ML14118A498

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Declaration of Mr. William A. Cross in Support of Fpl'S Answer Opposing Sace Request for Hearing
ML14118A498
Person / Time
Site: Saint Lucie NextEra Energy icon.png
Issue date: 04/28/2014
From: Cross W
Florida Power & Light Co
To:
NRC/OCM
SECY RAS
References
50-389-LA, License Amendment, RAS 25868
Download: ML14118A498 (26)
v • d • e


Text

ATTACHMENT 1 April 28, 2014 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Commission In the Matter of )

)

Florida Power & Light Company ) Docket No. 50-389

)

(St. Lucie Plant, Unit 2) )

DECLARATION OF MR. WILLIAM A. CROSS IN SUPPORT OF FPLS ANSWER OPPOSING SACE REQUEST FOR HEARING Mr. William A. Cross states as follows under penalty of perjury:

I. INTRODUCTION A. Declarants Background

1. I am the Nuclear Regulatory Programs Manager for Florida Power and Light Com-pany (FPL).

My educational background and qualifications include receiving a Bachelor of Sci-ence in Engineering (Nuclear Engineering) in 1974 from the University of Florida and holding a Nuclear Regulatory Commission (NRC) Senior Reactor Operator License for Crystal River Unit 3. Upon graduation from the University of Florida in 1974, I began my professional career with Florida Power Corporation as a Nuclear Licensing Engineer. From 1974 until 1975, I was employed by Tennessee Valley Authority as a Power Supply Planning Engineer. In 1975, I joined Florida Power Corporation at Crystal River Unit 3 as a Nuclear Plant Engineer. I continued my ca-reer advancement at Crystal River Unit 3 in positions with increasing responsibility and authority, including Technical Specifications Engineer, Reactor Engineer, Op-

erations Engineer and Nuclear Licensing Manager, until 1982. While employed by Florida Power Corporation, I received a Senior Reactor Operator License from the NRC. In 1982, I commenced work as a nuclear industry consultant specializing in nuclear licensing and regulatory compliance services with Southern Technical Ser-vices, Inc., a Florida corporation, which continued through 2002. From 2002 through 2008, I rendered small business consulting services as Crosservices, Incor-porated, a Florida Corporation. In 2008, I joined Florida Power & Light Company as Nuclear Projects Engineering Design Basis Supervisor, directing a team respon-sible for ensuring the technical and regulatory quality of the 10 C.F.R. § 50.59 re-views, as required, supporting engineering products including design changes, li-cense amendments, exemptions and responses to NRC requests for additional in-formation. As Design Basis Supervisor, I completed and maintain engineering qual-ifications for Section 50.59 Screening, Evaluation and Verification. In 2010, I joined the Nuclear Licensing Department as Fleet Project Licensing Manager, with responsibility for governance and oversight of Nuclear Projects and Nuclear Engi-neering activities needed to achieve timely and successful licensing actions for tran-sition to NFPA 805, implementation of Cyber Security Programs and resolution of GSI-191.

2. As Nuclear Regulatory Programs Manager, I am responsible for leadership of a team in the development of licensing actions for the nuclear fleet to meet regulatory requirements, improve operating margins, extend allowed completion times, im-prove outage effectiveness, avoid unwarranted plant shutdowns and prevent unnec-essary regulatory actions. Based on my design engineering qualifications, I con-2

ducted a review of the 10 C.F.R. § 50.59 Evaluation for the St. Lucie Unit 2 re-placement steam generators for technical and regulatory quality.

B. SACEs Hearing Request

3. I have reviewed and am familiar with Southern Alliance for Clean Energys (SACE) request for a hearing on the NRC Staffs alleged de facto amendment of FPLs operating license for St. Lucie Unit 2, which was filed with the Secretary on March 10, 2014. 1 I have also reviewed, and am familiar with, the Declaration pro-vided by Mr. Arnold Gundersen in support of SACEs hearing request. 2 II. REPLACEMENT OF ST. LUCIE UNIT 2 STEAM GENERATORS PROPERLY PURSUANT TO 10 C.F.R. § 50.59
4. SACE erroneously asserts that FPL made changes to the St. Lucie steam generators pursuant to 10 C.F.R. § 50.59 that required an amendment to FPLs operating li-cense, and that the design of the replacement steam generators put the reactors operation outside of both the original design basis and the license renewal design basis. Hearing Request at 11; Gundersen Decl. at ¶¶ 51-52.
5. In May of 2006, FPL sought from the NRC an amendment to St. Lucie Unit 2s op-erating license to change, among other things, the technical specifications relating to steam generator tube integrity in anticipation of steam generator replacement. In this regard, the amendment required that a program be established for the replace-ment steam generators to ensure that steam generator tube integrity would be main-tained. See Letter from Gordon L. Johnson, Acting Vice President, St. Lucie Plant, 1

Southern Alliance for Clean Energys Hearing Request Regarding De Facto Amendment of St. Lucie Unit 2 Op-erating License (Mar. 10, 2014) (Hearing Request).

2 Declaration of Arnold Gundersen (Mar. 9, 2014), Attachment 1 to Hearing Request (Gundersen Decl.).

3

to NRC, L-2006-094 (May 25, 2006), Attachment 1 at 10 (ADAMS Accession No. ML061510346). On July 18, 2006, the NRC published a notice of a proposed no significant hazards consideration determination concerning FPLs license amend-ment request, including a notice of opportunity for interested persons to file a hear-ing request. 71 Fed. Reg. 40,742, 40,743, 40,747-48 (July 18, 2006). No com-ments or hearing requests were filed in response to this notice. See 72 Fed. Reg.

