ML13010A387

Response to Friends of the Earth 10 CFR 2.206 Petition

ML13010A387
Person / Time
Site:	San Onofre
Issue date:	01/09/2013
From:	St.Onge R Southern California Edison Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML13010A387 (19)

Text

SOUTHERN CALIFORNIA Richard J. St. Onge EDSO E D I SE.0 N 0 Director, Nuclear Regulatory Affairs and Emergency Planning An EDISON INTERNATIONAL Company January 9, 2013 10 CFR 2.206 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Subject:

Docket Nos. 50-361 and 50-362 Response to Friends of the Earth 10 CFR 2.206 Petition San Onofre Nuclear Generating Station, Units 2 and 3

Dear Sir or Madam:

On June 18, 2012, Friends of the Earth (FOE) submitted a petition to intervene and request for hearing to the Commission claiming, among other things, that Southern California Edison Company (SCE) violated 10 CFR 50.59 when it replaced the San Onofre Nuclear Generating Station (SONGS) steam generators in 2010 and 2011 without a license amendment for certain design changes. In a November 8, 2012 decision (CLI-12-20), the Commission denied FOE's request for hearing regarding the alleged 10 CFR 50.59 violation, and referred that portion of the request to the Executive Director for Operations for consideration as a 10 CFR 2.206 Petition.

SCE provides the attached response to this 10 CFR 2.206 Petition and requests that the Nuclear Regulatory Commission (NRC) deny the Petition in its entirety. The NRC has already evaluated SCE's 10 CFR 50.59 evaluations for the replacement steam generators, and FOE's Petition identifies no significant new information. Additionally, nothing identified by FOE constitutes a violation of 10 CFR 50.59. Therefore, under the NRC's guidelines for petitions under 10 CFR 2.206, the Petition should be denied.

There are no new regulatory commitments contained in this letter. If you have any questions or require additional information, please contact me at (949) 368-6240.

Sincerely,

Enclosure:

Southern California Edison Company's Response to Friends of the Earth 10 CFR 2.206 Petition cc: B. J. Benney, NRC SONGS Petition Manager R. W. Borchardt, NRC Executive Director for Operations E. E. Collins, Regional Administrator, NRC Region IV R. Hall, NRC Project Manager, San Onofre Units 2 and 3 G. G. Warnick, NRC Senior Resident Inspector, San Onofre Units 2 and 3 A. T. Howell, Team Manager SONGS Special Project P.O. Box 128 062 San Clemente, CA 92672 . --

ENCLOSURE Southern California Edison Company's Response to Friends of the Earth 10 CFR 2.206 Petition I. PURPOSE On June 18, 2012, Friends of the Earth (FOE) submitted a petition to intervene and request for hearing (Petition) claiming, among other things, that Southern California Edison Company (SCE) violated 10 CFR 50.59 when it replaced the San Onofre Nuclear Generating Station (SONGS) steam generators (SGs) without a license amendment for certain design changes.

FOE's Petition attached a declaration from Arnold Gundersen (Gundersen Declaration), dated May 31, 2012, that addresses in part the 50.59 allegations.

In a decision on November 8, 2012, the Commission decided to treat that claim as a petition under 10 CFR 2.206. SCE is providing the following response to FOE's 2.206 Petition. As demonstrated in this response, the design changes associated with the replacement steam generators (RSGs) did not require a license amendment pursuant to 10 CFR 50.59, and FOE's 2.206 Petition should be denied.

II. BACKGROUND SONGS Units 2 and 3 are pressurized water reactors using a Combustion Engineering design.

SCE procured RSGs for SONGS Units 2 and 3 in the mid-2000s from Mitsubishi Heavy Industries (MHI) and received and installed the RSGs in 2009-2011. The RSGs included a number of design changes relative to SONGS' original steam generators (OSGs).

On June 27, 2008, SCE submitted a license amendment request (LAR) under 10 CFR 50.90 for certain technical specification changes associated with the RSGs. 1 The LAR explained that

"[t]he proposed changes reflect revised SG inspection and repair criteria and revised peak containment post-accident pressure resulting from installation of the replacement SGs."2 The NRC published a FederalRegister notice on the LAR on September 23, 2008, including a proposed determination that the amendment involves no significant hazards consideration.3 The NRC approved the LAR on June 25, 2009.4 SCE also performed 10 CFR 50.59 evaluations for the design changes associated with the RSGs and concluded that those changes did not require a license amendment. These evaluations consisted of an initial screening of the activities and changes resulting from the RSGs, and then a full 50.59 evaluation was performed for the issues identified by the initial screening. Those issues are documented in SCE's periodic Facility Change Report submitted 1 Letter from J. Reilly, SCE, to NRC, Amendment Application Numbers 252 and 238, Proposed Change Number NPF-10/15-583, Replacement Steam Generators, San Onofre Nuclear Generating Station, Units 2 and 3 (June 27, 2008), available at ADAMS Accession No. ML081830421.

2 Id. at 2.

3 Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 73 Fed. Reg. 54,862, 54,867-868 (Sept. 23, 2008).

Letter from J. Hall, NRC, to R. Ridenoure, SCE, San Onofre Nuclear Generating Station, Units 2 and 3 - Issuance of Amendments Re: Technical Specification Changes in Support of Steam Generator Replacement (TAC Nos. MD9160 and MD9161) (June 25, 2009), availableat ADAMS Accession No. ML091670298.

