ML15082A074

Catawba Nuclear Station Unit 1, Relief Request Serial Number 15-CN-001, Proposed Alternative Repair for Main Stem System Braided Flex-Hose

ML15082A074
Person / Time
Site:	Catawba
Issue date:	03/19/2015
From:	Henderson K Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15082A073	List: CNS-15-028, Drawing CN-1593-01.07, Flow Diagram of Main Steam Supply (SM) ML15082A074
References
CNS-15-028
Download: ML15082A074 (13)

Text

Kelvin Henderson DUKE Vice President ENERGY Catawba Nuclear StationDuke EnergyCNO1VP 1 4800 Concord RoadYork, SC 29745o: 803.701.4251 CNS-15-028 f: 803.701.3221 March 19, 2015 10 CFR 50.55aU.S. Nuclear Regulatory Commission Attention:

Document Control DeskWashington, DC 20555-0001

Subject:

Duke Energy Carolinas, LLC (Duke Energy)Catawba Nuclear Station, Unit 1Docket Number 50-413Relief Request Serial Number 15-CN-001, Proposed Alternative Repair for MainSteam System Braided Flex-Hose

-Submitted Pursuant to 10 CFR 50.55a(z)(2)

Pursuant to 10 CFR 50;55a(z)(2),

Duke Energy hereby submits Relief Request 15-CN-001 requesting approval to use an alternative repair for a Main Steam System leaking braided flex hose. Thebasis for the proposed relief request is provided in the enclosure to this letter.Duke Energy requests NRC approval of this relief request at your earliest possible convenience so that the repair may be implemented.

There are no regulatory commitments contained in this relief request submittal.

If you have any questions or require additional information, please contact L.J. Rudy at(803) 701-3084.

Very truly yours,Kelvin Henderson Vice President, Catawba Nuclear StationLJR/sEnclosure www.duke-energy.com U.S. Nuclear Regulatory Commission March 19, 2015Page 2xc (with enclosure):

V.M. McCreeRegional Administrator U.S. Nuclear Regulatory Commission

-Region IIMarquis One Tower245 Peachtree Center Ave., NE Suite 1200Atlanta, GA 30303-1257 G.A. Hutto, IIINRC Senior Resident Inspector Catawba Nuclear StationG.E. Miller (addressee only)NRC Project Manager (Catawba)

U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 8 G9A11555 Rockville PikeRockville, MD 20852-2738 Enclosure Duke Energy Carolinas, LLCCatawba Nuclear Station, Unit 1Relief Request Serial No. 15-CN-001 Proposed Alternative Repair for Main Steam System Braided Flex-Hose, Submitted Pursuant to10 CFR 50.55a(z)(2)

Relief Request 15-CN-001 Page 2 of 51.0 ASME Code Component AffectedCatawba Nuclear Station, Unit 1 ASME Class 2 Level Instrument Flex-Hose.

The following information is applicable to this component:

System: Main Steam (SM)Design Pressure:

1200 psiaDesign Temperature:

600°FPipe Size and Material:

1/2"-Sch. 80 / SA376 TP304Flex-Hose Size and Material:

1/2"-2500#

/ SA213 TP304Root Valves: 1/2"-1500#

/ SA182, Gr. F316 / globe valve (Model 09J-574)The flex-hose is part of level instrument 1SMLS-5710.

This degraded flex-hose islocated downstream of the two stainless steel root valves on each side of the flex-hose as detailed in Catawba Unit 1 Instrument Weld Isometric drawing CNI-SM-1571 (Attachment

2) and Weld Isometric drawing CN-1SM-0082 (Attachment 3).Instrument 1SMLS-5710 is located within the containment isolation boundary of theMain Steam 1A Containment Penetration (M113) identified as Item No. 91 withinUFSAR Table 6-77, Unit 1 Containment Isolation Valves Data.2.0 Applicable Code Edition and AddendaASME Boiler and Pressure Vessel Code, Section Xl, 1998 Edition with the 2000Addenda3.0 Applicable Code/Regulatory Requirements 3.1 IWC-3516 specifies that acceptance standards for Examination Category C-H,All Pressure Retaining Components are in the course of preparation, and thatthe standards of IWB-3522 may be applied.