33,779, 33,787-88 (June 19, 2007). The NRC granted the license amendment on May 29, 2007, before the steam generators were replaced. See Letter from Brenda L. Mozafari, NRC Office of Nuclear Reactor Regulation, to J. A. Stall, Senior Vice President, Nuclear and Chief Nuclear Officer, FPL (May 29, 2007) (ADAMS Ac-cession No. ML071490483); Letter from Brenda L. Mozafari, NRC Office of Nu-clear Reactor Regulation, to J. A. Stall, Senior Vice President, Nuclear and Chief Nuclear Officer, FPL (Aug. 8, 2007) (ADAMS Accession No. ML072140147) (cor-rection to amendment number in Technical Specification pages).

6. In part because the Technical Specifications had already been revised, the replace-ment itself was properly implemented pursuant to Section 50.59, which is the com-mon industry practice, after a thorough screening and evaluation. In accordance with Section 50.59, FPL prepared an evaluation demonstrating that the replacement steam generators satisfied the existing Updated Final Safety Analysis Report (UFSAR) acceptance criteria and Technical Specification Limits. In addition, the Section 50.59 evaluation found that none of the criteria warranting a license amendment, as specified in 10 C.F.R. § 50.59(c)(2), applied to the replacement steam generators for St. Lucie Unit 2. No further changes to St. Lucie Unit 2 Tech-4

nical Specifications were required, nor were any changes to the Emergency Operat-ing Procedures required as a result of the replacement of the steam generators. Af-fected sections of the UFSAR were updated, but the new steam generators were bounded by the original calculations for the original steam generators.

7. The 10 C.F.R. § 50.59 evaluation of the St. Lucie Unit 2 replacement steam genera-tors was reviewed by the NRC Staff as part of a three-month inspection, which in-cluded the St. Lucie Unit 2 steam generator replacement inspection. St. Lucie Nu-clear Plant - NRC Integrated Inspection Report 05000335/2007005, 05000389/2007005 (Feb. 1, 2008) (ADAMS Accession No. ML080350408), En-closure at 28. The inspectors reviewed the change screening and/or evaluation for all [plant change modifications] reviewed to verify that the modifications were properly evaluated in accordance with 10 CFR 50.59. Id. No findings of signifi-cance were identified. Id. at 33. See also Letter from Gordon L. Johnston, Site Vice President, St. Lucie Plant, to NRC, L-2008-148 (June 26, 2008), at 8 (AD-AMS Accession No. ML081840111) (summarizing the results of the Section 50.59 evaluation).
8. FPLs replacement of the St. Lucie Unit 2 steam generators in accordance with 10 C.F.R. § 50.59 did not modify the operating license or afford FPL any greater oper-ating authority.
9. SACE also erroneously claims that the baseline in-service inspection performed at Unit 2 during the current refueling outage must cover components that are listed in FPLs Aging Management Program (AMP) - specifically, the stay cylinder and 5

lattice tube supports - but that were removed or altered by FPL when it replaced the steam generators. Hearing Request at 12; Gundersen Decl. at ¶¶ 10, 57, 66.

10. FPL revised the design basis for the AMP in March 2007, prior to the steam genera-tor replacement, to reflect the fact that, as part of the replacement, the stay cylinder would be removed and tube support plates would be substituted for the lattice tube supports. Exhibit A to my Declaration shows the changes that were made to Table 5.2-2.7 in Revision 1 to the AMPs design basis document. Those changes were made as part of the engineering change process that FPL undertook in anticipation of Unit 2s steam generator replacement. The revisions are handwritten, consistent with FPL procedures. Those handwritten changes were reflected in the next revi-sion (Rev. 2) of the design basis document, dated April 2010. I am providing the relevant pages of that document in Exhibit B to this Declaration. Accordingly, as of the time the Section 50.59 analysis for the replacement steam generators was per-formed, and well before the current refueling outages in-service inspection, the St.

Lucie Unit 2 AMP reflected the removal of the stay cylinders as well as the re-placement of lattice tube supports with support plates and anti-vibration bars.

III. EXTENDED POWER UPRATE LICENSE AMENDMENT FOR ST. LUCIE UNIT 2 WITH REPLACEMENT STEAM GENERATORS

11. On February 25, 2011, FPL requested a license amendment to permit an extended power uprate (EPU) at St. Lucie Unit 2. Letter from Richard L. Anderson, Site Vice President, St. Lucie Plant, to NRC, L-2011-021 (Feb. 25, 2011) (ADAMS Ac-cession No. ML110730116). On September 1, 2011, the Commission published a notice of the license amendment request and of an opportunity to request a hearing.

6

76 Fed. Reg. 54,503 (Sept. 1, 2011). No hearing requests or petitions to intervene were filed.

12. In reviewing the EPU request, both the NRC Staff and the Advisory Committee on Reactor Safeguards (ACRS) evaluated the steam generator tube wear in the St.

Lucie Unit 2 steam generators in light of the unique tube-to-tube wear observed at San Onofre Nuclear Generating Station (SONGS) Unit 3. Certain of the requests for additional information (RAIs) from the NRC Staff addressed St. Lucie tube wear issues and the potential for tube-to-tube wear experienced at SONGS Unit 3.

On June 22, 2012, the ACRS Subcommittee on Power Uprates reviewed the St.

Lucie Unit 2 EPU license amendment request and the associated NRC safety evalu-ation. On July 11-13, 2012, the ACRS full committee reviewed the St. Lucie Unit 2 EPU license amendment request. In the letter to the NRC Executive Director for Operations recommending approval of the St. Lucie Unit 2 EPU license amend-ment, the ACRS commented on pertinent differences between the types and extent of steam generator wear observed at the two plants and concluded that [t]hese con-siderations and the licensees action plan adequately address concerns about [steam generator] tube integrity. Letter from J. Sam Armijo, Chairman, ACRS, to R.W.

