amendment proceeding.

32 The NRC Staff issued RAIs to SCE on December 26, 2012 (RAIs 1-32), March 18, 2013 (RAIs 33-67), and March 15, 2013 (RAIs 68-72). See SCE's Eighth Notification of Responses to RAIs (Apr. 23, 2013).

14 The NRC Staff nevertheless argues that the CAL process "does not involve issuing [a license] amendment.

Instead, closing out a CAL would 'permit plant restart or resumption of affected licensee activities."'

NRC Staff's Answering Brief at 32 (quoting NRC Enforcement Manual at 3-36). "If the licensee or Staff determined a license amendment was required," argues the NRC Staff, "that would be done separately from the CAL close-out process." NRC Staffs Answering Brief at 32 n. 157.The short answer to this argument is that "it is the substance of the NRC action that determines entitlement to a section 189a hearing, not the particular label the NRC chooses to assign to its action." Citizens Awareness Network, Inc. v. NRC, 59 F.3d 284, 295 (1st Cir.1995). Consistent with the Commission's directive in CLI-12-20, it is this Board's responsibility to scrutinize the substance of this CAL process to determine whether it constitutes a de facto license amendment proceeding.

To resolve that issue, our inquiry must extend to determining whether the Unit 2 Return to Service Report, in which SCE seeks a CAL close-out that "permit[s a] plant restart" (NRC Enforcement Manual at 3-36), constitutes a de facto license amendment proceeding that triggers a hearing opportunity under section 189a of the AEA.2.. The Language in the Commission's Referral Order Supports Petitioner's Interpretation Regarding the Scope of the Referred Issue The above conclusion is compelled by the plain language in the Commission's referral order. The Commission explicitly stated that Petitioner "contend[ed]

that the [CAL] issued to SCE, including the process for resolving the issues raised in the [CAL], constitutes a de facto license amendment proceeding" (CLI-12-20, 76 NRC at __.) (slip op. at 4)), and it was "this portion of the petition" that the Commission referred to the ASLBP for resolution.

Id. at 4-5.Insofar as the Commission referred a de facto license amendment claim that "includ[ed a challenge to] the process for resolving the issues raised in the [CALJ' (CLI-12-20, 76 NRC at _(slip op. at 4) (emphasis added)), we conclude that the referred issue requires us to determine whether this process, in which SCE seeks a CAL close-out resulting in a plant restart, 15 constitutes a de facto license amendment proceeding.

It is true that there can be no actual license amendment until (and unless) it is issued by the NRC Staff. See 10 C.F.R. § 50.92. It might therefore be argued that this Board should refrain from resolving the de facto license amendment issue until the Staff completes the CAL process by, for example, authorizing the start up of Units 2 and 3.This we decline to do for three reasons. First and foremost, we see no indication in CLI-12-20 that the Commission intended this Board to stay its hand until the Staff has taken final action in the CAL process. Second, if the hearing provision in section 189a of the AEA is to serve its intended purpose, the parties in interest should be afforded a meaningful opportunity to request a hearing before the NRC Staff takes final action that could result in authorizing SCE to operate in a manner that is beyond the ambit of its existing license. Cf. Citizens Awareness Network, Inc., 59 F.3d at 294-95 ("[I]f section 189a is to serve its intended purpose, surely it contemplates that parties in interest be afforded a meaningful opportunity to request a hearing before the Commission retroactively reinvents the terms of an extant license by voiding its implicit limitations on the licensee's conduct.").

Third, all the parties urge this Board to resolve the referred issue without awaiting final Staff action. See Tr. at 59 (SCE), 27 (Petitioner), 112 (NRC Staff). To do otherwise could result in years of delay. See Tr. at 59 (SCE advises that, in its estimation, the CAL close-out for Unit 3 is "not imminent" and is not likely to occur for several years).3. Common Sense Supports Petitioner's Interpretation Regarding the Scope of the Referred Issue Common sense also supports the conclusion that the Commission did not intend this Board to limit its review to the four corners of the March 27, 2012 confirmatory action letter.Otherwise, it would have resolved the issue itself, concluding

-- without difficulty

-- that this austere four-page document, viewed in isolation at the incipient stage of the CAL process, does not constitute a de facto license amendment.

16 However, by referring the issue to the ASLBP, and by acknowledging that Petitioner's claim "include[ed]

the process for resolving the issues raised in the [CAL]" (CLI-1 2-20, 76 NRC at _ (slip op. at 4) (emphasis added)), it may fairly be concluded that the Commission intended a Licensing Board to examine the entire CAL process, and to determine whether any aspect of that process -- including a close-out of the CAL that results in a plant start up pursuant to SCE's Unit 2 Return to Service Plan -- constitutes a de facto license amendment proceeding.

3 3 SCE advances a policy reason in support of its argument that this Board should focus exclusively on the March 27, 2012 CAL and conclude that it is not a de facto license amendment.

Namely, to do otherwise may discourage licensees in the future from agreeing to a CAL, thus (1) diminishing the NRC Staffs use of this important regulatory tool in the future;and (2) undermining the Staffs discretion to select the enforcement action that best fits the factual circumstances.

See SCE Brief at 20-23.This argument lacks merit. First, whether a CAL process constitutes a de facto license amendment proceeding is a highly fact-specific question, and there is no reason to believe that this Board's resolution of this fact-specific issue in this exceptionally unusual case will influence other licensees when they are considering whether to agree to a CAL. Second, "unreviewed Board rulings do not constitute precedent or binding law" (Baltimore Gas & Elec. Co. (Calvert Cliffs Nuclear Power Plant, Units 1 and 2), CLI-98-25, 48 NRC 325, 343 n.3 (1998)), which fortifies our conclusion that our resolution of the referred issue in this unique case will not impact the decision-making process of other licensees when they are considering whether to agree to a CAL. Finally, and dispositively, SCE's policy argument cannot trump the Commission's directive in CLI-1 2-20 that a Licensing Board examine this CAL process and determine whether it 33 We thus agree with the NRC Staffs assertion (see NRC Staffs Answering Brief at 35)that if we were to limit our review to the March 27, 2012 letter, we would conclude that this document, viewed in isolation, is not a de facto license amendment.

In our judgment, however, the Commission eschewed such a facile analytic approach by referring Petitioner's claim to the ASLBP, "including the process for resolving the issues raised in the CAL." CLI-1 2-20, 76 NRC at _ (slip op. at 4).

17 constitutes a de facto license amendment proceeding.