Duke Energy has chosen to applythe acceptance standards of IWB-3522 to address leakage from this Class 2component.

3.2 IWA-4400 specifies requirements for welding,

brazing, defect removal, andinstallation of pressure retaining items.3.3 IWA-4133 provides alternative requirements for repairs using mechanical clamping devices using Mandatory Appendix IX.3.4 The ASME Boiler and Pressure Vessel Code,Section XI, Appendix IX,Mechanical Clamping Devices for Class 2 and 3 Piping Pressure Boundaryprovides the requirements for using and designing mechanical clampingdevices for piping pressure boundary.

4.0 Reason for Request4.1 On August 12, 2014 a steam leak was discovered from a braided flex-hose partof level instrument 1 SMLS-571

0. This level switch is located in the Unit 1Exterior Doghouse and controls valve 1SM-89 which dumps accumulated condensate to the Unit 1 main condenser as required.

The flow diagram andweld isometric drawings showing this configuration are shown in Attachments 1, 2, and 3. The leaking ASME Class 2 flex-hose was isolated to comply withCNS SLC 16.5-5 "Structural Integrity

-The structural integrity of the ASMECode Class 1, 2, and 3 components shall be maintained."

Once isolated it wasdiscovered that minor leakage still occurred from the flex-hose due to valve Relief Request 15-CN-001 Page 3 of 5seat leakage past the upstream root valves. To enable isolation of the leak tofacilitate a code repair/replacement of the leaking flex-hose in accordance withIWA-4000, the alternative documented in this request is proposed.

Theproposed alternative consists of monitoring the leak until corrective action isrequired, then installing a mechanical clamping device using ASME Section XI,Appendix IX and injecting sealant into the pipe between the two root valves tofully isolate the leaking component.

Following sealant injection, the leaking flex-hose shall be removed and replacement pressure boundary material (pipe capsor plugs) shall be attached to the downstream side of the root valves to whichthe flex-hose had been connected.

Because Appendix IX prohibits the use ofmechanical clamping devices on portions of piping systems that form thecontainment

boundary, relief is required to permit use of this mechanical clamping device to facilitate the isolation of these valves to replace the leakingflex-hose.

4.2 As defined in UFSAR Table 6-77, "Unit 1 Containment Isolation Valve Data",the Main Steam containment isolation valves are not required to be leak ratetested. This is due to the Main Steam line being connected to the secondary side of the steam generator which is kept at a higher pressure than the primaryside immediately after a LOCA occurs. Any leakage between the primary andsecondary sides of the steam generator is directed inward to containment (i.e.,main steam header pressure is maintained higher than peak containment pressure).

This penetration is effectively sealed by steam header pressureagainst leakage from containment after a LOCA.This Main Steam 1A Header leak has been evaluated against consequence during the postulated Steam Generator Tube Rupture (SGTR) Design BasisAccident.

It was concluded that post SGTR radiation doses for this scenariowould be bounded by those for the SGTR with a failed open Power OperatedRelief Valve (PORV) on the ruptured Steam Generator.

The presence of thissteam leak does not invalidate any Safety Analysis calculations with regards toSGTR (i.e., an unanalyzed condition does not exist).4.3 NRC Inspection Manual, Part 9900 Technical

Guidance, Appendix C.12Operational Leakage From ASME Code Class 1, 2, and 3 Components, states"The NRC staff does not consider through-wall conditions in components, unless intentionally designed to be there such as sparger flow holes, to be inaccordance with the intent of the ASME Code or construction code and,therefore, would not meet code requirements, even though the system orcomponent may demonstrate adequate structural integrity."

The guidanceprovided in Part 9900 implies that the NRC does not accept that IWC-3000 ofthe ASME Code, Section Xl allows through wall leakage in Class 2components.

Since, a through-wall flaw in a flex-hose cannot be evaluated using an applicable and NRC endorsed code case, relief is required to complywith this guidance.

5.0 Proposed Alternative and Basis for Use5.1 Proposed Alternative In lieu of the requirement of IWB-3522.1 to correct the degraded condition priorto continued

service, Duke Energy requests NRC approval to allow continued operation until such time that an ASME Code, Section Xl repair/replacement Relief Request 15-CN-001 Page 4 of 5activity can be performed in accordance with IWA-4000.