Borchardt, NRC Executive Director of Operations (July 23, 2012), at 4 (ADAMS Accession No. ML12198A202), Exhibit C hereto. The Commission issued the re-quested license amendment on September 24, 2012. Letter from Tracy J. Orf, Pro-ject Manager, Plant Licensing Branch II-2, to Mano Nazar, Executive Vice Presi-dent and Chief Nuclear Officer, FPL (Sept. 24, 2012) (ADAMS Accession No. ML12235A463). The NRC Staffs Safety Evaluation Report supporting the license 7

amendment specifically found that the licensee demonstrated that tube integrity will continue to be maintained and will continue to meet the performance criteria in NEI 97-06 and the requirements of 10 C.F.R. § 50.55a following implementation of the proposed EPU. NRC staff also found the proposed EPU acceptable with respect to steam generator tube in-service inspection. Id., Enclosure 2 (Safety Evaluation) at 39.

IV. ST. LUCIE UNIT STEAM GENERATOR REPLACEMENT AND EPU HAVE BEEN PROPERLY IMPLEMENTED PURSUANT TO NRC REGULATIONS AND PROCEDURES

13. Particularly in view of the detailed safety evaluation conducted by the NRC Staff and the ACRS prior to granting the St. Lucie Unit 2 EPU license amendment, SACEs claim that the NRC Staff has informally amended FPLs operating license on multiple occasions by approving continued operation with equipment that is clearly outside the reactors design basis (Hearing Request at 12) is unfounded.

The St. Lucie Unit 2 steam generators were replaced appropriately pursuant to 10 C.F.R. § 50.59; this was confirmed by an NRC review during an inspection of steam generator replacement with no adverse findings. The EPU of St. Lucie Unit 2 with the replacement steam generators was accomplished pursuant to a license amendment, during which the tube wear at St. Lucie Unit 2 was considered and evaluated in light of the unique tube-to-tube wear at SONGS Unit 3. FPL and AREVA, as well as the NRC Staff and the ACRS, were satisfied that the design, fabrication and operations of the replacement steam generators in St. Lucie Unit 2 were sufficiently different from SONGS Unit 3 to be less likely susceptible to tube-to-tube wear. Results of inspections of St. Lucie Unit 2 steam generators and op-8

erational assessments prepared in support of each restart after such inspections have confirmed the consensus technical view. Mr. Gundersens speculation does not ad-dress the considerable record in the EPU license amendment proceeding.

I declare under penalty of perjury that the foregoing is true and correct.

Executed in Accord with 10 C.F.R. § 2.304(d)

William A. Cross Nuclear Regulatory Programs Manager 700 Universe Blvd Juno Beach, FL 33408 Phone: (561) 561-2970 E-mail: William.Cross@fpl.com 9

Exhibit A

  • St. Lucie Unit 2 SELECTED LICENSING ISSUES (SLI)

OPERATING LICENSE RENEWAL

  • Document No. DBD..SLI-OLR-2 Revision 1 Page 53 of 200 TABLE 5.2-2~7 REACTOR COOLANT (System 01)

STEAM GENERATORS Component/Commodity Aging Effects Group Intended Requiring

[GALL Reference] Function Material Environment Management Program/Activity Internal Environment Pnmary heads Pressure boundary Low alloy steel Treated water- Cracking Chemistry Control Program (IV 01.1.8] with sta1nless primary ASME Section XI, Subsections IWB,IWC, and steel cladding

- 4. g~u..allfll~' ~QI T~:ate'l .. ,.\11~- C!!i~tkiF~fl' IWD lnseMce Inspection Program GR~:~~IW,' Qa~~~E~IIJFBftP&M f- Wth etei~laee ~ '""'

-'Lin">

. 1-Plimary m:;~nway covers Pressure boundary Low alloy steel Treated water - Cracking Chemistry Control Program with-AIIe'J" primary

-6QQ.Ist.alnlass steel diaphrnam Pnmary Inlet and outlet Pressure boundary Low alloy steel Treated water- Cracking Chemistry Control Program nozzles with stainless primary ASME Section XI, SubGoc!ions IWB, IWC, and steel cladding IWD lnservtce lnsp_~otion Program Primary inlet nozzle safe ends

- Pressure boundary Carbon stael Treated water- Crackmg Chemistry Control Program Plimary outlet noxzla safe with stainless primary ASME Section XI, Subsections IWB, IWC, and ends stool cladding IWD lnsarv1ce Inspection Program

'fubcsheets Pressure boundary Low alloy llteel Treated water~ Cracking Chemlstry Conlrol Program wilh Alloy..e;ee. pnmary ASME Section XI, Subsections JWB, IWC, and cladding c:.~o IWO lnserv1ce Inspection Program Low alloy steel Treated water- Crnoklng ASME Section XI, Subsections lWB, IWC, and secondary IWD lnservlce Inspection Program Loss of rna tan aI Chem1Jilry Control Program Primary instrument nozzles Pressure boundary Alloy-see- Treated water- Cracking Chem1stry Control Program (IV 01.1.10] 6cto primary Allay 600 Inspection Program ASME Section XI, SlJbsectlons IWB, IWC, and IWD lnsarvice Inspection Program lube plugs Pressure boundary Alloy 9138 Treated water* Cracking Steam Generator Integrity Program

~IVD1.2.3] Alloy 690TT prtmary ChemishyControl Program Divider plates Flow distribution Elb.ilni!O!e el!!!!) Treated water- Cracking Chemistry Control Program I I pnmal)'_ J

                • - --*rr** --::~-
  • St. Luc1e Unit 2 SELECTED LICENSING ISSUES (SLI)

