ML22151A265

Presubmittal Licensee Presentation for 3.3.2 ESFAS Proposed LAR

ML22151A265
Person / Time
Site:	Robinson
Issue date:	05/31/2022
From:	Kappopoulos E Duke Energy Progress
To:	Tanya Hood NRC/NRR/DORL/LPL2-2
Hood T
References
Download: ML22151A265 (21)

Text

Robinson / NRC Pre-submittal Meeting:

LAR to Add Feedwater Isolation on SG High-High & Remove Obsolete Methods ( ~ DUKE

~ ENERGY June 7, 2022

Duke Energy Attendees Ryan Treadway (Manager, Nuclear Fleet Licensing)

Joshua Duc (Nuclear Fleet Licensing)

Jeff Abbott (Manager, Fleet Nuclear Fuels Engineering)

Christy Ray (Fleet Nuclear Fuels Engineering)

Scott Jackson (RNP Engineering)

Brad Hearne (RNP Engineering)

Fred Lane (RNP Operations) 2 2

Agenda Current and Proposed Technical Specification (TS) / Surveillance Requirement (SR)

Steam generator overfill protection: add feedwater isolation on SG level high-high to TS 3.3.2 Revise TS 2.1.1.1 and TS 5.6.5.b to reflect obsolete analytical methods System Design and Operation Reason for the Proposed Change Justification Precedent Schedule 3

Current and Proposed TS / SR No changes, included for reference only 4

4

Current and Proposed TS / SR Note:

SR 3.3.2.1 = CHANNEL CHECK SR 3.3.2.4 = CHANNEL OPERATIONAL TEST (COT)

SR 3.3.2.7 = CHANNEL CALIBRATION 5

5

Current and Proposed TS / SR SLs 2.0 2.0 SAFETY LIMITS {SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest cold leg temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:

2. 1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained

1.141 for the HTP correlation aA~ ~ 1.17 fer the X~~B eerrelatieA.

2.1.1.2 The peak fuel centerline temperature shall be maintained < [4901 -

(1.37 x 10-3 x (Burnup, MWD/MTU))] °F.

6 6

Current and Proposed TS / SR TS 5.6.5, Core Operating Limits Report (COLR) 7 7

Current and Proposed TS / SR Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements (continued)

9. XN NF 621(A), "XNB Critical Heat Flu)( Correl.ation,ff approved version as specified in the CO.L R

• Replace text in 10. Deleted items 9 and 11 with "Dell eted" 11. XN-NF~B2 06~t\). "Qualification of Exxon Nuolear Fuel for Extended Bumup," approved version as specified in the COLR.

• i '

12. Deleted

13. Deleted 8

8

Current and Proposed TS / SR 5.6 Reporting Requirements (comlnued) 5.6.5 CORE OPERATING LIMITS REPORT (COlRI (continued)

14. Deleted

15. Deleted I

16. ANF 88 054(P), 'PDC 3: l\,d,,aAeee Nuslear Fuels C..lp<>l'atioA Pe"er OistnbuVaA Callfrol for P , e s s ~

A!ipHeetien ef POC 3 te H.B. RelliASOA UAit2,' appro,*ed '"BFSkJA .a&

speeified IA the OObRc

17. ANF 88 133 (PJ{A), "Q\Jallliaalloo of. AllvaAGe<I NuGl&a<-filel8'-l>WR DeeigA Mel~edele9)' ror Red Bumups er e2 G"fil!Mll I,' appR11<ed version as specified iMl>e CGbft

18. ANF 89 161 (AJ, ' A"lF REI.AP Methadel"!l'f ror i2re&S1mad \llratar ReaeleFS: l<.Aelysis el NBA bOC.A.CllaJ>ler 15 E*,ame," approved Re place text in VBF6iOA-B&-specified-i~

items 16, 17, 18,

19. EMF 9i! Q81(AJ, 'SteliS!isal SetpelAt,/TraRaiellt l(ett,gdQlogy !Qr 19, 21, 22, and 23 Westingliouse-+ype-Reactors" app,eved *,ersieA as spedfied IA with "Deleted " the-00\cR, .

20. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiffng ConelaUon for High Thermal Performance Fuel," approved version as specified in the COLR.