B. Legal Standards That Address License Amendments

1. Relevant Statutory Provisions Related to License Amendments It is imperative that the terms of a reactor operating license be clear and unambiguous, and also that a licensee scrupulously adhere to those terms, because section 101 of the AEA makes it "unlawful

... for any person within the United States to ... use ... any utilization...

facility except under and in accordance with a license issued by the Commission." 42 U.S.C.§ 2131.34 Section 182a of the AEA addresses what must be included in a reactor operating license. It states that such licenses must include "technical specifications" that include, inter alia, "the specific characteristics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization.., of special nuclear material ... will provide adequate protection to the health and safety of the public." 42 U.S.C. § 2232(a).3 5 The Commission is empowered to issue an order amending any license as it deems necessary to "effectuate the provisions of [the AEA]" (42 U.S.C. § 2233) -- that is, to "promote the common defense and security or to protect health or to minimize danger to life or property." Id. § 2201; see also id. § 2237. Additionally, the Commission "may at any time ... before the expiration of the license, require further written statements

[from the licensee]

to determine whether...

a license should be modified." Id. § 2232(a).Finally, section 189a of the AEA states that "[i]n any proceeding under [the AEA], for the 34 A "utilization facility" includes a commercial nuclear power reactor. See 10 C.F.R.§ 50.2.35 "The AEA, however, leaves it up to the Commission to determine, and prescribe by rule or regulation, what additional information should be included in technical specifications to ensure public health and safety and the common defense and security." Dominion Nuclear Connecticut, Inc. (Millstone Nuclear Power Station, Units 2 and 3), CLI-01-24, 54 NRC 349, 351 (2001).

18... amending of any license .... the Commission shall grant a hearing upon the request of any person whose interest may be affected by the proceeding, and shall admit any such person as a party to such proceeding." 42 U.S.C. § 2239(a)(1)(A).

2. Relevant Regulatory Provisions Related to License Amendments 10 C.F.R. §§ 50.90 to 50.92 provide the applicable process when a licensee wishes to request a license amendment.

Specifically, section 50.90 authorizes applications to amend existing operating licenses; section 50.91 provides for notice and comment regarding license amendment applications, as well as consultation with the State in which the facility is located;and section 50.92 provides the standard considered by the NRC when determining whether to issue an amendment.

Section 50.59 establishes standards for a licensee to request a license amendment before it may make "changes in the facility as described in the [updated]

final safety analysis report [UFSAR 3 6], make changes in the procedures as described in the [UFSAR], and conduct tests or experiments not described in the [UFSAR]." 10 C.F.R. § 50.59(c)(1).

Section 50.59 states that a licensee need not request a license amendment pursuant to section 50.90 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." Id. § 50.59(c)(1)(i)-(ii).

Restated, a licensee must request a license amendment if the proposed action requires that existing technical specifications be changed (see 10 C.F.R. § 50.59(c)(1)(i)), 3 7 or if a 36 A final safety analysis report (FSAR) is part of the application for an operating license, and it contains "a description of the facility; the design bases and limits on operation; and the safety analysis for the structures, systems, and components (SSC) and of the facility as a whole." Changes, Tests, and Experiments:

Proposed Rule, 63 Fed. Reg. 56,098, 56,099 (Oct.21, 1998). "When a plant is licensed, the NRC states in its Safety Evaluation Report (SER) why it found each FSAR analysis acceptable." Id. Licensees must periodically update their FSARs to reflect changes to the facility "so that the [updated FSAR (UFSAR)] remains a complete and accurate description and analysis of the facility." Id.37 Because changes to technical specifications require a license amendment, the 19 change, test, or experiment satisfies any of the eight criteria in section 50.59(c)(2).

See id.§ 50.59(c)(1)(ii).

The section 50.59(c)(2) criteria require a licensee to seek a license amendment if the proposed change, test, or experiment would (i) Result in more than a minimal increase in the frequency of occurrence of any accident previously evaluated in the [UFSAR];(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 [UFSAR];(iii) Result in more than a minimal increase in the consequences of an accident previously evaluated in the [UFSAR];(iv) Result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the [UFSAR];(v) Create a possibility for an accident of a different type than any previously evaluated in the [UFSAR];(vi) Create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the [UFSAR];(vii) Result in a design basis limit for a fission product barrier as described in the[UFSAR] being exceeded or altered; or (viii) Result in a departure from a method of evaluation described in the [UFSAR]used in establishing the design bases or in the safety analyses.Id. § 50.59(c)(2).

3 8 Commission has instructed that technical specifications should be limited to "'those plant conditions most important to safety."'

Millstone, CLI-01-24, 54 NRC at 360 (quoting Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, 58 Fed. Reg.39,132, 39,135 (July 22, 1993)). Thus, technical specifications "should be reserved for those reactor operation

'conditions or limitations

... necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health or safety."'

Id.at 361 (quoting Technical Specifications, Final Rule, 60 Fed. Reg. 36,953, 36,957 (July 19, 1995)). See also 10 C.F.R. § 50.36 (identifying criteria to be used in determining what items must be included in technical specifications).

38 The term "design bases" to which section 50.59(c)(2)(vii) and (viii) refer is defined as follows: Design bases means that information which identifies the specific functions to be performed by a structure, system, or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for a design. These values may be (1) restraints derived from generally accepted "state of the art" practices for achieving functional goals, or 20* Finally, 10 C.F.R. § 2.105 implements the hearing opportunity provision for license amendment procedures that is mandated by section 189a of the AEA, and Subpart C of 10 C.F.R. Part 2 contains the general rules governing hearing requests and subsequent hearing-related activities.

In sum, Congress has commanded that licensees may not, under penalty of law, deviate from the terms of their reactor operating licenses.

See 42 U.S.C. § 2131. If a licensee is unable to operate a reactor in strict accordance with its license, it must seek authorization from the NRC for a license amendment (10 C.F.R. §§ 50.59, 50.90 to 50.92), which is a process that triggers a right to request an adjudicatory hearing by persons whose interests may be affected by the proceeding.

See 42 U.S.C. § 2239(a)(1)(A);

10 C.F.R. § 2.105.3. De Facto License Amendments As shown above, amending a license is, by design, a carefully considered process that is closely regulated by the NRC and in which "any person whose interest may be affected" is entitled to request a hearing. 42 U.S.C. § 2239(a)(1)(A).

As discussed below, however, there have been occasions when the NRC has taken action that effectively constituted a license amendment, but it failed to recognize that its actions effectively amended the license.In other words, there have been occasions when the NRC has -- without formally amending a license and without providing the public with the opportunity for a hearing as required by section 189a of the AEA -- authorized activity by the licensee that was incompatible with the statutory requirement that the facility operate "in accordance with" its existing operating license. 42 U.S.C. § 2131. Such NRC action is characterized as a de facto license amendment.

According to Petitioner, this CAL process is a de facto license amendment (2) requirements derived from analysis (based on calculation and/or experiments) of the effects of a postulated accident for which a structure, system, or component must meet its functional goals.10 C.F.R. § 50.2.

21 proceeding because SCE seeks effectively to amend its license via the CAL process.Specifically, Petitioner argued to the Commission that "the [CAL] issued to SCE, including the process for resolving the issues raised in the [CAL], constitutes a de facto license amendment proceeding within the hearing provision of section 189a of the AEA, and therefore an adjudicatory hearing is required." CLI-12-20, 76 NRC at _ (slip. op at 4). The Commission referred that claim to the ASLBP for resolution.

