ML20099F241

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Co., NRC Safety Pre& 8208;Application Meeting North Anna Power Station (NAPS)
ML20099F241
Person / Time
Site: North Anna  Dominion icon.png
Issue date: 04/09/2020
From:
Dominion Energy Co
To:
Office of Nuclear Reactor Regulation
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ML20099F240 List:
References
Download: ML20099F241 (22)
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Text

NRC Safety PreApplication Meeting North Anna Power Station (NAPS)

April 9, 2020 1

North Anna Overview

  • Two Westinghouse 3loop PWRs
  • Net Capacity: Each unit is
  • ~800 MWnet => 2940 MWt
  • Sister Plant to Surry Power Station
  • Located in Louisa County, VA OLs 40 Yrs 60 Yrs 80 Yrs Unit 1 1978 2018 2038 2058 Unit 2 1980 2020 2040 2060 2

Plant Aerial

Discussion Topics

  • Same experienced team using same methodologies from Surry
  • Insights from Surry SLR were incorporated into the North Anna SLRA to the extent practical
  • Commonality in plant designs and many Fleet AMPs
  • TLAAs address first LR TLAAs and extensive review performed for new TLAAs 4

Integrated Plant Assessment

  • NEI 1701 methodology utilized throughout IPA process
  • North Anna methodology was consistent with Surry:

Revalidation of various design inputs for inscope determinations (a)2 methodology consistent with current standards and expectations

  • Higher degree of AMR consistency with GALLSLR:

> 99% Consistency (only 11 of 7495 lines with FJ Notes) 60 Note E Lines (plant specific AMPs and/or AMP exceptions)

  • 10Year OE Search identified no new aging effects GALLSLR comprehensive
  • FER items requiring plant specific review resulted in aging management considerations 5

New SLR AMPs

  • XI.M32 OneTime Inspection
  • XI.M33 Selective Leaching
  • XI.M35 ASME Code Class 1 Small Bore Piping
  • XI.E3B Inaccessible Instrument and Control Cables Not Subject to 10 CFR 50.49
  • XI.E3C Inaccessible LowVoltage Power Cables Not Subject to 10 CFR 50.49
  • XI.E6 Electrical Cable Connections Not Subject to 10 CFR 50.49
  • XI.E7 High Voltage Insulators 6

AMPs With Exceptions

  • XI.M3 Reactor Head Closure Stud Bolting Yield strength and ultimate tensile strength (SPS)
  • XI.M20 OpenCycle Cooling Water System Recirculation spray HT EX test interval (SPS)
  • XI.M27 Fire Water System Fire pump suction strainer Main drain test frequency (SPS)
  • XI.M29 Atmospheric Metallic Storage Tanks Caulking/sealant and concrete missile shield (SPS)
  • XI.M42 Internal Coatings/Linings (ISGTBD) 7

AMPs With Enhancements

  • 12 consistent with GALL no enhancements
  • 13 minimal enhancements:

Scope and/or standard GALL language enhancement 4 or less enhancements with straight forward guidance to be incorporated

  • 11 many enhancements:

>4 enhancements to be incorporated more involved enhancements to be incorporated 8

Incorporation of Revisions to GALLSLR

  • AMPs XI.M2 & XI.M21A EPRI Chemistry Guidelines
  • AMP XI.M12 Thermal Aging Embrittlement
  • AMP XI.M16A PWR Vessel Internals (MRP227R1A)
  • AMP XI.M42 Internal Coatings/Linings
  • AMP XI.E7 High Voltage Insulators
  • AMR items: Table 3.51
  • AMR items: Generic FJ Notes (15 items) 9

Flaw Tolerance Evaluation for RCS CASS

  • Flaw tolerance evaluation is consistent with Grimess Letter and NUREG 2191 Section XI.M12, Thermal Aging Embrittlement of CASS
  • Delta ferrite for check chemistry is less than 25% and evaluated with ASME Section XI Appendix C or ASME Section XI Code Case N838, as applicable, based on delta ferrite content
  • The Unit 1 crossover leg delta ferrite content per Hulls Equivalent Factors for Ladle and Check chemistry conditions is 24.57% and 28.29%.
  • Delta ferrite for one ladle value greater than 25% is evaluated using Category 2 weld Zfactors from Table C63301 of the 2019 Edition of Section XI. Use of the Category 2 weld Zfactor provided in Table C6330 1 of the 2019 Edition of Section XI for delta ferrite greater than 25% is a more stringent penalty factor compared to the SAW Zfactor for delta ferrite between 14% and 25%.
  • A Plant Specific Evaluation will be included for assessment of the fitting with ladle value greater than 25% delta ferrite.