• 21 . XN NF 86 92(PJ!A), "EJGcoo-NIIGlear Uranitim DioxideJGadolmla lrradiatio~-Examlna~OA*and-Themlal-Gollductivlty-Reaulls,'

e~re~ed versieA ae ~ l ,r lhe COL.R

22. EMF 96 929(P)(I'<), "Reaeter A~l~s System ro, PWR&."

approVed*V81SK>IH 1&-speeffiedaiFHl,e,.CObR,

23. EMF 9i! 116, *oerierie t.1eetiaAlsal'Oesi!!R Clileria ror f'IA~ ~u..1 Designs." appre*,ed veraleA as spesillell iR the COi R

24. EMF-2103(PXA), "Realistic Large Break*LOCA Methodology tor Pressurtted water Reactors." approved version as spedfled in the COLR.

9 9

Current and Proposed TS / SR l.i.6 Reporting Requtemenl$ loonllnued) 5.6.5 CORE OPERATING LIMITS REPORT (COLRJ (oon!i!lued)

25. e:MF 231G(P)~6,), "8ft° Chapter 16 NeA bOC,tr. Mettleeele!l) fer

. Replace text in Pfess~12e!I W,aler Reaete'9," a~pre,all ,al'l!iel'I eupeemed il'I IM item 25 with CQlA.-

"Deleted"

26. BAW-10240(P)(A), *1ncorporatron !)fl/15 Properties In Framalome ANP ,'\pproved Methods," approved version as specified in the COLR. .

. 27. EMF-2328(P)(A), "PWR Small Break LOCA Evaluation lllodel, S-REt.AP5 Based." approved version as specified In the COLR.

28. DPC-NE-20CJ5.P-A, "Thermal,.Hydraulfc Stalistlcal Core Desl:gn Methodology," approved versJon as specified iri the COLR.

29. DPC-NE-1008-P-A, "Nuclear Design Methodology Using CASM0-5/SIMULATE-3 for Wminghouse Reactors." as approved by NRC Safety Evaluation dated May 18, 2017.

30. Df'C-NF-2010-A, "Nuclear Physics Methodology for Reload Design,".as apl)fbved by NRC Safely Evaluation dated l/lay 18, 2017.

81. Df'c-NE*2011-P-A. "Nuclear Design Mettiodology Report for Core Operating Limits of Westinghouse Reactors* as approved by NRC Safety Evaluation dated May 18, .2017. *

32. DPC-NE-300a..J>..A, "lhemial-Hydrauoc Models for Transient Analysis," as approved by NRC Safety E~aluation dated Apfil 10, 2018 . .

33. DPC-NE-3009-P-A. "FSAR / UFSAR Chapter 15 Tl'an9ient All81ysis l/lethodalogy," as approved by NRC Safety Eva1u11~on dated

• April 10, 2018.

34. BAW~10231P-A, .COPERN IC Fuel Rod Design Computer Coch!," 10 a p ~ version as specified.in the OOLR. 10

System Design and Operation Robinson (RNP) is Westinghouse-designed 3-loop pressurized water reactor (PWR)

Main Feedwater (MFW) system supplies water to Steam Generators (SGs)

Main flow path: Main Feedwater Isolation Valve (MFIV) and Main Feedwater Regulation Valve (MFRV)

Bypass flow path: two bypass valves FW Isolation on SG Level High-High at 75% level (SG overfill protection)

Trips main turbine and MFW pumps Closes MFRV and bypass valves on the affected SG 2 out of 3 channel initiating logic that is safety related Not part of Engineering Safety Features Actuation System (ESFAS) 11 11

System Design and Operation

Background

Generic Letter (GL) 89-19 SG overfill protection design is sufficiently separate from MFW control system TS include periodic verification of operability of SG overfill protection system NRC closeout letter for RNP dated June 24, 1991 (ML14184A839)

RNP conversion to standard TS (STS) - safety evaluation dated October 24, 1997 (ML020560172, ML14175A922, ML14175A924)

FW isolation on SG high-high not included in new ESFAS TS Table 3.3.2-1

- RNP STS submittal: This Function is not classified as an Engineered Safety Feature in the plant design basis and current licensing basis.

RNP UFSAR 15.1.2, Feedwater System Malfunctions that Result in an Increase in Feedwater Flow Currently not an explicit analysis, bounded by UFSAR 15.1.3 and 15.4.1 events Does not acknowledge that feedwater must be terminated to prevent SG overfill Approved via safety evaluation dated November 7, 1984 (ML020520268) 12 12

Reason for the Proposed Change SG overfill can occur within ~5 minutes of event initiation if MFW is not isolated Concerns documented in GL 89-19 still valid and could challenge fission product barrier (1) the increased dead weight and potential seismic loads placed on the main steam line and its support should the main steam line be flooded; (2) the loads placed on the main steam lines as a result of the potential for rapid collapse of steam voids resulting in water hammer; (3) the potential for secondary safety valves sticking open following discharge of water or two-phase flow; (4) the potential inoperability of the main steam line isolation valves (MSIVs), main turbine stop or bypass valves, feedwater turbine valves, or atmospheric dump valves from the effects of water or two phase flow Restore compliance with GL 89-19 Correct non-conservative TS to ensure UFSAR Chapter 15 Condition II event does not transition to more serious Condition IV event (Condition II acceptance criteria)