Id.Determining whether the CAL process constitutes a de facto license amendment proceeding "is a highly fact-specific question." NRC Staffs Answering Brief at 10. Case law, however, provides a straight-forward analytic framework for assessing the relevant facts. For example, in Cleveland Elec. Ilium. Co. (Perry Nuclear Power Plant), CLI-96-13, 44 NRC 315 (1996), the Commission considered whether the NRC Staffs decision to authorize changes to a material specimen withdrawal schedule was a de facto license amendment.

Examining decisions from the U.S. Courts of Appeals for the First Circuit and the District of Columbia Circuit, the Commission distilled the following factors that are material to determining whether NRC actions constitute a de facto license amendment:

In evaluating whether challenged NRC authorizations effected license amendments within the meaning of section 189a, courts repeatedly have considered the same key factors: did the challenged approval grant the licensee any "greater operating authority," or otherwise "alter the original terms of a license"?

If so, hearing rights likely were implicated.

For example, in Citizens Awareness Network, Inc. v. NRC, 59 F.3d 284, 295 (1st Cir. 1995) (CAN), ...the court found that the challenged NRC approval "undeniably supplement[edj the original license. The agency had permitted the licensee to dismantle major structural components, an activity that the court found unauthorized by the original license and agency rules. Similarly, in another case [San Luis Obispo Mothers for Peace v. NRC, 751 F.2d 1287 (D.C. Cir. 1984) (SLO)], where the NRC Staff extended the duration of a low-power license, a reviewing court viewed the Staff approval to be a license amendment changing a term of the license, and therefore triggering an opportunity for a hearing under section 189a.44 NRC at 326-27 (footnotes omitted).

Guided by CAN and SLO, the Commission in Perry considered whether the Staffs action (1) "alter[ed]

the.., license," or (2) "permit[ted]

the licensee to operate 'in any greater capacity' than [the original license prescribes]." Id. After 22 examining the relevant terms and technical specifications in the license, the Commission resolved both inquiries in the negative.3 9 As illustrated in the Perry case, a de facto license amendment claim typically involves a tribunal "looking backward" to determine whether action already taken by the NRC Staff effectively constituted a license amendment.

Here, however, consistent with the Commission's referral order, we are tasked with looking at an ongoing CAL process to determine whether that process constitutes a de facto license amendment proceeding.

See supra Part II.A. To resolve that issue, this Board must determine whether the requested change in authority to operate Unit 2 sought by SCE pursuant to the CAL process is strictly "in accordance with" the terms and technical specifications in its existing license. 42 U.S.C. § 2131.In other words, this Board must consider the following connate factors: whether SCE's start-up request, if granted, would permit SCE to operate (1) in a manner that deviates from a technical specification in its existing license; (2) beyond the ambit, or outside the restrictions, of its existing license; or (3) in a manner that is neither delineated nor reasonably encompassed 39 For additional pronouncements on standards employed by tribunals in the context of considering de facto license amendment issues, see, eq., Perry, CLI-96-13, 44 NRC at 319 ("Because technical specifications are an integral part of an operating license, changes to technical specifications require a license amendment.");

id. at 320 (the UFSAR "can be modified without a license amendment, so long as the modifications do not involve a change to the technical specifications or an unreviewed safety question");

CAN, 59 F.3d at 294 ("[B]y its nature a license is presumptively an exclusive

-- not an inclusive

-- regulatory device. ...Regulated conduct which is neither delineated, nor reasonably encompassed within delineated categories of authorized conduct, presumptively remains unlicensed.");

id. at 295 (NRC's actions constitute de facto license amendment when they authorize licensee to "engage in[activities]

beyond the ambit of [its] original license");

Massachusetts

v. NRC, 878 F.2d 1516, 1520-21 (1st Cir. 1989) (NRC's actions in requiring 47 improvements, granting an exemption from emergency drills, and lifting a license suspension did not require a license amendment, because the licensee can "operate[]

in accordance with its unaltered license" and need not be"exempted

... from following a specific license requirement");

In re Three Mile Island Alert, Inc., 771 F.2d 720, 729 (3d Cir. 1985) (NRC's lifting of license suspension and authorizing restart under stipulated restrictions was not a license amendment because "nothing in this record ...indicates

... that license amendments are necessary to permit the licensee to operate in accordance with the restrictions which have been imposed"), cert. denied, 475 U.S. 1082 (1986).

23 within the prescriptive terms of its existing license. See supra note 39 and accompanying text.4 0 In assessing the referred issue, this Board can refer to 10 C.F.R. § 50.59, which -- as discussed supra Part ll.B.2 -- identifies situations where a licensee must request a license amendment.

In our view, reference to the criteria in section 50.59 is eminently appropriate here, because the ultimate question before this Board is whether SCE's request that the Staff close out the CAL by permitting a plant restart constitutes a de facto license amendment proceeding that triggers a hearing opportunity under section 189a of the AEA. To resolve this question, we must look at SCE's Unit 2 Return to Service Plan to determine whether SCE is seeking authority from the NRC Staff to deviate from a technical specification or to otherwise operate in a manner that is beyond the ambit, or inconsistent with the prescriptive terms, of its existing license.Section 50.59 establishes standards that may guide this Board in resolving that issue.Contrary to arguments advanced by the NRC Staff (see NRC Staff Answer at 43-47; Tr.at 140), the fact that section 50.59 is designed for a licensee to determine whether it must seek a license amendment ab initio poses no impediment to this Board referring to those same regulatory standards as guides in determining whether this CAL process constitutes a de facto license amendment proceeding.

The standards in section 50.59 -- which establish when a"licensee shall obtain a license amendment" (10 C.F.R. § 50.59(c)(2))

-- have the imprimatur of the Commission and therefore, a fortiori, are appropriate guides for determining whether SCE's Unit 2 Return to Service Plan requires a license amendment, thereby converting the CAL process into a de facto license amendment proceeding.

Our use of section 50.59 as a tool in resolving the referred issue is to be distinguished from scrutinizing the actual actions taken by SCE under section 50.59. The latter is prohibited 40 At the March 22, 2013 oral argument, counsel for the NRC Staff was asked whether the need for a license amendment is limited to circumstances that involve an increase in licensing authority, or whether a license amendment would also be required where, for example, the Staff were to change the licensing authority by decreasing the maximum operating thermal power for a nuclear reactor. Counsel responded that a license amendment would be required for both situations.

See Tr. at 130.

24 by case law, which establishes that "[a] member of the public may challenge an action taken under 10 C.F.R. § 50.59 only by means of a petition under 10 C.F.R. § 2.206." Yankee Atomic Elec. Co. (Yankee Nuclear Power Station), CLI-94-3, 39 NRC 95, 101 n.7 (1994). Contrary to the NRC Staffs assertion (see NRC Staff Answer at 44-49; Tr. at 141), any reference we might make to section 50.59 will not run afoul of this rule, because the issue presented here is not a challenge to SCE's previous actions taken under section 50.59.41 Rather, the Commission directed us to determine whether this CAL process constitutes a de facto license amendment proceeding.

To resolve this issue, it is manifestly appropriate for this Board to consider, and to be guided by, all relevant analytic tools, including

-- if warranted

-- the standards in section 50.59. Cf. Tr. at 31-32, 59-60 (SCE and Petitioner both agree that this Board can properly refer to section 50.59 for purposes of resolving whether this CAL process constitutes a de facto license amendment proceeding).