Page 1

to the NRC according to 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2). 5 The Facility Change Report describes the 50.59 evaluations as follows:

This activity replaces the design disclosure documentation and reference documentation for the San Onofre Nuclear Generating Station (SONGS) Unit 2 Original Steam Generators (OSGs) with that for the Replacement Steam Generators (RSGs) and performs functional testing of the RSGs. This replacement and testing is required as a result of physically replacing the OSGs with RSGs.

Having the OSGs replaced with the RSGs will improve the efficiency and reliability of Unit 2 by replacing a large number of plugged or otherwise degraded heat transfer tubes in each OSG with new tubes made from thermally-treated Alloy 690, which is less susceptible to degradation than the mill annealed Alloy 600 material used for OSG heat transfer tubing. Replacement of the steam generators is a replacement in-kind in terms of an overall fit, form, and function with no, or minimal, permanent modifications 6

to the plant [Systems, Structures, and Components] (SSCs).

Based on the 50.59 evaluations, the Facility Change Report stated that SCE concluded that the changes could be made without prior NRC approval. 7 The NRC has performed a number of inspections of the RSG project and the associated 50.59 evaluations. These inspections are documented in Inspection Report 2009-007 for the Unit 2 RSG, Inspection Report 2010-009 for the Unit 3 RSG, Inspection Report 2012-007 for the Augmented Inspection Team (AIT) inspection following the recent steam generator degradation issues, and Inspection Report 2012-010 for an AIT follow-up inspection. None of these inspections has concluded that SCE should have sought license amendments for the design changes in the RSGs.

SCE replaced the Unit 2 steam generators in January 2010 and the Unit 3 steam generators in January 2011. On January 31, 2012, SCE identified a leak in a tube in one of the Unit 3 RSGs.

This leak was well below allowable limits in the technical specifications, and presented no hazard to the public health and safety. Pursuant to established procedures, SCE shut down Unit 3. At the time Unit 2 was already shutdown and undergoing a refueling outage.

Ill. REGULATORY STANDARDS A. 10 CFR 2.206 10 CFR 2.206 provides the regulatory mechanism for requests for action. Section 2.206(a) states in part that "[a]ny person may file a request to institute a proceeding pursuant to § 2.202 to modify, suspend, or revoke a license, or for any other action as may be proper."

Letter from R. St. Onge, SCE, to NRC, Facility Change Report, San Onofre Nuclear Generating Station Units 2 and 3 and the Independent Spent Fuel Storage Installation (June 8, 2011),

available at ADAMS Accession No. ML11161A158.

6 Id., Enclosure 1A, at 4.

7 Id.

Page 2

The NRC's Management Directive (MD) for reviewing 10 CFR 2.206 petitions is provided in MD 8.11, "Review Process for 10 CFR 2.206 Petitions," which was most recently revised on October 25, 2000. MD 8.11 provides Handbook 8.11, which details the procedures for staff review and disposition of 2.206 petitions.

MD 8.11 explains that the NRC will reject a petition under certain circumstances, including:

The petitioner raises issues that have already been the subject of NRC staff review and evaluation either on that facility, other similar facilities, or on a generic basis, for which a resolution has been achieved, the issues have been resolved, and the resolution is applicable to the facility in question. This would include requests to reconsider or reopen a previous enforcement action (including a decision not to initiate an enforcement action) or a director's decision. These requests will not be treated as a 2.206 petition unless they present significant new information. 8 MD 8.11 also explains that the NRC, if appropriate, will request the licensee to provide a voluntary response to the NRC on the issues raised in a 2.206 petition, but "[t]he licensee may voluntarily submit information relative to the petition, even if the NRC staff has not requested any such information."9 B. 10 CFR 50.59 The applicable provisions in 10 CFR 50.59 state as follows:

(a) Definitions for the purposes of this section:

(1) Change means a modification or addition to, or removal from, the facility or procedures that affects a design function, method of performing or controlling the function, or an evaluation that demonstrates that intended functions will be accomplished.

(2) Departurefrom a method of evaluation describedin the FSAR (as updated)used in establishingthe design bases or in the safety analyses means:

(i) Changing any of the elements of the method described in the FSAR (as updated) unless the results of the analysis are conservative or essentially the same; or (ii) Changing from a method described in the FSAR to another method unless that method has been approved by NRC for the intended application.

(3) Facilityas describedin the final safety analysis report (as updated) means:

(i) The structures, systems, and components (SSC) that are described in the final safety analysis report (FSAR) (as updated),

(ii) The design and performance requirements for such SSCs described in the FSAR (as updated), and 8 Handbook 8.11, Part Ill.C.2.b.

9 Id. Part IV.A.2.a.

Page 3

(iii) The evaluations or methods of evaluation included in the FSAR (as updated) for such SSCs which demonstrate that their intended function(s) will be accomplished.

(c)(1) A licensee may make changes in the facility as described in the final safety analysis report (as updated), make changes in the procedures as described in the final safety analysis report (as updated), and conduct tests or experiments not described in the final safety analysis report (as updated) without obtaining a license amendment pursuant to § 50.90 only if:

(i) A change to the technical specifications incorporated in the license is not required, and (ii) The change, test, or experiment does not meet any of the criteria in paragraph (c)(2) of this section.

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

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

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

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

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

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

(vi) Create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the final safety analysis report (as updated);

(vii) Result in a design basis limit for a fission product barrier as described in the FSAR (as updated) being exceeded or altered; or (viii) Result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses.