OPERATING LICENSE RENEWAL

  • Document No. DBD-SLI-OLR-2 Rev1s1on 1 Pa e 54 of200 TABLE 5.2M2,7 (continued)

REACTOR COOLANT (System 01}

STEAM GENERATORS Component/Commodity Aging Effects Group Intended Requiring

[GALL Reference] Function Material Environmen Management ProgramfActivlty Internal Environment ig2ntinuedl U-tubes Pressure boundary Alloy690~ Treated water

  • Cracking Chemistry Control Program

[IV 01.2.1] Heat transfer TT primary Steam Generator Integrity Program ASME Section XJ, Subsections JWB. JWC, and IWD lnservlce Inspection Program Treated water* Loss of material Chemistry Control Program secondary Crackmg Steam Generator lnteglity Program ASME Section XI, Subsections IWB. IWC, and IWD lnserv1ce Inspection Program Upper a_nd lower shells Pressure boundary Low alloy steel Treated wale r- Cracking ASME Section XI, Subsections IW8, IWC, and

[IV 01.1.3] secondary JWD lnservice lnsoection Pro1:1ram Lo:;;s of mate rial Chemimrv Control Program Transition cones Pressure boundary Low alloy stelll Treated water- Crackmg ASME Section XI, SubsectionsiW1:3, IWC, and

[IV 01.1.4] secondary IWD !nserv1ce Inspection Program Loss of materia\ Chamlstrv Control Program Secondary heads Pressure boundary Low alloy stee I Treated watar- Cracking ASME Section XI, Subsections IWB, JWC, and

[IV 01.1.1] secondary IWD lnservlce Inspection ProQram Loss of matenal Chemistry Control Program Feedwater nozzles~ Pressure boundary Low alloy steel Treated water -

secondary Cracking ASME Section XI, Subsections IWB, JWC, and IWD lnservice Inspection Program

[IV 01.1.5! Loss of material Chemistry Control Program Flow Accelerr;~ted Corrosion Program O!ellm o~:~Yet t'leeerl~ ea~ i'RQ'W Pressure boundary Low alloy steel Treated water - Cracking ASME Section XI, Subsections IWB, IWC, and Steam outlet nozzles secondary IWD lnsef\llcc Inspection Program

[IV 01.1.2] Loss of matenal Chemistry Control Program Flow Acoelaratad Corrosion Program Slowdown nozzles Pressure boundary be 1 ellej* eleel Treated water* Cracking ASME Section XI, Subsections IWB, JWC, and secondary lWD lnservice lnsp!Jction PrpQr-.!lm Ct.tvlton Loss of matenal Chemistry Control Program S1"e.e.l Flow Accelerated Corros1on Program

_,,..,.~ ,, -*~*~-* ,,..,.., ...... *** ****~-*-***o.*~, ,.,...,.... ,.,., ,,,.,.,,, ***<ol"io*""'"~"'-*----~ ................... _ , _ _ __, _ _ ._~** ................ -,....,......,_,.,._,....~.,~.~-**-,__....., __ ...._ __ ,,,.,,_._~~~ .....**- - - * * - ,.,.,,,, r .. ,.,&.. ,, .... '" -~-*--*

  • St. Lucie Unit 2 SELECTED LICENSING ISSUES (SLI)

OPERATING LICENSE RENEWAL

  • Document No. DBD-SLI-OLR-2 Revision 1 Page 55 of 200 TABLE 5.2-2.7 (continued)

REACTOR COOLANT (System 01)

STEAM GENERATORS ComponentJCommodity Aging Effects Group Intended Requiring rGALL Referancel Function Material Environment Management PrCXlram!Activitv Internal Environment (continued)

Secondary instrument nozzles Pressure boundary Carbon steel Treated water- Cracking ASME Section XI, Subsections IWB, IWC, and secondary IWD lnservi~ Inspection Proaram I' Loss of material Chemistry Control F>rogram Secondary manway and Pressure boundary " tee!. Treated water- Chemistry Control Program I l:rot~kj~q handhole closure~_'b. .5 t-.ai~!u.rftfAI-1 secondary_

T'Jbo bundle Wnlppern and ...., ~cturaf ~;:upport Carbon steel Treated water- Loss of material Chemistry Control Program wrapper supports secondary Tube st2ppo1l h"tijco b(lre ' Stru'\ral support j Treated water- Chemistry Control Program r~:tinlarr ftul secondary C..-ctek;ll'l...~

"'\: ,....

{IVD.1.2.21 Steam Generator lnta~rity Pro~ ram Tube Stabilizers (Stakes) '~~~support Stainless Steel Treated water- Cracking Chemistry CMIJ'ol Program Alloy690 secondary R.ecr r-Gc.c l~t"*on n o1!2 (es ....

etnd en~ GQ;fS'

"-' ~ e. r c.cff o~"t f !ct"tet o.r1J qt1"t;-v;"t"ctt'~ra f&,ttt'J' Steam 0 wtle.t 1f..taie.J c. fr.\ck;,., Clte.....irfty Cctt1iol ft>oJifo.M1.

rJozl.lt. Ventwrir lhrottl;n, Alloy""'~o W'lter-re..toiAdary Lors of A.\qfUlQ( C"em;1rry ColfliraC P'"J'..-t

  • n .---rr---
  • St. Luc1e Unit 2 SELECTED LICENSING ISSUES (SLI)

OPERATING LICENSE RENEWAL

  • Document No. DBD-SLI-OLR-2 Revision 1 Page 56 of 200 TABLE 5.2-2.7 (continued)