C. This CAL Process Constitutes a De Facto License Amendment Proceedinq We turn now to the first of the two issues referred by the Commission:

whether this CAL process for the start up of SONGS Unit 2 constitutes a de facto license amendment proceeding.

4 2 As discussed supra Part ll.B.3, to constitute a de facto license amendment proceeding, this CAL process must involve proposed actions by SCE that, if authorized, would allow SCE to deviate from a technical specification or otherwise operate Unit 2 in a manner that is inconsistent with existing licensing requirements or restrictions.

We conclude that this CAL process constitutes a de facto license amendment proceeding for the following three independent reasons: 41 Indeed, it is impossible on the present record -- as a legal and factual matter -- for Petitioner to challenge, or for this Board to review, SCE's section 50.59 analysis for the Unit 2 Return to Service Plan because a copy of SCE's analysis has not even been filed with this Board.42 As stated supra note 29, although our analysis focuses on Unit 2, it would necessarily apply to Unit 3 if SCE sought to restart it without a license amendment.

25 (1) The restart of Unit 2 would grant SCE authority to operate without the ability to comply with all applicable technical specifications; (2) The restart of Unit 2 would allow SCE to operate beyond the scope of its existing license; and (3) SCE's Unit 2 Return to Service Plan includes a test or experiment that meets the criteria in 10 C.F.R. § 50.59 that require a license amendment.

Below, we provide a factual backdrop for our analysis, after which we discuss each of the above reasons in turn.The unprecedented extent of tube wear and failures that SCE experienced in the SONGS Unit 3 replacement steam generators reveal that these steam generators have serious design and operational issues (see SCE's Answering Brief at 10; supra Part I.A), placing them beyond the envelope of experience with U-tube steam generators.

SCE's investigation into the cause of the multiple tube leaks indicates that the design is prone to tube-to-tube wear caused by in-plane fluid elastic instability, which "had not been previously experienced in U-tube steam generators." SCE's Answering Brief at 10.As mentioned supra Part I.A, fluid elastic instability results from the combination of localized high steam velocity, high steam void fraction, and insufficient contact forces between the tubes and the anti-vibration bars. The fluid elastic instability caused vibration of steam generator tubes in the in-plane direction resulting in rapid, localized tube wear. See SCE's Unit 2 Restart Plan at 2; Assessment for Tube-to-Tube Wear at 15."In contrast to the extensive

[tube-to-tube wear] in Unit 3, [tube-to-tube wear in Unit 2]existed in only a single pair of tubes ... in one of the two [replacement steam generators]." SCE's Answering Brief at 9. Although the Unit 2 steam generators did not experience the accelerated and extensive tube-to-tube wear suffered in the Unit 3 steam generators, they nevertheless are the identical design as those in Unit 3 and they operate under similar conditions.

See SCE's Answering Brief, Att. 18, SONGS UFSAR Excerpt at 5.4-20 [hereinafter SONGS UFSAR]; Brabec Aft. at 4-6, 18.

26 SCE claims that the fact that steam generator tube-to-tube wear was significantly less in Unit 2 than in Unit 3 is attributable to the differences in meeting fabrication tolerances.

See SCE's Answering Brief at 10, 92. Fabrication tolerances permit small differences between components designed to the same specifications, and SCE attributes the large difference in steam generator operational performance to very small differences in their construction.

4 3 More precisely, SCE asserts that the difference in steam generator tube wear between Unit 3 and Unit 2 is due in large part to differences in contact between the steam generator tubes and the anti-vibration bars arising from differences in meeting fabrication tolerances.

SCE explains the role played by anti-vibration bars in preventing in-plane vibrations as follows: "The effect of flat bar supports with small clearance is to act as apparent nodal points for flow-induced tube response.

They not only prevent out-of-plane mode as expected but also in-plane modes." Assessment for Tube-to-Tube Wear at 17.But "[w]ear at [anti-vibration bar] locations will degrade in-plane support effectiveness over time." Assessment for Tube-to-Tube Wear at 104. Such degradation can be caused "by a combination of turbulence and out-of-plane fluid-elastic excitation." Id. at 15. As contact is lost between the tube and the bar, the restraining effect of the anti-vibration bars in the in-plane direction decreases.

These decreases, when combined with certain thermal hydraulic conditions, allow in-plane vibration and tube-to-tube wear to develop over time at locations 43 Manufacturing of components is never perfectly exact. Thus, if the nominal design specifies a required distance between adjacent steam generator tubes, it will also specify how closely the manufacturer must come to that required distance.

This permitted variance from the design is referred to as the fabrication tolerance.

See SONGS Unit 2 Return to Service Report, Att. 6 -App. D, Operational Assessment of Wear Indications in the U-bend Region of San Onofre Unit 2 Replacement Steam Generators at 100-02 (ADAMS Accession No.ML12285A269, which is entitled "Attachment 6: Appendix A: Estimate of FEl-Induced TTW Rates" on ADAMS, but also contains Appendix D, starting on page 78 of 209 of the ADAMS portable document format (PDF) version).

Ironically, SCE indicates that the steam generators for Unit 3 were built more closely to design specifications than those in Unit 2, and it maintains that this greater manufacturing precision rendered the Unit 3 steam generators more susceptible to in-plane tube vibration.

See SCE's Answering Brief at 92; accord Unit 2 Return to Service Report at 36.

27 where it previously had not occurred.

See id. at 104; SONGS Unit 2 Return to Service Report, Aft. 6 -App. B, SONGS U2C17 Generator Operational Assessment for Tube-to-Tube Wear at 21 (ADAMS Accession No. ML12285A268).

Moreover, tube-to-tube wear "due to in-plane fluid elastic instability is a unique degradation mechanism because one unstable tube can drive its neighbor into instability through repeated impact events." Assessment for Tube-to-Tube Wear at 18. It is thus possible for in-plane instability to develop in a single tube and propagate to a larger number of tubes in the vicinity.Wear of steam generator tubes is of critical importance to evaluations performed in the FSAR, because the tubes are part of the reactor coolant pressure boundary, and assurance of their integrity is required by General Design Criterion

14. Numerous analyses are grounded on the assumed integrity of steam generator tubes, and technical specifications exist to assure their integrity.

4 5 Any new phenomenon that could negatively impact tube integrity can affect, and possibly negate, assumptions used in FSAR analyses.SCE and its contractors have evaluated the in-plane tube-to-tube wear due to fluid elastic instability and have developed a theory to explain its occurrence and to predict how it can be avoided. As a result of comparing the thermal hydraulic conditions in the SONGS replacement steam generators with those of other steam generators, SCE concluded that the likelihood of fluid elastic instability will decrease if the steam quality in the steam generators is reduced (i.e., if the moisture content of the steam is increased).

See Unit 2 Return to Service Report at 37. SCE determined that a reduced steam quality results in greater "damping" within the steam generators, which decreases the potential for fluid elastic instability.