The NRC has established guidance for implementing 10 CFR 50.59 in Regulatory Guide 1.187, "Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments."1 0 In Regulatory Guide 1.187, the NRC staff endorsed Revision 1 of NEI 96-07, "Guidelines for 10 CFR 50.59 Evaluations," and concluded that NEI 96-07 "provides methods that are 10 Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments (Nov. 2000).

Page 4

acceptable to the NRC staff for complying with the provisions of 10 CFR 50.59."l1 Therefore, NEI 96-07 provides approved guidance for evaluating whether 10 CFR 50.59 allows changes to the facility or procedures and conduct of tests or experiments without prior NRC approval.

As discussed in Section 1.3 of NEI 96-07, changes are evaluated under 10 CFR 50.59 using a multi-step process. 12 First, a licensee must determine that a proposed change is safe and effective through appropriate engineering and technical evaluations. The 10 CFR 50.59 process is then applied to determine if a license amendment is required prior to implementation.

Thus, the 50.59 process is a licensing process to determine whether a change requires a license amendment from the NRC; it is not a process to determine whether the change is safe.

Instead, the safety determination is made before the 50.59 process is initiated.

The 50.59 process utilizes three basic steps:13

1) Applicability and Screening: Determine if a 10 CFR 50.59 evaluation is required.

2) Evaluation: Apply the eight evaluation criteria of 10 CFR 50.59(c)(2) to determine if a license amendment must be obtained from the NRC.

3) Documentation and reporting: Document and report to the NRC activities implemented under 10 CFR 50.59.

Section 4.1 of NEI 96-07 further addresses the applicability of 10 CFR 50.59. Section 4.1.1 explains that 10 CFR 50.59 is applicable to tests or experiments not described in the Updated Final Safety Analysis Report (UFSAR) and to changes to the facility or procedures as described in the UFSAR, with a few exceptions.14 One exception is that a change to the technical specifications must be made using the license amendment process in 10 CFR 50.90.15 Another exception is that changes to the facility or procedures that are controlled by other more specific requirements and criteria established by regulation (e.g., 10 CFR 50.54(q) change requirements for emergency plans) are excluded from the scope of 10 CFR 50.59.16 If 10 CFR 50.59 does not apply to an issue, then no 50.59 evaluation is required for that issue.

Section 4.2 of NEI 96-07 further explains that, once it has been determined that 10 CFR 50.59 is applicable to a proposed activity, a screening should be performed to determine whether a 50.59 evaluation is required.' 7 If the screening determines that an activity is (1) a change to the facility or procedures as described in the UFSAR or (2) a test or experiment not described in the UFSAR, then a 50.59 evaluation is needed.' 8 In making the first determination, a design modification may be screened out if it does not affect:

• a design function of an SSC;

• a method of performing or controlling the design function; or 11 Id. at 2; NEI 96-07, Guidelines for 10 CFR 50.59 Evaluations (Rev. 1, Nov. 2000).

12 NEI 96-07 at 4.

13 Id.

14 Id. at 23.

15 Id.; 10 CFR 50.59(c)(1)(i).

16 NEI 96-07 at 23; 10 CFR 50.59(c)(4).

17 NEI 96-07 at 29.

18 Id.

Page 5

• an evaluation that demonstrates that intended design functions will be accomplished.1 9 Additionally, Section 4.2.1 of NEI 96-07 further explains that changes that have neutral or positive effects on these three areas may be screened out, because only adverse changes have the potential to increase the likelihood of malfunctions, increase consequences, create new accidents or otherwise meet the 10 CFR 50.59 evaluation criteria.

If there is a change in the method of evaluation as described in the UFSAR, then a 50.59 evaluation is required. However, a change in the method of evaluation does not require a license amendment if:

• the results of the new analysis are conservative or essentially the same; or

• the new method has been approved by the NRC for the intended application.2 ° IV. EVALUATION OF FOE PETITION A. Issues Raised by FOE Already Have Been Reviewed and Evaluated by NRC As noted above, NRC guidance in MD 8.11 states that a 10 CFR 2.206 petition will be rejected if the petition raises issues that already have been evaluated by the NRC, and there is no significant new information. 2 1 Because the NRC already has reviewed and evaluated the issues raised by FOE that are part of the 10 CFR 2.206 Petition, and FOE has not provided any significant new information, the Petition should be rejected.

The NRC staff already has reviewed whether the RSGs were appropriately evaluated under 10 CFR 50.59. This included reviews that took place at the time of the Units 2 and 3 steam generator replacements. More recently, the NRC staff reviewed the 50.59 evaluations for the RSGs in detail as part of the AIT inspection. The staff's review of the 50.59 evaluations for the RSGs is documented in detail in the July 18, 2012 AIT Report.2 2 The AIT Report explains that the NRC staff had conducted inspections regarding the Units 2 and 3 RSGs, which "included a review of selected portions of modifications associated with the replacement steam generators to determine if the changes were done in accordance with 10 CFR 50.59."23 The AIT Report states that the results of the RSG inspections are documented in Inspection Reports 2009-007 and 2010_009.24 These inspection reports do not identify any violations related to 10 CFR 50.59 evaluations.2 5 The AIT Report stated that "[t]he team determined that the licensee's evaluation for changes in the updated final safety analysis report's design methodologies for the replacement steam 19 Id.