REACTOR COOLANT (System 01}

STEAM GENERATORS ComponentJCommodity Aging Effects Group Intended Requiring

[GALL Reference} Function Material Environment Management Pro{:! ram/Activity External Environment Primary heads Pressure boundary Containment a1r Nona None reQuired

[IV 01.1.81 Low allov steel Borated water leaks Loss of material Boric Acid Wastage Program Primary manway covers Pressure boundary .QQFbeA eweJ. CORtainment air None Nona required

[IV 01.1.111 Low alloy steel Borated water laaks Loss of material Boric Acid Wastage Program Primary inlet and outlet Pressure boundary Carbon_steel - - Containmentair _None - - - - - - - - - - - - Nomneo.ulrad-------- --- -- --- ---- --------

nozzles and safe ends Low alloy steel Borated water leaks Loss of material Boric Acid Was!aga. Program Conical skirt Structurnl support Low alloy steel Con!ainment air NonA None r-.quirad Borated waterleaka Loss of material Bone Acsd Wastage~ Program Upper and lower shells Pressure boundary Low alloy steel Contaitu11ent air None None reqUired Secondary heads Transition cones Feedwaterno~es~

~*

Steam outlet nozzles~

~ nozzles Second arv closure cove ra UpPer vessel clev1ses and Structural support Law alloy steel Containment air None None required shear keys Primary instrument nozzles Pressure bound<Jry Alloy~ Containment air None None required no Borated water leaks Secondary Instrument nozzles Pressure boundary Carbon steel Containment a1r None None required Re.circu. (it1i~n ~ ~cin:w /,;;{.,., IIOU lei q~J ~d (.Q fr Blow<-lot.m no i.z(er

  • --r----. --* ... -. ... ,.,.~.

-!:1"****

Exhibit B FLORIDA POWER AND LIGHT COMPANY ST. LUCIE UNIT 2 SELECTED LICENSING ISSUES OPERATING LICENSE RENEWAL Design Basis Document DBD-SLI-OLR-2 REVISION 2

St. Lucie Unit 2 Document No. DBD-SLI-OLR-2 SELECTED LICENSING ISSUES (SLI) Revision 2 (4/10)

OPERATING LICENSE RENEWAL Page ii VII. ST. LUCIE UNIT 2 OPERATING LICENSE RENEWAL TABLE OF CONTENTS SECTION Page Title Page i Table of Contents ii 1.0 SCOPE STATEMENT OF THE ISSUE 1 2.0 LICENSE RENEWAL CRITERIA AND GUIDANCE 1 3.0 HISTORICAL

SUMMARY

1 4.0 METHODS AND POSITIONS 4 5.0 INFORMATION INCLUDED IN THIS DBD 10 6.0 CONFIGURATION MANAGEMENT 14

7.0 REFERENCES

15 TABLES 23 - 200

St. Lucie Unit 2 Document No. DBD-SLI-OLR-2 SELECTED LICENSING ISSUES (SLI) Revision 2 OPERATING LICENSE RENEWAL Page 53 of 200 TABLE 5.2-2.7 REACTOR COOLANT (System 01)

STEAM GENERATORS Component/Commodity Aging Effects Group Intended Requiring

[GALL Reference] Function Material Environment Management Program/Activity Internal Environment Primary heads Pressure boundary Low alloy steel Treated water - Cracking Chemistry Control Program

[IV D1.1.8] with stainless primary ASME Section XI, Subsections IWB, IWC, and steel cladding IWD Inservice Inspection Program Primary manway covers Pressure boundary Low alloy steel Treated water - Cracking Chemistry Control Program with stainless primary steel diaphragm Primary inlet and outlet Pressure boundary Low alloy steel Treated water - Cracking Chemistry Control Program nozzles with stainless primary ASME Section XI, Subsections IWB, IWC, and steel cladding IWD Inservice Inspection Program Primary inlet nozzle safe ends Pressure boundary Carbon steel Treated water - Cracking Chemistry Control Program Primary outlet nozzle safe with stainless primary ASME Section XI, Subsections IWB, IWC, and ends steel cladding IWD Inservice Inspection Program Tubesheets Pressure boundary Low alloy steel Treated water - Cracking Chemistry Control Program with Alloy 690 primary ASME Section XI, Subsections IWB, IWC, and cladding IWD Inservice Inspection Program Low alloy steel Treated water - Cracking ASME Section XI, Subsections IWB, IWC, and secondary IWD Inservice Inspection Program Loss of material Chemistry Control Program Primary instrument nozzles Pressure boundary Alloy 690 Treated water - Cracking Chemistry Control Program

[IV D1.1.10] primary Alloy 600 Inspection Program ASME Section XI, Subsections IWB, IWC, and IWD Inservice Inspection Program Tube plugs Pressure boundary Alloy 690 TT Treated water - Cracking Steam Generator Integrity Program

[IV D1.2.3] primary Chemistry Control Program Divider plates Flow distribution Alloy 690 TT Treated water - Cracking Chemistry Control Program primary Alloy 600 Inspection Program

St. Lucie Unit 2 Document No. DBD-SLI-OLR-2 SELECTED LICENSING ISSUES (SLI) Revision 2 OPERATING LICENSE RENEWAL Page 54 of 200 TABLE 5.2-2.7 (continued)

REACTOR COOLANT (System 01)

STEAM GENERATORS Component/Commodity Aging Effects Group Intended Requiring

[GALL Reference] Function Material Environment Management Program/Activity Internal Environment (continued)

U-tubes Pressure boundary Alloy 690 TT Treated water - Cracking Chemistry Control Program