See id.44 10 C.F.R. Part 50, App. A -General Design Criteria for Nuclear Power Plants, Criterion 14, states: "Reactor Coolant Pressure Boundary.

The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture." 45 See, e.g., SCE's Answering Brief, AUt. 9, SONGS Technical Specification 5.5.2.11, Steam Generator Program [hereinafter SONGS Unit 2 Technical Specifications].

28 SCE provided the following explanation regarding the relation between steam quality and damping, and the effect of damping on fluid elastic instability:

Damping is the result of energy dissipation and delays the onset of [fluid elastic instability].

Damping is greater for a tube surrounded by liquid compared to a tube surrounded by gas. Since quality describes the mass fraction of a vapor in a two-phase mixture, it provides insight into the fluid condition surrounding the tube. A higher steam quality correlates with dryer conditions and provides less damping. Conversely, lower steam quality correlates with wetter conditions resulting in more damping, which decreases the potential for [fluid elastic instability].

Unit 2 Return to Service Report at 38.When compared to steam generators at other plants that do not experience fluid elastic instability, SCE calculated that the steam quality in the SONGS replacement steam generators was higher when operated at 100% power. On the other hand, when SONGS steam generators were operated at 70% power, steam quality was in the same range as those steam generators that did not experience fluid elastic instability.

See Assessment for Tube-to-Tube Wear, Figures 4-3 and 5-1.SCE concluded that limiting the power generated at SONGS Unit 2 to 70% would reduce steam quality and hydrodynamic pressure to values that would eliminate the thermal hydraulic conditions that cause fluid elastic instability and associated tube-to-tube wear in the SONGS Unit 2 steam generators.

See SCE's Unit 2 Restart Plan at 3; Unit 2 Return to Service Report at 37.46 SCE's most recent assessment indicates that, after operating for less than two years (i.e., 20.6 months), tube integrity for the Unit 2 steam generators can be guaranteed only for another eleven months of operation at 100% power. See SCE's Fifth Notification of Responses to RAIs, Enc. 1, Docket No. 50-361, Operational Assessment for 100% Power Case Regarding 46 See also Transcript of Briefing Before Commission on Steam Generator Tube Degradation (Feb. 7, 2013) at 48 (MHI agrees that a reduction to 70% power would improve the thermal hydraulic condition in the steam generators by reducing the steam quality and bringing it into a range seen in other steam generators manufactured by MHI).

29[CAL] Response (TAC No. ME9727) [SONGS], Unit 2 (Mar. 14, 2013) [hereinafter SCE's Fifth Notification of Responses to RAIs].Against the above backdrop, we explain below why we conclude that this CAL process is a de facto license amendment proceeding.

1. Under SCE's Return to Service Plan, Unit 2 Cannot be Operated "Over the Full Range Of Normal Operating Conditions" Up to 100% Power, Which is, Inconsistent with a Technical Specification and Therefore Requires a License Amendment SONGS Unit 2 Technical Specification 5.5.2.1 lb. 1 requires that "[a]ll inservice 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.

4 7 Under its current license, SCE is authorized to operate Unit 2 up to 3,438 megawatts thermal, which is defined as 100% power. See SCE's Answering Brief, Aft. 19, SONGS Operating License 226 at 3.In its Unit 2 Return to Service Report, SCE proposes administratively to limit Unit 2 to 70% reactor power prior to a mid-cycle inspection outage. See SCE's Unit 2 Restart Plan at 3.Based on its analyses, asserts SCE, a 70% power-level limit will provide adequate margin to preclude the onset of in-plane fluid elastic instability and excessive tube wear. See id.If, pursuant to the CAL process, the NRC Staff were to authorize SCE to operate Unit 2 at a power limit not to exceed 70%, this condition would result in a deviation from the technical specification requirement that tube integrity be maintained over the "full range of normal operation conditions" up to 100%. Such a deviation from a technical specification requires a license amendment, thus converting this CAL process to a de facto license amendment proceeding.

4 8 47 See NRC's Answering Brief, Att. 8, Docket No. 50-361, SONGS Unit 2 Facility Operating License No. NPF-10 Excerpts at 5.0-14.48 In SCE's license amendment request for Unit 2 (see supra note 28), SCE seeks the following licensing revisions:

30 2. Unit 2 Cannot Operate Within the Scope of its Operating License,49 Which Requires that the License Must be Amended SONGS Unit 2 is currently licensed to operate anywhere in the normal power range from 0% to 100% power with steam generators that meet the original design specifications.

The original steam generators in SONGS Unit 2 (and Unit 3) were replaced without a license amendment arising from design differences, which SCE claims was in compliance with 10 C.F.R. § 50.59. See Tr. at 79-81. As discussed in greater detail supra Part ll.B.2, section 50.59 permits changes with respect to components (ji.(., steam generators) without a license amendment under prescribed conditions that assure the replacement components are sufficiently similar to the original so that safety requirements are maintained or improved.

See 10 C.F.R. § 50.59(c)(2).

The replacement steam generators in SONGS Unit 3, however, unexpectedly demonstrated significant in-plane vibrations due to fluid elastic instability.

The vibrations were The proposed amendment requests that Technical Specification 5.5.2.11 .b.1 be revised to add a footnote to require that compliance with the steam generator structural integrity performance criterion (SIPC) be demonstrated up to 70%Rated Thermal Power (2406.6 megawatts thermal) and that Facility Operating License Condition 2.C(1) "Maximum Power Level" be revised to add a footnote to restrict operation of SONGS Unit 2 to no more than 70% Rated Thermal Power for the SONGS Unit 2, Cycle 17.Docket No. 50-361, Amendment Application Number 263, Steam Generator Program [SONGS], Unit 2 (Apr. 5, 2013) at 1. Although SCE's license amendment request addresses the first reason underlying our conclusion that this CAL process constitutes a de facto license amendment proceeding, it does not address the alternative reasons underlying our conclusion (see infra Parts II.C.2 and I1.C.3) and it, thus, does not fully address, much less moot, the first issue referred by the Commission.

49 Although the term "scope of an operating license" does not have a regulatory definition, it is a useful concept in the instant context, because the Court of Appeals for the First Circuit has held that actions by the NRC Staff constitute a de facto license amendment when they authorize a licensee to "engage in [activities]

beyond the ambit [i.e., scope] of [its] original license." CAN, 59 F.3d at 295; accord Perry, CLI-96-13, 44 NRC at 327. As described by the Commission, an operating license reflects a specific facility-design basis, a safety analysis documented in an FSAR, facility-specific technical specification, and NRC regulations.

See 63 Fed. Reg. 56,098, 56,099-100.

These factors comprise the scope of an operating license as we use the term in this Memorandum and Order.

31 severe enough to cause tube-to-tube contact resulting in accelerated wear of the tube wall and premature wall failure. See Assessment for Tube-to-Tube Wear at 18. This phenomenon has never before been seen in a U-tube steam generator (see SCE's Answering Brief at 10), which supports a conclusion that the replacement steam generators differ in significant respects from the originals.

Because the Unit 3 steam generators are identical in design to the Unit 2 steam generators (see SONGS UFSAR at 5.4-20; Brabec Aff. at 4-6, 18), we conclude that the latter steam generators likewise differ in significant respects from the originals.