20 Id. at 14-15.

21 Handbook 8.11, Part III.C.2.b.

22 NRC Inspection Report 05000361/2012007 and 05000362/2012007, San Onofre Nuclear Generating Station - NRC Augmented Inspection Team Report (July 18, 2012) (AIT Report),

available at ADAMS Accession No. ML12188A748.

23 Id. at 4-5.

24 Id.

25 NRC Inspection Report 05000361/2009007, San Onofre Nuclear Generating Station - Unit 2 Steam Generator Replacement Project (Mar. 4, 2010), available at ADAMS Accession No. ML100630838; NRC Inspection Report 05000362/2010009, San Onofre Nuclear Generating Station - Unit 3 Steam Generator Replacement Project (May 10, 2011), available at ADAMS Accession No. ML111300448.

Page 6

generators was consistent with SONGS procedures for implementation of 10 CFR 50.59 requirements."2 6 The staff explained that SCE determined as part of its 50.59 screening evaluation that the proposed activity did not adversely affect a design function, or the method of performing or controlling a design function described in the UFSAR; did not change a procedure in a manner that adversely affected how an UFSAR design function is performed or controlled; and did not involve a test or experiment not described in the UFSAR. This included review of the following UFSAR design functions as part of the screening:

• Steam Generator Design Functions

• Reactor Coolant System Structural Integrity

" Emergency Core Cooling System Performance

• Non-Loss of Coolant Accident Transients

• Containment Pressure-Temperature Analysis

• Low Temperature Overpressure Protection

" Reactor Protection System, Engineered Safety Features Actuation System, Core Operating Limit Supervisory System, and Core Protection Calculations

• Nuclear Steam Supply System Performance

• Non-Safety Related Control Systems Performance2 8 As detailed in the AIT Report, the 50.59 screening process identified three methods of analysis described in the UFSAR that were affected by the RSG project and required further evaluation against the 50.59 criteria: (1) Seismic Analysis of Reactor Vessel29Internals; (2) Reactor Coolant System Structural Integrity; and (3) Tube Wall Thinning Analysis.

Based on this review, the AIT Report concluded: "The team determined that no significant differences existed in the design requirements of Unit 2 and Unit 3 replacement steam generators. Based on the updated final safety analysis report description of the original steam generators, the team determined that the steam generators major design changes were reviewed in accordance with the 10 CFR 50.59 requirements."30 The AIT Report also documented the Office of Nuclear Reactor Regulation (NRR) review of the 50.59 evaluation for the RSGs. 3 1 This review was detailed, and included the following scope:

The NRR technical specialist reviewed all of the design changes associated with the replacement steam generators to determine whether the changes to the facility or procedures, as described in the updated final safety analysis report, had been reviewed and 26 AIT Report at 33-34.

27 Id. at 34.

28 Id.

29 Id. at 34-35.

30 Id. at 36.

31 Id. at 63-65.

Page 7

documented in accordance with 10 CFR 50.59 requirements. The technical specialist reviewed the various information used by SCE to review the changes being made to the replacement steam generators, including calculations, analyses, design change documentation, procedures, the updated final safety analysis report, the technical specifications, and plant drawings. The evaluation process used by the technical specialist included determining if the design changes to the replacement steam generators were a change to the facility or procedures as described in the updated final safety analysis report or a test or experiment not described in the updated final safety analysis report. The technical specialist also verified that safety issues related to the changes were resolved. The technical specialist compared the safety evaluations and supporting documents to the guidance and methods provided in NEI 96-07, "Guidelines for 10 CFR 50.59 Implementation," Revision 1, as endorsed by NRC Regulatory Guide 1.187, "Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments," to determine the adequacy of the 10 CFR 50.59 evaluations.3 2 This review resulted in one unresolved item (URI 05000362/2012007-10), "Evaluation of Departure of Method of Evaluation for 10 CFR 50.59 Processes."3 3 This unresolved item, however, was closed during a November 9, 2012 AIT follow-up report.34 Although the AIT follow-up report identified a minor violation regarding 10 CFR 50.59(d)(1), the NRC 3concluded that the issues related to the unresolved item did not require a license amendment. 1 The AIT members reviewed a wide variety of documents related to the RSGs as part of its inspection, including calculations, design basis documents, design change notifications/supplier deviation requests, drawings, engineering reports, modifications, nuclear notifications, procedures, vendor documents, and other miscellaneous documents.36 Among these, the AIT members specifically reviewed SCE's 50.59 screening and evaluation for the RSGs.3 7 In summary, the NRC performed a detailed review of the 50.59 process utilized for the SONGS RSGs, and did not identify any design changes that required a license amendment (other than the LAR approved by the NRC on June 25, 2009). The NRC's review was very broad and encompassed all of the design changes associated with the RSGs. The AIT Report concluded:

"Based on the updated final safety analysis report description of the original steam generators, the steam generators major design changes were appropriately reviewed in accordance with the 10 CFR 50.59.38 Therefore, the NRC's evaluation of the 50.59 process for the RSGs encompasses all of the issues raised by FOE in the 2.206 Petition. FOE does not identify any significant new information. Consistent with MD 8.11, the Petition should be rejected because it raises issues that already have been evaluated by the NRC.

32 Id. at 63-64.

33 Id. at 65.

34 NRC Inspection Report 05000361/2012010 and 05000362/2012010, San Onofre Nuclear Generating Station - NRC Augmented Inspection Team Follow-up Report (Nov. 9, 2012),

available at ADAMS Accession No. ML12318A342.