Remove analytical methods no longer in use due to approval of new Duke Energy methods 13 13

Justification UFSAR 15.1.2 reevaluated using NRC-approved Duke method DPC-NE-3009-P-A MFRV to one SG failed open, increasing feedwater flow FW isolation upon reaching uncertainty adjusted SG high-high setpoint Turbine trip would also concurrently occur but is conservatively not modeled in this new analysis Significant margin to Departure from Nucleate Boiling (DNB) and Centerline Fuel Melt (CFM) acceptance criteria Peak primary and peak secondary pressures bounded by UFSAR 15.2.2 event Allowable value of 76.16% calculated consistent with other ESFAS setpoints Process measurement accuracy, sensor calibration accuracy, sensor and rack drift, the effects of temperature and pressure on the racks and sensors, etc.

Below maximum reliable indicated level (MRIL) of Westinghouse NSAL-02-4 14 14

Justification Proposed MODES are 1, 2, and 3 Except in MODES 2 and 3 when all MFIVs, MFRVs, and associated bypass valves are closed or isolated by a closed manual valve, as feedwater isolation would be satisfied in this condition In MODES 4, 5 and 6, the MFW system and the turbine generator are not in service Consistent with STS NUREG-1431 Utilize existing Condition D Specific to channel operability and applies to functions with 2 out of 3 logic Existing Completion Times more conservative than STS NUREG-1431 (72 hrs) 15 15

Justification SRs 3.3.2.1 (Channel Check), SR 3.3.2.4 (COT), SR 3.3.2.7 (Channel Calibration)

Frequency controlled by the Surveillance Frequency Control Program Consistent with other existing Table 3.3.2-1 ESFAS channel-based protection functions NUREG-1431 specifies SR 3.3.2.1, SR 3.3.2.5, SR 3.3.2.9, and SR 3.3.2.10 Proposed SRs are consistent with SR 3.3.2.1, SR 3.3.2.5, and SR 3.3.2.9 SR 3.3.2.10 to verify ESFAS response times is not applicable to RNP, as it was not included in the approved RNP conversion to STS 16 16

Justification GL 89-19: sufficient separation between MFW control and overfill protection All 3 level indication channels used for both overfill protection and MFW control (median selector)

Signal isolation devices for separation of protection and control Power supplies through different circuits off the same instrument buses Power to instrument buses is reliable source with alternate supply available The 3 channels do not share common routing 2 protection channels remain operable in the event of a fire in MFW control system 17 17

Justification Methods approved to allow Duke Energy in-house analysis for RNP DPC-NE-2005-P, Thermal-Hydraulic Statistical Core Design Methodology, safety evaluation March 8, 2016 (ML16049A630)

DPC-NE-1008-P, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors, safety evaluation May 18, 2017 (ML17102A923)

DPC-NF-2010, Nuclear Physics Methodology for Reload Design, safety evaluation May 18, 2017 (ML17102A923)

DPC-NE-2011-P, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors, safety evaluation May 18, 2017 (ML17102A923)

DPC-NE-3008-P, Thermal-Hydraulic Models for Transient Analysis, safety evaluation April 10, 2018 (ML18060A401)

DPC-NE-3009-P, FSAR / UFSAR Chapter 15 Transient Analysis Methodology, safety evaluation April 10, 2018 (ML18060A401)

BAW-10231P-A, COPERNIC Fuel Rod Design Computer Code, safety evaluation April 29, 2019 (ML18288A139)

The methods and DNB correlation limit proposed for deletion are no longer planned for use in RNP design/safety analysis Administrative change to remove obsolete information 18 18

Precedent GL 89-19 required response from licensees in early 1990s Turkey Point added SG overfill protection, safety evaluation dated April 28, 1994 (ML013380368)

Similar 3-loop Westinghouse PWR with 2 out of 3 SG overfill logic Stated in TS and TS Bases that SG overfill is not part of ESFAS Remove obsolete methods Harris safety evaluation dated April 8, 2021 (ML21047A470)

Based on the same Duke Energy in-house methods Harris safety evaluation dated March 30, 2012 (ML12058A133)

RNP safety evaluation dated December 29, 2011 (ML11342A165) 19 19

Schedule Submit LAR by June 30, 2022 Implementation within 120 days of receipt of safety evaluation 20 20

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