[IV D1.2.1] Heat transfer primary Steam Generator Integrity Program ASME Section XI, Subsections IWB, IWC, and IWD Inservice Inspection Program Treated water - Loss of material Chemistry Control Program secondary Cracking Steam Generator Integrity Program ASME Section XI, Subsections IWB, IWC, and IWD Inservice Inspection Program Upper and lower shells Pressure boundary Low alloy steel Treated water - Cracking ASME Section XI, Subsections IWB, IWC, and

[IV D1.1.3] secondary IWD Inservice Inspection Program Loss of material Chemistry Control Program Transition cones Pressure boundary Low alloy steel Treated water - Cracking ASME Section XI, Subsections IWB, IWC, and

[IV D1.1.4] secondary IWD Inservice Inspection Program Loss of material Chemistry Control Program Secondary heads Pressure boundary Low alloy steel Treated water - Cracking ASME Section XI, Subsections IWB, IWC, and

[IV D1.1.1] secondary IWD Inservice Inspection Program Loss of material Chemistry Control Program Feedwater nozzles Pressure boundary Low alloy steel Treated water - Cracking ASME Section XI, Subsections IWB, IWC, and

[IV D1.1.5] secondary IWD Inservice Inspection Program Loss of material Chemistry Control Program Flow Accelerated Corrosion Program Steam outlet nozzles Pressure boundary Low alloy steel Treated water - Cracking ASME Section XI, Subsections IWB, IWC, and

[IV D1.1.2] secondary IWD Inservice Inspection Program Loss of material Chemistry Control Program Flow Accelerated Corrosion Program Blowdown nozzles Pressure boundary Carbon steel Treated water - Cracking ASME Section XI, Subsections IWB, IWC, and secondary IWD Inservice Inspection Program Loss of material Chemistry Control Program Flow Accelerated Corrosion Program

St. Lucie Unit 2 Document No. DBD-SLI-OLR-2 SELECTED LICENSING ISSUES (SLI) Revision 2 OPERATING LICENSE RENEWAL Page 55 of 200 TABLE 5.2-2.7 (continued)

REACTOR COOLANT (System 01)

STEAM GENERATORS Component/Commodity Aging Effects Group Intended Requiring

[GALL Reference] Function Material Environment Management Program/Activity Internal Environment (continued)

Secondary instrument nozzles Pressure boundary Carbon steel Treated water - Cracking ASME Section XI, Subsections IWB, IWC, and Recirculation nozzles and end secondary IWD Inservice Inspection Program caps Loss of material Chemistry Control Program Secondary manway and Pressure boundary Stainless Steel Treated water - Cracking Chemistry Control Program handhole closure diaphragms secondary Tube bundle wrappers and Structural support Carbon steel Treated water - Loss of material Chemistry Control Program wrapper supports secondary Tube support plates and anti- Structural support Stainless Steel Treated water - Cracking Chemistry Control Program vibration bars [IV D.1.2.2] secondary Steam Generator Integrity Program Tube Stabilizers (Stakes) Structural support Stainless Steel Treated water - Cracking Chemistry Control Program Alloy 690 secondary Steam Outlet Nozzle Venturis Throttling Alloy 690 Treated water - Cracking Chemistry Control Program secondary Loss of material Chemistry Control Program

St. Lucie Unit 2 Document No. DBD-SLI-OLR-2 SELECTED LICENSING ISSUES (SLI) Revision 2 OPERATING LICENSE RENEWAL Page 56 of 200 TABLE 5.2-2.7 (continued)

REACTOR COOLANT (System 01)

STEAM GENERATORS Component/Commodity Aging Effects Group Intended Requiring

[GALL Reference] Function Material Environment Management Program/Activity External Environment Primary heads Pressure boundary Low alloy steel Containment air None None required

[IV D1.1.8] Borated water leaks Loss of material Boric Acid Wastage Program Primary manway covers Pressure boundary Low alloy steel Containment air None None required

[IV D1.1.11] Borated water leaks Loss of material Boric Acid Wastage Program Primary inlet and outlet Pressure boundary Carbon steel Containment air None None required nozzles and safe ends Low alloy steel Borated water leaks Loss of material Boric Acid Wastage Program Conical skirt Structural support Low alloy steel Containment air None None required Borated water leaks Loss of material Boric Acid Wastage Program Upper and lower shells Pressure boundary Low alloy steel Containment air None None required Secondary heads Transition cones Feedwater nozzles Steam outlet nozzles Recirculation nozzles Secondary closure covers Upper vessel clevises and Structural support Low alloy steel Containment air None None required shear keys Primary instrument nozzles Pressure boundary Alloy 690 Containment air None None required Borated water leaks Secondary instrument nozzles Pressure boundary Carbon steel Containment air None None required Recirculation nozzles and end caps Blowdown nozzles

Exhibit C UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, DC 20555 - 0001 July 23, 2012 Mr. R.W. Borchardt Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

FINAL SAFETY EVALUATION REPORT ASSOCIATED WITH THE FLORIDA POWER AND LIGHT ST. LUCIE, UNIT 2, LICENSE AMENDMENT REQUEST FOR AN EXTENDED POWER UPRATE

Dear Mr. Borchardt:

During the 596th meeting of the Advisory Committee on Reactor Safeguards, July 11-13, 2012, we completed our review of the license amendment request (LAR) for the extended power uprate (EPU) of St. Lucie, Unit 2, (St. Lucie 2) and the associated draft Safety Evaluation (SE).

Our Subcommittee on Power Uprates reviewed this matter in a meeting on June 22, 2012.

During these reviews, we met with representatives of the staff, Florida Power and Light Company (FPL or the licensee), and their consultants. We did not review the St. Lucie 2 spent fuel pool analysis, which is still under review by the staff. We had the benefit of the documents referenced.