35 Id. at 22-26.

36 AIT Report, Attachment 1.

37 Id., Attachment 1, at 14.

38 Id. at ii.

Page 8

B. Issues Raised by FOE Do Not Require a License Amendment Per 10 CFR 50.59 Even had the NRC not already reviewed the 50.59 process utilized by SCE for the SONGS RSGs, the issues raised by FOE do not require a license amendment under 10 CFR 50.59.

These include allegations regarding design changes, changes in computer codes, and nonconformances in the RSGs. These categories are each discussed below.

1. Design Changes FOE argues that SCE violated 10 CFR 50.59 because it made "major changes" in the steam generators and the RSGs "differ significantly" from the OSGs.39 FOE claims that design changes "such as removal of the stay cylinder, replacement of the egg crate tube support with a broached plate tube support, or the thickening of the tube sheet" are "major design changes" and therefore required a license amendment. 40 That is not the correct standard under 10 CFR 50.59 for evaluating whether a license amendment is needed. The fact that the design is changed, or even that a41design change is "major," does not automatically result in the need for a license amendment.

As discussed above, design changes may42 be screened out under 10 CFR 50.59 if the changes do not adversely affect a design function. As explained in UFSAR Section 5.1, the RSGs have two design functions: (1) Function as a part of the reactor coolant pressure boundary (RCPB) as a barrier to the release of fission products; and (2) Transfer the heat generated in the reactor from the reactor coolant system into the secondary system.

The design changes identified by FOE did not have an adverse effect on the heat transfer function of the RSGs, and FOE does not allege anything to the contrary. Additionally, as shown on the table provided in Appendix 1, the design changes identified by FOE did not have an adverse effect on the RCPB function of the RSGs. Because the design changes identified by FOE did not have an adverse effect on any design function, the changes were properly screened out, and there was no violation of 10 CFR 50.59. The adverse condition that later resulted in a tube leak was a deficiency associated with the design and was not known at the time the 50.59 evaluation was performed.

Mr. Gundersen provided a table that purports to show that certain design changes would increase the probability or consequences of an accident, create the possibility of a new accident, or have other effects.43 The design changes addressed in the table include removing the stay cylinder, changing the tubesheet, adding tubes, changing the tube support, adding a flow restrictor, additional water volume, and a feedwater distribution ring.4 4 Mr. Gundersen, 39 Petition at 2-3, 17.

40 Id. at 2 n.2.

41 FOE further alleges that "[t]he NRC failed to follow its own regulations, in particular 10 C.F.R. § 50.59, which require a formal licensing proceeding be convened and a license amendment granted before changes can be made to the facility that affect the final safety analysis." Id. at 18.

This too is the wrong standard. A change to the UFSAR or analyses found therein does not automatically require a license amendment. Instead, such changes require a license amendment only if they meet one of the eight criteria in 10 CFR 50.59(c)(2).

42 NEI 96-07 at 29.

43 Gundersen Declaration at 9-10; Petition at 18-19.

44 Gundersen Declaration at 9.

Page 9

however, provides no explanation or information to support those claims in the table. Therefore, the table does not provide a basis for a 50.59 violation. These design changes are further addressed in the table in Appendix 1.

In some places, Mr. Gundersen appears to be taking the position that any change in the UFSAR requires a license amendment under 50.59. That position is inconsistent with 10 CFR 50.59 and the associated NRC-endorsed guidance. As discussed above, a license amendment is needed only if (1) the design change adversely affects a design function; and (2) the impact on the design function satisfies one of the eight criteria in 10 CFR 50.59(c)(2). This is not the case for the design changes associated with the RSGs.

Moreover, the AIT Report evaluated and rejected the very issues raised by FOE regarding design changes. The AIT Report stated:

With regard to the major design changes between the original and replacement steam generators, the updated final safety analysis report did not specify how the original steam generators relied on special design features such as the stay cylinder, tubesheet, tube support plates, or the shape of the tubes to perform the intended safety functions. ...

Consistent with [NEI 96-07], SCE's 50.59 screening evaluated the differences in subcomponents between the original steam generators and replacement steam generators as to whether the differences adversely affected the design function (reactor coolant pressure boundary) of the steam generators. The replacement steam generators were designed and fabricated in accordance with quality assurance requirements, and 10 CFR 50.59 does not require the licensee to presume deficiencies in the design or fabrication....

The team determined that no significant differences existed in the design requirements of Unit 2 and Unit 3 replacement steam generators. Based on the updated final safety analysis report description of the original steam generators, the team determined that the steam generators major design changes were reviewed in accordance with the 10 CFR 50.59 requirements.45 In summary, SCE's 50.59 activities appropriately evaluated whether any design changes associated with the RSGs required a license amendment. This conclusion is consistent with the AIT Report, and FOE has not demonstrated that any of the design changes (if properly implemented) would have adversely affected a design function.

In addition, the design changes associated with the replacement steam generators are similar to design changes made in replacement steam generators for other plants, and the changes in those plants have not required a license amendment. As stated in a June 11, 2012 letter from NRC Chairman Jaczko to Senator Barbara Boxer:

45 AIT Report at 36.

Page 10

NRC regulations at 10 CFR 50.59 and associated guidance in Regulatory Guide 1.187 include criteria for a licensee to determine when a license amendment is required for proposed changes to a facility. Historically, RSGs have been evaluated against these criteria and no license amendment was required.