Concerning the FSAR analysis of steam generator tube integrity, SCE states that "[t]he original analysis was fine if we had simply received steam generators that met our specifications" (i.(, were like-for-like replacements), but "[w]hat we had is a degraded or nonconforming condition in our steam generators where they did not perform per the procurement specifications." See Tr. at 98. The extent to which the replacement steam generators failed to perform per the procurement specifications is graphically illustrated by the fact that the original steam generators lasted about twenty-eight years, whereas SCE's most recent operational assessment indicates that, after less than two years of operation (i.e., 20.6 months), tube integrity for Unit 2 steam generators can be guaranteed only for another eleven months of operation at 100% power. See SCE's Fifth Notification of Responses to RAIs.Significantly, the UFSAR for the original steam generators for SONGS Units 2 and 3 excluded the possibility of in-plane vibrations caused by fluid elastic instability when evaluating the conditions necessary to maintain steam generator tube integrity.

In this regard, the UFSAR states: The steam generator was designed to ensure that critical vibration frequencies are well out of the range expected during normal operation and during abnormal conditions.

The tubing and tubing supports are designed and fabricated with considerations given to both secondary side flow-induced vibration and reactor coolant pump-induced vibrations.

32 SONGS UFSAR at 5.4-21 ;50 see also id. at 5.4-23 to 5.4-26 (analysis in section 5.4.2.3.1.3 evaluating conditions necessary to maintain tube integrity in the original steam generators based on the assumption that vibrations caused by in-plane fluid elastic instability will not occur).However, the UFSAR assumption for the original steam generators that in-plane vibrations caused by fluid elastic instability were excluded by design is demonstrably unjustified for the replacement steam generators.

This renders inadequate the UFSAR section 5.4.2.3.1.3 analysis of steam generator tube integrity, which places the replacement steam generators outside the scope of the operating license." 1 We conclude that until the tube degradation mechanism is fully understood, until reasonable assurance of safe operation of the replacement steam generators is demonstrated, and until there has been a rigorous NRC Staff review appropriate for a licensing action, the operation of Unit 2 would be outside the scope of its operating license because the replacement steam generator design must be considered to be inconsistent with the steam generator design specifications assumed in the FSAR and supporting analysis.

In short, the start-up of Unit 2 pursuant to the CAL process would transform that process into a de facto license amendment 50 The reference in the UFSAR to "critical vibration frequencies" and "secondary side flow-induced vibration" subsume the in-plane vibrations caused by fluid elastic instability experienced in the SONGS replacement steam generators.

See generally SONGS Unit 2 Return to Service Report, Att. 6 -App. D, Operational Assessment of Wear Indications in the U-bend Region of San Onofre Unit 2 Replacement Steam Generators at 10-12 (ADAMS Accession No.ML12285A269, which is entitled "Attachment 6: Appendix A: Estimate of FEI-Induced TTW Rates" on ADAMS, but also contains Appendix D, starting on page 78 of 209 of the ADAMS portable document format (PDF) version);

cf. SCE's Answering Brief, Att. 5, MHI Document L5-04GA564 Tube Wear of Unit-3 RSG Technical Evaluation Report at 11 (MHI states that incident to the design of the SONGS replacement steam generators, "only out-of-plane vibration of the[steam generator]

U-tubes was evaluated").

51 The purpose of the UFSAR section 5.4.2.3.1.3 analysis is to verify that General Design Criterion 14 -- which concerns maintaining integrity of the reactor coolant pressure boundary (see supra note 44) -- is satisfied.

We now know that General Design Criterion 14 cannot be satisfied for the steam generator tubes without an analysis of in-plane fluid elastic instability.

33 proceeding by allowing steam generator operation with a tube degradation mechanism not considered in the FSAR -i.e., in-plane vibrations due to fluid elastic instability.

5 2 3. A Unit 2 Start-Up Pursuant to SCE's Return to Service Report Would Result in SCE Conducting a Test or Experiment Pursuant to 10 C.F.R. § 50.59(c)(2)(viii), Which Requires a License Amendment In Part ll.B.3 supra, we determined that we may use the standards in section 50.59 --which establish when a "licensee shall obtain a license amendment" (10 C.F.R. § 50.59(c)(2))

-- as guidance to determine whether implementation of SCE's Unit 2 Return to Service Report requires a license amendment.

As relevant here, section 50.59 requires a licensee to seek a license amendment before implementing a "test or experiment" that will "[r]esult in a departure from a method of evaluation described in the [UFSAR] used in establishing the design basis or in the safety analysis." 10 C.F.R. § 50.59(c)(2)(viii).

Guided by that provision, we conclude that the authority to operate sought by SCE in its Unit 2 Return to Service Report is such a "test or experiment" that requires a license amendment and, thus, transforms this CAL process into a de facto license amendment proceeding.

5 3 SCE's analysis of the cause of the excessive tube wear and the measures it proposes to implement to preclude such wear are based on a theory as applied to U-tube steam generators, 52 The required change to the current FSAR analysis is that it must be augmented with a vibration analysis to assure that steam generator tubes do not fail prematurely due to tube-to-tube wear and that tubes are thus able to satisfy their design bases. As the Commission has explained, a licensee must seek a license amendment "at the point in time [when] the revised method [in the FSAR] becomes the means used for purposes of satisfying FSAR safety analysis or design bases." Changes, Tests, and Experiments:

Final Rule, 64 Fed. Reg. 53,582, 53,598 (Oct. 4, 1999).53 Although Petitioner's briefs rely heavily on 10 C.F.R. § 50.59 in support of its argument that this CAL process constitutes a de facto license amendment proceeding (see, e.., Petitioner's Brief at 19-23), they do not specifically reference section 50.59(c)(2)(viii).

We do not view this omission as a waiver, however, because Petitioner's brief included an argument based on the rationale in section 50.59(c)(2)(viii).

See Petitioner's Brief at 13; Large Affidavit at 5; see also Tr. at 42-44. Indeed, SCE understood Petitioner to be advancing such an argument, as evidenced by the fact that SCE endeavored to rebut it. See SCE's Answering Brief, App. A, Examples of Mischaracterizations in the FOE Brief, Affidavits, and NRDC Brief at 118-19.

34 although that theory is not yet supported by actual experience.

5 4 SCE nevertheless proposes to implement the following sequence of steps incident to the start-up and operation of Unit 2: (1) Unit 2 will be operated at 70% power for a limited duration; (2) this duration will be selected in such a manner that if the calculations are wrong, tube-to-tube wear will likely not progress far enough to cause any tube failures; (3) Unit 2 will then be shut down; and (4) 100% of the steam generator tubes will be inspected, and the inspection results can be compared to current wear data to determine the wear rate and provide confirmation vel non of the theoretical analysis.See SCE's Answering Brief at 10-11.The above steps satisfy the regulatory definition of "tests or experiments not described in the [UFSAR,]" which constitute "any activity where any structure, system, or component is utilized or controlled in a manner which is either: (i) [o]utside the reference bounds of the design bases as described in the [UFSAR] or (ii) [i]nconsistent with the analyses or descriptions in the [UFSAR]." 10 C.F.R. § 50.59(a)(6).