10

Reactor Internals AMP - FER 3.1.2.2.9 80 Year Reference basis for developing and defining the aging management of PWR reactor vessel internals components:

  • MRP227 R1A (3002017168) PWR Reactor Internals Inspection and Evaluation Guidelines including NRC Safety Evaluation and associated RAI responses (ML19339G364)
  • MRP 2018022, Interim Guidance for PWR Internals I&E Guidelines, MRP227A, for SLR: Westinghouse and Combustion Engineering (ML19081A061)
  • Recent & related Industry operating experience:

Lower Girth Welds: MRP 2019009 (ML19249B102)

CRGT Sheaths and CTubes: WCAP17451 Rev 2 (ML19262E593)

  • dhd 11

Reactor Vessel (RV) Support Steel Configuration 12

Irradiation of Concrete: FER 3.5.2.2.2.6

  • Fluence Projections:

Concrete biological shield (CBS) fluence projections at 72 EFPY are below the fast neutron (E> 0.1 MeV) fluence threshold of 1x1019 n/cm2 and the gamma dose threshold of 1x108 Gy (1x1010 rad)

[Westinghouse Proprietary Report]

  • CBS Concrete Maximum Temperature:

The maximum temperature in the CBS concrete including radiationinduced heating is less than FER 3.5.2.2.2.2 limit of 200oF for local areas and 150oF for general areas with sufficient margin.

(Gamma heating reference: EPRI TR 3002013051) 13

Irradiation of RV Support Steel

  • NAPS Reactor Vessel Support is through a Neutron Thermal Shield configuration like Surry
  • Dominion will use fracture mechanics (NUREG1509) to demonstrate structural integrity through the SPEO.
  • Some differences between Surry and NAPS:

Loading on NST for SPS based upon Framatome Break Opening Time (BOT) previously analyzed for evaluation of the reactor vessel slide foot support.

Loading on NST for NAPS will be based upon original BOT which is overly conservative.

Fracture toughness for SPS was based upon KIR/KIa curve while the fracture toughness for NAPS will be based upon the KIc curve 14

Reactor Vessel Integrity

  • Materials Properties Confirmation Material properties reviewed and updated (PWROG18005 NP) to ensure consistency with ASME Code & BTP 53 Generic Rotterdam USE, Cu, Ni values for Rotterdam Welds and Forgings (PWROG17090)
  • LTOP LTOP enable temperature is 269F - System is enabled per Technical Specification LCO 3.4.12.
  • Pressurized Thermal Shock (PTS)

RTPTS screening criteria values satisfied through SLR (72 EFPY) 15

Reactor Vessel Integrity

  • Upper Shelf Energy (USE)

USE values are above 50 ftlbs at 72 EFPY except for Unit 1 inlet nozzle forging 11 equal to 50 ftlbs at 72 EFPY, and Unit 2 intermediate forging 04 less than 50 ftlbs at 72 EFPY Equivalent Margins Analysis completed by PWROG 19407 for USE items and will be submitted for NRC review and approval.

  • Heatup and Cooldown Curves Confirmed that existing PT curves are acceptable through 72 EFPY
  • Surveillance Capsule Withdraw Schedule Letter 19390 submitted November 25, 2019 requesting NRC review and approval to amend the SC withdrawal schedules.

Early withdrawal dates correspond to the first RFO after when the capsules will reach a fluence equal to 80 years of fluence For SLR, potential withdraw dates correspond to the first RFO after when the capsules will reach a fluence equal to 100 years of fluence.