2. Changes in Computer Codes FOE also alleges that "[t]he computer code MHI used for design validation simply was not capable of analyzing the reactor design at San Onofre; rather, the code 46 was qualified only for Westinghouse generators, which are not similar to CE generators."

Changes in computer codes, however, do not require a 50.59 evaluation unless the code is different than that specified in the UFSAR. As stated in 10 CFR 50.59(c)(2)(viii), a change in a code requires a license amendment only if it represents a "departure from a method of evaluation described in the FSAR (as updated)"(emphasis added). The codes mentioned by FOE are not described in the UFSAR. Therefore, changes in those codes did not require a 50.59 evaluation, and did not involve any violation of 50.59.

This conclusion is consistent with the NRC staff's evaluation of this issue as described in the AIT Report:

The team noted that a key methodology for the design of the replacement steam generators was the thermal-hydraulic code used to model the flow conditions in the steam generators.

Mitsubishi's FIT-Ill thermal-hydraulic code was accepted by SCE for the design of the replacement steam generators. The team noted that the updated final safety analysis report did not describe the thermal-hydraulic code used for the design of the original steam generators and therefore the use of the FIT-Ill thermal-hydraulic code did not constitute a change in methodology or a change in an element of a methodology described in the updated final safety analysis report. The updated final safety analysis report did describe the computer code CRIB as the code used to analyze overall steam generator performance. As described in the updated final safety analysis report, CRIB was used to establish the recirculation ratio and fluid mass inventories as a function of power level in the original steam generators.4 7 As a result of its recent evaluations, SCE has determined that MHI's thermal-hydraulic analysis code did not predict the fluid elastic instability that occurred in the RSGs. That concern, however, was not known during the design and manufacturing of the RSGs. Therefore, those concerns could not have been a basis for a license amendment and do not provide any basis for an allegation that SCE violated 50.59 in 2009-2011.

In summary, SCE's 50.59 activities appropriately evaluated whether any changes in computer codes associated with the RSGs required a license amendment. This conclusion is consistent with the AIT Report, and FOE has not demonstrated that a change in codes represents a departure from a method of evaluation described in the UFSAR.

46 Petition at 21; Gundersen Declaration at 12.47 AIT Report at 35-36.

47 AIT Report at 35-36.

Page 11

3. Nonconformances in the RSGs FOE further refers to the leak in one of the RSGs and the unexpected tube wear as a basis for its allegation that SCE violated 10 CFR 50.59, claiming that the design changes resulted in risks not considered in the UFSAR.'4 The leak and unexpected tube wear are nonconforming conditions. If the RSGs had been designed and manufactured in accordance with the procurement specification, the leak and tube wear would not have occurred. These nonconformances were not known at the time the 50.59 evaluations were performed.

Therefore, the nonconformances do not indicate any violation of 10 CFR 50.59 or any need for a license amendment in 2009-2011.

As explained above, SCE evaluated all of the changes related to the RSGs, and the changes met the screening criteria for not performing a 50.59 evaluation, or a 50.59 evaluation determined that no license amendment was necessary. Later-identified errors in an evaluation or nonconformances do not mean that an earlier 50.59 evaluation, such as SCE's 50.59 analysis for the RSGs, was deficient or that a license amendment should have been obtained.

And FOE has not demonstrated otherwise in their 10 CFR 2.206 petition.

This concept is further discussed in Section 4.4 of NEI 96-07. As that section indicates, a 50.59 evaluation is not needed for a nonconforming condition, unless the licensee proposes to change its licensing basis to accept the nonconforming condition. Thus, 10 CFR 50.59 does not apply to nonconforming conditions unless the licensee accepts such conditions; instead, nonconforming conditions are addressed through a licensee's corrective action program pursuant to Appendix B to 10 CFR Part 50.

4. Other Claims by FOE Many of FOE's allegations pertain to the current safety of the RSGs. Those claims are being dealt with under the subject of NRC staff inspections and evaluations, and do not pertain to whether SCE violated 10 CFR 50.59 in 2009-2011.

V. CONCLUSIONS The NRC already has evaluated SCE's 10 CFR 50.59 evaluations for the RSGs, and FOE's Petition identifies no significant new information. Additionally, nothing identified by FOE warrants a license amendment under 10 CFR 50.59. Therefore, under the NRC's guidelines for petitions under 10 CFR 2.206, the Petition should be denied.

48 Petition at 3, 18, 20.

Page 12

APPENDIX 1 FOE Allegation, Anallysis The Petition states: "The key fabrication The number of tubes is not specified in the change in the new generators was the UFSAR, and therefore a change in the number decision to add almost 400 tubes to each of tubes did not involve a change in the steam generator, increasing the total number UFSAR.

of tubes by more than 4%." Petition at 17.

SCE's 50.59 Screens evaluated the addition of The Gundersen Declaration states that "[t]he tubes to the steam generators and determined key fabrication change supplanted to the San that the increase did not adversely affect the Onofre steam generators by the Edison/MHI function of the tubes as described in the team increased the total number of tubes in UFSAR, and therefore did not require a each steam generator by almost 400 tubes to license amendment.

more than 104 percent of each generator's original design. Each Original Steam Generator contained 9350 tubes while the There are other plants that have more tubes Replacement Steam Generators each contain per steam generator than SONGS Units 2 and 9727 tubes." Gundersen Declaration at 4-5. 3. The SONGS RSGs have 9727 tubes each.

In comparison, each of the original and replacement steam generators for one other Combustion Engineering (CE) facility, has more than 11,000 tubes, and each of the replacement steam generators for another CE facility has more than 10,000 tubes.