CONCLUSIONS AND RECOMMENDATION

1. The FPL LAR for an EPU of St. Lucie 2 should be approved subject to the conditions imposed in the staffs draft Safety Evaluation.
2. Fuel thermal conductivity degradation (TCD) phenomena at St. Lucie 2 are addressed by the license condition that FPL maintain more restrictive operational/design radial power fall-off (RFO) curve limits.
3. The licensees action plan addresses our concerns related to further wear of the tubes in the replacement steam generators (SGs).

BACKGROUND The two unit St. Lucie Nuclear Power Plant is located on Hutchinson Island, near Ft. Pierce, Florida in St. Lucie County. Unit 2 is a 2x4 loop pressurized water reactor, designed by Combustion Engineering and licensed in 1983 to operate at 2560 MWt. In 1985, the unit was approved for a 5% stretch uprate to the currently licensed thermal power (CLTP) of 2700 MWt.

In this LAR, FPL requested approval of a power uprate of 10% above the CLTP and a 1.7%

measurement uncertainty recapture (MUR) to allow a maximum core power level of 3020 MWt.

FPL plans to implement this EPU in the fall of 2012.

There are no changes in the reactor coolant system, reactor vessel internals, and fuel type due to the EPU. The two St. Lucie 2 SGs were replaced in 2007 with AREVA Model 86/19TI SGs.

The reactor vessel closure head was replaced in 2007.

The EPU will change some core design parameters including fuel enrichment and radial peaking factor, but the maximum linear heat rate will remain the same. The core average coolant temperature will increase from 573.3°F to 578.5°F. The reactor vessel head temperature will increase from about 595°F to 604°F. Design steam mass flow rates in each of the SGs will increase by approximately 13%.

Safety-related changes include control room air conditioning, charging pump control circuits, neutron absorption materials in spent fuel pool storage racks, nuclear steam supply system setpoints, environmental qualification of electrical equipment, component cooling water supports, and the SG low-level trip setpoints. Installation of a Leading Edge Flow Measurement system reduces flow measurement uncertainty and allows a 1.7% increase in power. Some of these changes have already been implemented, and the remaining changes will be completed during the fall 2012 outage.

DISCUSSION We reviewed the staffs evaluation of the EPU effects on station blackout, component material degradation, risk, and electrical power systems. In addition, we considered the licensee's power ascension test program. Issues of special interest that arose during our review are discussed in this letter.

Fuel Thermal Conductivity Degradation (TCD)

NRC Information Notice 2009-23, Nuclear Fuel Thermal Conductivity Degradation, describes an issue concerning the ability of legacy thermal-mechanical fuel modeling codes to accurately predict the exposure-dependent fuel TCD.

The NRC-approved FATES3B fuel rod performance model used by FPL for predicting fuel centerline temperature at high burnup does not model TCD. In response to staff concerns, FPL proposed a license condition that will impose more restrictive operational/design radial power fall-off (RFO) curve limits for St. Lucie 2. The new RFO curve limits were derived by comparing FATES3B fuel temperature predictions to results from Halden fuel tests. FATES3B predictions compared well to Halden data up to intermediate levels of rod average burnup (about 35 GWd/MTU). The predictions underestimate fuel centerline temperatures at higher burnups.

New RFO curve limits were determined by imposing in the analysis a penalty that increases from 0-200°F over the burnup range from 35-50 GWd/MTU and remains constant for higher burnups. The staff compared FPL FATES3B fuel temperature predictions that incorporate this RFO curve penalty to Halden data and performed independent FRAPCON-3.4 calculations.

Based on these comparisons, the staff concluded that this was acceptable for addressing TCD phenomena at St. Lucie 2. The more restrictive RFO curve limits will be verified as part of the Reload Safety Analysis Checklist process.

Steam Generator Performance Each replacement SG contains 8999 thermally-treated Alloy 690 tubes with broached stainless steel horizontal supports and an anti-vibration bar (AVB) system. The steam generator supplier, AREVA, performed design calculations with their codes to ensure that accumulated SG tube wear was acceptable for 110% CLTP and EPU conditions.

After their first 18 month cycle of operation at the CLTP level, an inspection revealed a number of tube-to-AVB wear indications (3700 indications on 1231 tubes in SG A and 2157 indications on 815 tubes in SG B). Approximately 90% of these wear indicators were less than 15% of the tube wall thickness. Although none of the tube wear reached the 40% wear limit that would require plugging, FPL conservatively plugged the 14 tubes with greater than 25% wear (e.g., 8 tubes in SG A and 6 tubes in SG B). After their second 18 month cycle of operation, inspections found an additional 2164 indications on tubes in SG A and 804 additional indications on tubes in SG B (bringing the total number of affected tubes to 1862 for SG A and to 1125 for SG B). The measured average wear rates reduced from 7.9 to 4.0 %/EFPY for SG A and from 7.7 to 1.6%/EFPY for SG B, but one tube in SG A reached the 40% wear limit for plugging. In addition, the licensee plugged any tubes with measured wear exceeding approximately 30% (a total of 16 additional tubes in SG A and 5 additional tubes in SG B).

The licensee completed a root cause evaluation that considered factors such as SG design, manufacturing processes, materials and associated tolerances, and potential operational effects. They concluded that the root cause was that the U-tubes were not effectively supported during SG manufacture, which caused the tubes to sag into the AVBs and led to slight AVB deformation that closed the tube-to-AVB gap at specific locations. This exacerbated tube wear in those locations. Supporting information for this root cause evaluation included updated AREVA analyses with revised gap distributions that predict wear similar to observed values after the first and second inspections.

The licensees analyses indicate that the increased steam flow rates associated with the EPU will have a negligible effect on the observed tube wear rates. Results from a third full 100%

bobbin coil inspection (scheduled for this fall) will provide additional information. In addition, a full 100% bobbin coil inspection will be conducted after EPU conditions are implemented.