Because the phenomenon of in-plane fluid elastic instability had not previously been observed in U-tube steam generators, and because tube 54 As evidenced by the following, SCE's prediction that accelerated tube wear will be precluded by plant operations limited to 70% power is grounded on theory that is not yet supported by actual experience.

First, SCE's Steam Generator Operational Assessment for Tube-to-Tube Wear by Areva states that "[i]n-plane modes that have never been observed to be unstable even though the computed fluid-elastic stability margins are well below 1." Assessment for Tube-to-Tube Wear at 16. In other words, in-plane vibrations due to fluid elastic instability have not occurred even though the theory predicts in-plane instability.

Second, regarding the tests conducted by Westinghouse, which developed the criteria for in-plane vibrations used for the Unit 2 steam generators, SCE states that the "[in-plane]

instability was never observed in any of [the] square-pitch U-bend tests despite early attempts to force its occurrence without any [anti-vibration bar] support for flows up to three times the [out-of-plane]

instability threshold." SONGS Unit 2 Return to Service Report, Aft. 6 -App. D, Operational Assessment of Wear Indications in the U-bend Region of San Onofre Unit 2 Replacement Steam Generators at 14 (ADAMS Accession No. ML12285A269, which is entitled "Attachment 6: Appendix A: Estimate of FEI-Induced TTW Rates" on ADAMS, but also contains Appendix D, starting on page 78 of 209 of the ADAMS portable document format (PDF) version).Additionally, SCE states that in subsequent tests using triangular arrays, "[a]s was the case for square array patterns, no in-plane instability was observed in these tests even for U-bend tubes with no supports above the top tube support plate." Id. In short, there is a dearth of applicable experiential data available for in-plane vibrational motion, because, as conceded by SCE, "tube-to-tube wear due to in-plane [fluid elastic instability]

ha[s] not been previously experienced in U-tube steam generators." SCE's Answering Brief at 10.

35 failures based on that phenomenon had not been envisioned, the FSAR did not include an analysis or description of it. See supra note 50 and accompanying text. Accordingly, any operation of Unit 2 that might result in in-plane vibrations due to fluid elastic instability, is"[i]nconsistent with the analyses or descriptions in the UFSAR" (10 C.F.R. § 50.59(a)(6)), which, in turn, is the type of "test or experiment" that triggers the obligation under section 50.59(c)(2)(viii) to seek a license amendment.

5 5 According to SCE, even if the sequence of start-up and operational steps in its Unit 2 Return to Service Report are viewed as tests or experiments that result in a "substantial change in an analysis" in the UFSAR, such a change "does not per se require a license amendment under 10 C.F.R. § 50.59." SCE's Answering Brief at 83. For example, "[i]f the analytical method is not described in the UFSAR," states SCE, "a change to that method does not require[a license amendment pursuant to section 50.59]." Id. "Furthermore, only changes to the'method of evaluation' are covered by 10 C.F.R. § 50.59(c)(2)(viii).

Changes to inputs to methods of evaluation are not covered by this provision" and, hence, do not trigger the requirement of seeking a license amendment.

Id.In other words, SCE claims that the standard in section 50.59(c)(2)(viii) has not been triggered because the tests or experiments embodied in its Unit 2 Return to Service Report are not inconsistent with the analysis or descriptions in the UFSAR. We disagree.The General Design Criteria in Appendix A of 10 C.F.R. Part 50 establish minimum requirements for the principal design criteria for water-cooled nuclear reactor plants. And as discussed supra note 44, General Design Criterion 14 refers to the reactor coolant boundary and includes steam generator tubes.55 The test or experiment proposed by SCE that must be the subject of a license amendment is required (1) to validate the vibration analysis that will become part of the FSAR (see supra note 52); and (2) to assure the steam generator tubes do not fail prematurely due to tube-to-tube wear and, thus, are able to satisfy their design bases. See id. (quoting 64 Fed.Reg. at 53,598).

36 Section 5.4.2.3.1 of the SONGS FSAR analyzes the maintenance of steam generator tube integrity.

Subsection 5.4.2.3.1.3.A describes the "Degraded Tube Evaluation." Its methodology essentially consists of calculating the maximum thinning for which tube integrity can be assured.5 6 Additionally, an inspection program, defined in Technical Specification 5.5.2.11, assures that tubes are removed from service before they reach maximum wall thinning.5 7 SCE's experience with SONGS Unit 3 forcefully demonstrates that the current analysis used to support the maintenance of steam generator tube integrity is inadequate for the replacement steam generators.

More specifically, the current analysis underlying tube inspections to prevent maximum thinning is inadequate to assure tube integrity in light of the accelerated wear mechanism that might occur in this type of steam generator, and that did occur in the Unit 3 steam generators.

Without question, the current analysis described in the FSAR failed to achieve its intended purpose, and it must therefore be changed. We view this change as sufficiently significant to trigger the license amendment requirement of section 50.59(c)(2)(viii), because it is "[i]nconsistent with the analyses or descriptions in the [UFSAR]." 10 C.F.R. § 50.59(a)(6)(ii).

Indeed, this change is a radical deviation from the prior analysis and description in the UFSAR, because without this change, tube integrity cannot be assured for the SONGS steam generators.

In sum, we conclude that SCE's Unit 2 Restart Plan, if implemented, would (1) grant SCE authority to operate without the ability to comply with all technical specifications; (2) grant SCE authority to operate beyond the scope of its existing license; and (3) grant SCE authority to 56 See SONGS UFSAR at 5.4-24, section 5.4.2.3.1.3.A.

57 See SONGS Unit 2 Technical Specification, section 5.5.2.11.

37 operate its replacement steam generators in a manner that constitutes a test or experiment that meets the criteria in 10 C.F.R. § 50.59(c)(2)(viii) for seeking a license amendment.

For these three independent reasons, this CAL process constitutes a de facto license amendment proceeding that is subject to a hearing opportunity under section 189a of the AEA.D. Because Our Resolution of the First Referred Issue Grants Petitioner All the Relief Its Contention Seeks, the Second Issue Referred by the Commission Is Moot The second issue referred to this Licensing Board is whether Petitioner "meets the standing and contention admissibility requirements of 10 C.F.R. § 2.309." CLI-12-20, 76 NRC at-(slip op. at 5).8 In its contention, Petitioner claims that "SONGS cannot be allowed to restart without a license amendment and attendant adjudicatory public hearing as required by 10 C.F.R. § 2.309, in which Petitioner and other members of the public may participate." Petition to Intervene at 16.In the course of resolving the first issue referred by the Commission (supra Part II.C), we concluded that this CAL process constitutes a de facto license amendment proceeding that is subject to a hearing opportunity.

As Petitioner conceded during oral argument (see Tr. at 29), such a conclusion grants all the relief sought in its contention.

Petitioner's contention, therefore, is moot.Were we to adjudicate either (1) the admissibility of a moot contention, or (2) the standing of a petitioner who sought to adjudicate a moot contention, we would be issuing an advisory opinion in derogation of Commission precedent.