The Petition refers to "removing the stay The UFSAR did not specify that the OSGs cylinder, which functioned as a support pillar to relied on the stay cylinder to perform the the tubesheet into which the U-tubes are intended safety functions. SCE's 50.59 inserted." Petition at 17; 19-20. Screens evaluated the removal of the stay cylinder and determined that removal of the The Gundersen Declaration states that "[t]he stay cylinder in conjunction with other changes Edison/MHI decision to add additional tubes did not adversely affect the functions of the and replace this key support pillar was part of steam generators and therefore did not require the cascading fabrication changes that caused a license amendment.

additional stresses and steam generator failure." Gundersen Declaration at 5. As part of the analysis for the RSGs, MHI evaluated the stresses in the RSGs and determined that no part was stressed beyond the allowable limits in the ASME Code.

SCE has evaluated whether the change from the stay cylinder to a divider plate led to the tube damage. SCE has determined that this modification was not a causal factor in the tube-to-tube wear.

Page 13

FOE Al*e*ation .Analysis The Petition refers to "thickening the tubesheet FOE has not alleged that the thicker tubesheet to compensate structurally for the removal of adversely affected any design function of the the stay cylinder." Petition at 17. RSGs. Therefore, this statement does not provide a basis for alleging that SCE violated The Gundersen Declaration states that 10 CFR 50.59.

"[b]ecause the tubesheet was no longer supported in the center by the stay cylinder, The change in the tubesheet did not require a Edison/MHI required the fabrication of a license amendment. The increase in thicker tubesheet so that it could bear the thickness of the tubesheet did not adversely additional stress without a stay cylinder." affect the functions of the steam generators.

Gundersen Declaration at 5.

The Petition refers to "reducing the volume of FOE does not allege that the change in water in the steam generator." Petition at 17. volume had any adverse effect on a function of the RSGs or any other system.

The Gundersen Declaration states that "[t]his change in the tubesheet thickness meant yet The primary side volume for each RSG is another design change by reducing the slightly greater than that of each OSG due to a volume of water in the steam generator." larger internal volume of the tube bundle. The Gundersen Declaration at 5-6. secondary side masses of the RSGs and OSGs are approximately the same at full-load operating conditions. SCE's 50.59 Screens evaluated the change in the primary side volume and determined that it had an insignificant impact on the safety analyses.

The Petition refers to "changing the flow A change in a flow pattern does not, in and of pattern." Petition at 17. itself, require a license amendment under 10 CFR 50.59. FOE does not allege that the The Gundersen Declaration states that "[t]his change in the flow pattern had any impact on a change in the tubesheet thickness meant yet safety function.

another design change by ... changing the flow pattern." Gundersen Declaration at 5-6. At the time the RSGs were designed, MHI evaluated the flow patterns and determined that fluid elastic instability (FEI) would not occur. The experience with SONGS Unit 3 indicates that that conclusion was not correct, and that the RSGs do not conform to the procurement specification. However, that was not known at the time, and therefore provides no basis for a claim that SCE violated 10 CFR 50.59.

Page 14

FOE Allegiption Analy is The Petition refers to "reducing the inspection This change pertains to access for access area below the tubesheet." Petition at inspections. This change is not relevant to the

17. design functions of the RSGs to transfer heat and act as part of the RCPB. The RSG design The Gundersen Declaration states that "[t]his meets the 10 CFR 50 Appendix A General change in the tubesheet thickness meant yet Design Criteria for inspection access.

another design change by ... also reducing the inspection access area beneath the tubesheet that is required to fit personnel and equipment for tube inspection." Gundersen Declaration at 5-6.

The Petition states: "These design As part of the analysis for the RSGs, MHI modifications altered the structural loads on evaluated the stresses in the RSGs and the tubesheet, a critical safety consideration determined that no part was stressed beyond as the tubesheet serves as the key barrier the allowable limits in the ASME Code.

keeping radiation inside the containment." Therefore, these modifications did not affect Petition at 17. the function of the tubesheet with respect to the RCPB.

The Gundersen Declaration states that

"[c]hanging the structural loads on the tubesheet have not only affected the reliability of the steam generators but also should have raised a serious safety concern because the tubesheet is the key barrier keeping radiation inside the containment." Gundersen Declaration at 6.

The Petition states: "Adding tubes also FOE does not allege that this design change required increasing the nuclear reactor core adversely affected any design function of the flow, on which the original design basis safety RSGs or any other system.

calculations for cooling the reactor are based."

Petition at 17-18. The change in the primary coolant flow rate as a result of the RSGs was minimal; an increase The Gundersen Declaration states that from 198,000 gpm to 209,880 gpm. This is

"[f]abricating more tubes increased nuclear approximately a 5% difference. The difference reactor core flow, which was unacceptable in heat transfer rate was even less; an because it changed the original design basis increase from 5.819 x 10' Btu/hr to 5.900 x 10'9 safety calculations for cooling the reactor." Btu/hr, or approximately a 1% difference. As Gundersen Declaration at 6. discussed below, this increase in flow did not adversely affect the safety analysis of other systems and components.

SCE's 50.59 Screens evaluated the impact of the RSG changes on the reactor coolant system (RCS). SCE concluded that the functional and performance requirements for the RCS were met and that the RCS will continue to perform its design functions with the RSGs installed.