The licensee performed an operational assessment for the next two cycles which included a cycle under EPU conditions. This assessment was based on wear rate data from the first two inspections. The analysis applied a factor of 1.24 to the wear rates to account for the increase in wear rate due to the change in flow conditions for the EPU. The factor of 1.24 is based on an analysis with the tube and support in contact in accordance with the root cause evaluation. The assessment does not credit any additional attenuation of the wear rates that may occur during the current cycle of operation. Assessment results indicate acceptable margin against tube structural integrity requirements, indicating a probability of loss of margin of 0.02 versus an allowable value of 0.05.

The tube wear observed at St Lucie 2 is primarily at AVB supports. This is different than the form of degradation reported to have occurred at San Onofre. There are a number of design differences between the SGs installed at San Onofre and those at St Lucie 2. We reviewed the FPL evaluation of these differences and concluded that the forms of degradation reported to have occurred at San Onofre are less likely to occur at St Lucie 2. This will be verified by the inspection following the first EPU cycle.

These considerations and the licensees action plan adequately address concerns about SG tube integrity.

CLOSING COMMENT In summary, the EPU license amendment request for St. Lucie 2 should be approved with the license conditions identified in the SER.

Sincerely,

/RA/

J. Sam Armijo Chairman REFERENCES

1. License Amendment Request for Extended Power Uprate, St. Lucie, Unit 2, Docket No.

50-389, Renewed License No. NPF-16, February 25, 2011, (ML110730116).

2. Draft NRC Safety Evaluation on St. Lucie 2 EPU, updated July 2012 (ML12145A032).
3. NRC Review Standard 001 (RS-001), Review Standard for Extended Power Uprate, Revision 0, December 2003 (ML033640024).
4. CENPD-132, Supplement 4-P-A, Calculative Methods for the C-E Nuclear Power Large Break LOCA Evaluation Model, April 2001 (ML011030417).
5. NRC Information Notice 2009-23, Nuclear Fuel Thermal Conductivity Degradation,"

October 8, 2009 (ML091550527).

6. CENPD-139-P-A, Fuel Evaluation Model, July 1974, (ML120960147).
7. CEN-161(B)-P-A, Improvements to Fuel Evaluation Model, August 1989, (ML120960155).
8. CEN-161(B)-P, Supplement 1-P-A, Improvements to Fuel Evaluation Model, January 1992, (ML120960175).
9. CENPD-275-P, Revision 1-P-A, C-E Methodology for Core Designs Containing Gadolinia-Urania Burnable Absorbers, May 1988.
10. CEN-372-P-A, Fuel Rod Maximum Allowable Gas Pressure, May 1990.
11. CENPD-275-P, Revision 1-P, Supplement 1-P-A C-E Methodology for PWR Core Designs Containing Gadolinia-Urania Burnable Absorbers, April 1999.
12. Letter from B. T. Moroney (NRC) to J. A. Stall (FP&L), St. Lucie Plant, Unit 2 - Issuance of Amendment Regarding Change in Reload Methodology and Increase in Steam Generator Tube Plugging Limit (TAC No. MC1566), January 31, 2005 (ML050120363).
13. CENPD-404-P-A, Revision 0, Implementation of ZIRLOTM Cladding Material in CE Nuclear Power Fuel Assembly Designs, November 2001 (ML013270123 and 013270127).
14. CEN-386-P-A, Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/kgU for Combustion Engineering 16x16 PWR Fuel, ABB Combustion Engineering, Inc.,

August 1992.

15. CENPD-384-P, Report on the Continued Applicability of 60 MWD/kgU for ABB Combustion Engineering PWR Fuel, ABB Combustion Engineering, Inc., September 1995.
10. CEN-372-P-A, Fuel Rod Maximum Allowable Gas Pressure, May 1990.
11. CENPD-275-P, Revision 1-P, Supplement 1-P-A C-E Methodology for PWR Core Designs Containing Gadolinia-Urania Burnable Absorbers, April 1999.
12. Letter from B. T. Moroney (NRC) to J. A. Stall (FP&L), St. Lucie Plant, Unit 2 -

Issuance of Amendment Regarding Change in Reload Methodology and Increase in Steam Generator Tube Plugging Limit (TAC No. MC1566), January 31, 2005 (ML050120363).

13. CENPD-404-P-A, Revision 0, Implementation of ZIRLOTM Cladding Material in CE Nuclear Power Fuel Assembly Designs, November 2001 (ML013270123 and 013270127).
14. CEN-386-P-A, Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/kgU for Combustion Engineering 16x16 PWR Fuel, ABB Combustion Engineering, Inc., August 1992.
15. CENPD-384-P, Report on the Continued Applicability of 60 MWD/kgU for ABB Combustion Engineering PWR Fuel, ABB Combustion Engineering, Inc., September 1995.

Accession No: ML12198A202 Publicly Available Y Sensitive N Viewing Rights: NRC Users or ACRS Only or See Restricted distribution OFFICE ACRS SUNSI Review ACRS ACRS ACRS NAME WWang WWang CSantos EMHackett EMH for JSA DATE 07/23/12 07/23/12 07/23/12 07/23/12 07/23/12 OFFICIAL RECORD COPY

Letter to R.W. Borchardt, EDO, from J. Sam Armijo, ACRS Chairman, dated July 23, 2012

SUBJECT:

FINAL SAFETY EVALUATION REPORT ASSOCIATED WITH THE FLORIDA POWER AND LIGHT ST. LUCIE, UNIT 2, LICENSE AMENDMENT REQUEST FOR AN EXTENDED POWER UPRATE ML#12198A202 Distribution:

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