Reference:

WCAP18363NP and WCAP18364NP 16

EAF - Components & Vessels

  • ASME Code Section III components and vessels
  • NB3200 evaluations reduced CUF and Fen multiplier so CUFen below unity CRDMs Reactor Vessel inlet nozzle, outlet nozzle and CETNA Pressurizer locations Steam Generator locations Reactor Coolant Pump

EAF Piping

  • Reactor coolant loop (RCL) piping was designed to USAS B31.7 Piping classified into thermal zones/transient sections Retained all NUREG 6260 locations Fen methodology from NUREG/CR6909 One material is not used to screen out another material Common basis stress evaluation (EPRI 1024995) used for one zone Uen based upon design transients for one zone with use of Code Case N779 (Ke) and Code Case N902 (EPRI 3002014121 - thickness and gradient factors)

Uen based upon design transients in one zone using 1979 Edition of B31.7

  • Fatigue for USAS B31.7 piping will be managed by Fatigue Monitoring program (Appendix L - 4 locations)
  • Revised Pressurizer Surge Line weld inspection for Initial License Renewal submitted to NRC in March 2020
  • Appendix L inspection (10 year frequency) for weld that attaches the reactor coolant hot leg nozzle to the pressurizer surge line piping

Other Plant Specific TLAAs

  • 3 PWROG Reports with NRC Safety Evaluations generically address TLAAs RCP Flywheel Fatigue Crack Growth Analysis (PWROG17011NPA Rev 2 ML19198A056)

Cracking Associated With Weld Deposited Cladding (PWROG17031NPA Rev 1 MLTBD)

Reactor Coolant Pump Code Case N481 (PWROG17033NP/PA Rev 1 - ML19266A666)

Documents To Be Docketed With SLRA

  • PWROG18005NP, Revision 2, Determination of Unirradiated RTNDT and UpperShelf Energy Values of the North Anna Units 1 and 2 Reactor Vessel Materials, September 2019
  • WCAP18015NP, Revision 2, Extended Beltline Pressure Vessel Fluence Evaluations Applicable to North Ann Units 1 & 2, September 2018
  • WCAP18363NP, Rev. 1, North Anna Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation, March 2020
  • WCAP18353NP, Rev 0, Reactor Internals Fluence Evaluation for a Westinghouse 3Loop Plant with Two Units, October 2018
  • WCAP18364NP, Rev. 1, North Anna Units 1 and 2 TimeLimited Aging Analysis on Reactor Vessel Integrity for Subsequent License Renewal (SLR), March 2020
  • WCAP11163P, Revision 1, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for North Anna Units 1 and 2 Nuclear Power Plants for the Subsequent License Renewal Program (80 Years) LeakBeforeBreak Evaluation, October 2019 20

Documents Previously Docketed

  • WCAP14040A, Rev. 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004 (ML050120209)
  • PWROG19047P, North Anna Reactor Vessel UpperShelf Fracture Toughness Equivalent Margin Analysis (MLTBD)
  • PWROG17011NPA, Revision 2, Update for Subsequent License Renewal:

WCAP14535A, Topical Report on Reactor Coolant Pump Flywheel Inspection Elimination and WCAP15666A, Extension of Reactor Coolant Pump Motor Flywheel Examination, January, 2019 (ML19198A056)

  • PWROG17031NPA, Revision 1, Update for Subsequent License Renewal:

WCAP15338A, A Review of Cracking Associated with Weld Deposited Cladding in Operating PWR Plants, May 2018 (MLTBD)

  • WCAP15338A, A Review of Cracking Associated with Weld Deposited Cladding in Operating PWR Plants, October 2002 (ML083530289)
  • PWROG17033NPA, Revision 1, Update for Subsequent License Renewal:

WCAP13045, 'Compliance to ASME Code Case N481 of the Primary Loop Pump Casings of Westinghouse Type Nuclear Steam Supply Systems,' November 2019 (ML19266A666) 21

Closing Remarks

  • North Anna is a sister plant to Surry - Many Fleet AMPs
  • Dominion is highly experienced with SLR
  • Dominion has been engaged and integrated with the development of GALLSLR ISGs and industry guidance
  • North Anna SLRA has incorporated Surry RAIs
  • North Anna SLRA will have a higher degree of consistency with GALLSLR
  • Dominion will submit a higher quality application to support an efficient NRC review
  • The North Anna SLRA is ahead of schedule with expected submittal in 3Q20 22
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