This we decline to do. See U.S.Dep't of Energy (High-Level Waste Repository), CLI-08-21, 68 NRC 351, 352 (2008); accord 58 SCE urged this Board to resolve the standing and contention admissibility issues before considering the de facto license amendment issue. See Tr. at 63-65. The NRC Staff and Petitioner disagreed (see Tr. at 138 (NRC Staff); Tr. at 150 (Petitioner)), arguing that SCE's suggested approach was inconsistent with the Commission's unequivocal directive "to consider whether: (1) the [CAL] ... constitutes a de facto license amendment that would be subject to a hearing opportunity

... ; and, if so, (2) whether the petition meets the standing and contention admissibility requirements." CLI-1 2-20, 76 NRC at __ (slip op. at 5). We agree with the NRC Staff and Petitioner that SCE's suggested approach is at odds with the Commission's clearly expressed instruction in CLI-12-20.

38 Texas Utilities Generating Co. (Comanche Peak Steam Elec. Station), ALAB-714, 17 NRC 86, 94 (1983).59 Ill. CONCLUSION For the foregoing reasons, we resolve the first issue referred by the Commission in the affirmative, concluding that the CAL process for SONGS Units 2 and 3 constitutes a de facto license amendment proceeding that is subject to a hearing opportunity under section 189a of the AEA. Our resolution of the first issue grants Petitioner the relief it seeks in its contention; namely, the opportunity for an adjudicatory hearing incident to the license amendment proceedings for the restart of Units 2 and 3. Petitioner's contention is thus moot, which renders moot the second issue referred by the Commission.

The proceeding before this Board is therefore terminated.

59 "It is well established that, absent compelling reasons, the Commission adheres to the ,case' or 'controversy' doctrine in its adjudicatory proceedings." Hydro Resources, Inc. (P.O.Box 777, Crownpoint, New Mexico 87313), LBP-05-17, 62 NRC 77, 91 (2005) (citing Texas Utilities Elec. Co. (Comanche Peak Steam Elec. Station), CLI-93-10, 37 NRC 192, 200 n.28 (1993)). Pursuant to this doctrine, a justiciable controversy must involve parties who raise questions "presented in an adversary context and in a form historically viewed as capable of resolution through the judicial process." Flast v. Cohen, 392 U.S. 83, 95 (1968). When -- as is the case here -- a petitioner obtains the relief it is seeking before the admissibility of its contention is resolved, the admissibility vel non of the contention is no longer justiciable, because it no longer presents a live controversy involving a true clash of interests that is susceptible to meaningful adjudicative relief. Cf. Moore v. Charlotte-Mecklenburg Bd. of Ed., 402 U.S. 47, 48 (1971) (per curiam) (dismissing appeal for lack of live controversy where both litigants desired the same result); David B. Kuhl (Denial of Senior Reactor Operator License), LBP-09-14, 70 NRC 193, 195-96 (2009) (dismissing hearing request as moot where petitioner's claim was not susceptible to meaningful adjudicative relief).

39 If a party wishes to appeal this decision, it must file a petition for review with the Commission within 25 days after service of this decision.

See 10 C.F.R. § 2.341(b)(1).

Unless otherwise authorized by law, a party to an NRC adjudicatory proceeding must seek Commission review before seeking judicial review of an agency action. See id.It is so ORDERED.THE ATOMIC SAFETY AND LICENSING BOARD IRA, E. Roy Hawkens, Chairman ADMINISTRATIVE JUDGE IRA!Dr. Anthony J. Baratta ADMINISTRATIVE JUDGE IRA!Dr. Gary S. Arnold ADMINISTRATIVE JUDGE Issued at Rockville, Maryland this 13th day of May 2013.

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of SOUTHERN CALIFORNIA EDISON CO.(San Onofre Nuclear Generating Station -Units 2 and 3))))))Docket Nos. 50-361-CAL 50-362-CAL CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing MEMORANDUM AND ORDER (Resolving Issues Referred by the Commission in CLI-12-20)

-LBP-13-07 have been served upon the following persons by Electronic Information Exchange.Office of Commission Appellate Adjudication U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: ocaamail(@nrc..ov Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Mail Stop -T-3 F23 Washington, DC 20555-0001 E. Roy Hawkens Chief Administrative Judge E-mail: roy.hawkens~cnrc..ov Anthony J. Baratta Administrative Judge Email: anthony.baratta(,nrc..qov Gary S. Arnold Administrative Judge Email: pary.arnold(.nrc.qov U.S. Nuclear Regulatory Commission Office of the Secretary of the Commission Mail Stop O-16C1 Washington, DC 20555-0001 Southern California Edison Company Douglas Porter, Esq.Director and Managing Attorney Generation Policy and Resources Law Department 2244 Walnut Grove Ave., GO1, Q3B, 335C Rosemead, CA 91770 Email: douqlas.porter(csce.com Counsel for Licensee Morgan, Lewis & Bockius, LLP 1111 Pennsylvania, Ave. N.W.Washington, D.C. 20004 Paul M. Bessette, Esq.Kathryn M. Sutton, Esq.Stephen J. Burdick, Esq.Steven P. Frantz, Esq.William E. Baer, Jr.Mary Freeze, Legal Secretary Lena M. Long, Legal Secretary E-mail: pbessefte(Dmorqanlewis.com sburdickC-morqanlewis.com ksutton morqanlewis.com wbaerdmorganlewis.com sfrantz(omorganlewis.com mfreeze(cmorganlewis.com llona'moraanlewis.com Hearing Docket E-mail: hearinqdocketanrc.gov San Onofre Nuclear Generating Station, Units 2 and 3, Docket Nos. 50-361 and 50-362-CAL MEMORANDUM AND ORDER (Resolving Issues Referred by the Commission in CLI-12-20)

-LBP-13-07 U.S. Nuclear Regulatory Commission Office of the General Counsel Mail Stop 15 D21 Washington, DC 20555-0001 Edward Williamson, Esq.David Roth, Esq.Catherine Kanatas, Esq.David Cylkowski, Esq.Jeremy Wachutka, Esq.Email: edward.williamson(cnrc.gov david. roth(cnrc.gov catherine.kanatas(onrc.gov david.cylkowski(cnrc.gov ieremy.wachutka(o-nrc.oov OGG Mail Center: oqcmailcenter(o.nrc.gov Dated at Rockville, Maryland this 1 3 1h day of May, 2013 Friends of the Earth Ayres Law Group 1707 L St., NW Suite 850 Washington, D.C. 20036 Richard E. Ayres, Esq.Jessica L. Olson, Esq.Kristin L. Hines, Esq.Email: avresr(aayreslawgroup.com olsoni*(ayreslawqroup.com hineskdayreslawgroup.com Natural Resources Defense Council Geoffrey H. Fettus, Esq.1152 1 5 th Street, NW Suite 300 Washington, DC 20005 Email: qfettus(fnrdc.orq

[Original signed by Herald M. Speiser Office of the Secretary of the Commission 2