Page 15

-FOE ,Allegation Aay The Petition refers to "changes to the tube The tube support plates (TSPs) are located in supports in an attempt to avoid increased the portion of the steam generators with the vibration in the tubes." Petition at 18. straight legs of the steam generator tubes.

However, the area of concern with respect to "MHI changed to broached plate tube supports the tube vibration was in the U-bend area of in the replacement steam generator design." the steam generators.

Petition at 21.

SCE and MHI evaluated the impact of "Mr. Gundersen explains in his Declaration broached plates in their 50.59 reviews and how the flow resistance of the broached plate concluded that they did not affect the function designed by MHI is much higher than the of the steam generators and that they provided original CE egg crate design because of the sufficient margin against vibration.

reduced spacing of the tubes in the broached plate." Petition at 21. Furthermore, SCE has recently evaluated whether the change in tube support led to the The Gundersen Declaration states: "The flow tube damage, and has determined that this resistance of the Mitsubishi broached plate is modification was not a causal factor in the much higherthan that of the original tube-to-tube wear.

Combustion Engineering egg crate design because the tubes are so tightly packed in the Use of TSPs with broached holes has been original CE San Onofre steam generators. By common since at least the 1970s.

reviewing the documents thus far produced, it Furthermore, the majority of the RSGs in the appears that due to Mitsubishi's fabrication United States have used broached holes for experience with broached plates, both Edison the TSPs. In contrast, few plants (ten units in and Mitsubishi missed this key difference in total in the United States, including the OSGs the design and fabrication of the new San for SONGS Units 2 and 3) have used the egg Onofre steam generators." Gundersen crate design. Therefore, there is nothing Declaration at 10. improper in using broached hole TSPs.

The Gundersen Declaration also states that "Mitsubishi fabricated a broached plate design that allows almost no water to reach the top of the steam generator." Gundersen Declaration at 11.

Page 16

FOE Allegation Analysis The Petition states that "both MHI and SCE SCE and MHI did not "miss" this issue. The missed [that the design change] has resulted tube bundle flow analysis was the subject of in almost no water reaching the top of the special design review meetings between SCE steam generator, creating regions where the and MHI. MHI provided a thermal-hydraulic U-tubes are almost dry. Without liquid in the analysis as part of the original design of the mixture, there is no damping against vibration, RSGs that showed that there would be no FEl.

resulting in a severe fluid-elastic instability. A fundamental problem in the steam generator In retrospect, SCE and MHI have determined causing the vibration and, consequently, the that MHI's thermal-hydraulic code had errors, tube wear is that there is too much steam and and that it did not accurately predict the too little water at the top of the steam thermal-hydraulic conditions. This was not generators in the U-bend region." Petition at known in 2009-2011 and therefore does not

21. provide a basis for a claim that SCE violated 10 CFR 50.59 in 2009-2011.

The Gundersen Declaration states: "In response to the Edison/Mitsubishi steam generator changes, the top of the new steam generator is starved for water therefore making tube vibration inevitable." Gundersen Declaration at 11.

The Gundersen Declaration also states that

"[tlhe real problem in the replacement steam generators at San Onofre is that too much steam and too little water is causing the tubes to vibrate violently in the U-bend region."

Gundersen Declaration at 11.

The Gundersen Declaration states: "Edison FOE does not allege that this change had any welded a flow-restricting ring into the steam adverse impact of the design functions of the generator nozzle in order to reduce the flow of RSGs.

cooling water back into the reactor to the original design parameters, which also A flow restrictor was added to the steam outlet changes the flow distribution to the tubes." from the RSGs to reduce the amount of Gundersen Declaration at 6. energy released to the containment during a postulated steam line break. It also reduces the loads on the internals of the steam generators during such an event. These are design improvements and did not require a license amendment.

Additionally, a flow restrictor was added to the reactor coolant inlet for the steam generators.

Its purpose is to ensure that the maximum allowable reactor coolant flowrate will not be exceeded. As a result, it also did not require a license amendment.

Page 17

FOE AlIpgatipn Analvsis Anaysi FOE lleatio The Gundersen Declaration states: "The FOE does not allege that this change had any feedwater distribution ring inside the steam adverse impact on the design functions of the generator was also dramatically modified in RSGs.

order to avoid a serious flow induced water hammer." Gundersen Declaration at 6. With respect to water hammer, the RSG feedwater distribution ring has a gooseneck design for preventing a water hammer; whereas, the OSGs did not have the gooseneck. This change represents an improvement in the design. SCE's 50.59 Screens evaluated this change and determined that it did not adversely affect the function of the RSGs and therefore did not require a license amendment.

The Gundersen Declaration states: "The At the time the RSGs were designed, MHI maximum quality of the water/steam mixture at performed analyses that demonstrated that the the top of the steam generator in the U-Bend steam in any area of the tube bundles would region should be approximately 40 to 50 be low enough to provide the required percent, i.e. half water and half steam. With damping, and that the quality of the steam in the Mitsubishi design the top of the U-tubes the vast majority of the secondary side of the are almost dry in some regions. Without liquid steam generators would be even less.

in the mixture, there is no damping against Furthermore, MHI analyzed the potential for vibration, and therefore a severe fluid-elastic fluid elastic vibration, and determined that instability developed." Gundersen Declaration conditions were stable.

at 11.

SCE's root cause evaluation has determined that FEI did occur. However, SCE had no evidence of that beforehand. Thus, FOE's allegation provides no basis for concluding that a license amendment was needed in 2009-2011.

Page 18