The following alternative requirements are proposed:

1. Surveillance of the flex-hose leakage shall be performed once per shift duringOperations rounds to confirm that the leakage from the flex-hose has notincreased significantly.

2. If a significant increase in leakage is detected, a non-code repair shall beperformed to stop the leakage using a mechanical clamp, 1/8" NPT injection valve, and injection sealant.

An ASME class 2 mechanical clamp shall beinstalled followed by installation of an ASME class 1 injection valve. Afterinstalling the injection valve, a 3/16" diameter hole shall be drilled in the pipe tofacilitate sealant injection.

After sealant injection is completed, the mechanical clamp and closed injection valve shall serve as part of the class 2 pressureboundary until a code repair/replacement activity complying with IWA-4000can be performed.

A drawing of the mechanical clamp is provided inAttachment

4. Additionally, after verification that the leak has been fullyisolated, the damaged braided flex-hose shall be removed and code compliant caps (or plugs) shall be installed on the end of the outboard root valves. Thesecaps (or plugs) will also serve as part of the class 2 pressure boundary untilthe flex hose, affected root valves and piping can be replaced.

3. The mechanical clamp, injection valve and sealant injection may be usedbetween one or both sets of root valves located upstream and downstream ofthe degraded flex hose to fully isolate the leakage.5.2 Basis for Proposed Alternative Hardship:

A code-compliant repair cannot be performed without fully isolating and depressurizing the affected component.

The root valves that areavailable to isolate this component are not leak-tight and are locatedupstream of the main steam isolation valves (MSIVs).

Therefore, in order toisolate the affected component, a unit shutdown would be required tofacilitate the repair. Therefore, the only way to perform a code-compliant repair in accordance with IWA-4000 would be to shutdown the unit in order todepressurize the line and replace the affected components.

Compliance withthe specified requirement would result in hardship without a compensating increase in the level of quality and safety.If the leakage from the flex-hose stabilizes, use of the proposed leak injection alternative will not be necessary.

However, if continued surveillance of theleak identifies any significant increase in leakage rate, the proposed clampand leak injection alternative (and installation of temporary pressureboundary materials) shall be implemented.

The proposed alternative to install an engineered clamp with an injection valve and inject sealant between the leaking root valves will enable theleaking component to be fully isolated and depressurized to permit removal ofthe leaking flex-hose and installation of pipe caps (or plugs) to restore theleak-tight integrity of the system. The piping between the leaking root valveshas been evaluated and the structural and leak-tight integrity of this pipingshall be maintained during the installation of the clamp, during leak injection, and during subsequent operation until a permanent repair/replacement activity can be performed.

Relief Request 15-CN-001 Page 5 of 56.0 Duration of Proposed Alternative The proposed alternatives to the ASME Code are applicable for the third 10-yearInservice Inspection (ISI) Interval at Catawba Nuclear Station, Unit 1.* The Catawba Unit 1 third Inservice Inspection Interval began on June 29, 2005and is currently scheduled to end on June 29, 2016.Use of the proposed alternative is requested until Code repair/replacement activities can be performed on the level instrument piping and flex-hose during refueling outage 1EOC22 (fall, 2015) or during a forced outage of sufficient duration beforerefueling outage 1EOC22.7.0 References 7.1 1998 Edition through 2000 Addenda, ASME Code,Section XI, "Rules forInservice Inspection of Nuclear Power Plant Components."

7.2 US NRC Regulatory Issue Summary 2005-20, Rev. 1, Revision to NRCInspection Manual Part 9900 Technical

Guidance, "Operability Determinations

& Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety".

The 3 drawings specifically referenced Enclosures 1, 2,& 3 have been processed intoADAMS.These drawings, can beaccessed within the ADAMSpackage or by performing asearch on theDocument/Report Number.DOI -D03X Attachment IFlow Diagram of Main Steam System (SM)Coordinates E-1 Attachment 2Instrumentation Weld Isometric Drawing Attachment 3Weld Isometric Drawing Attachment 4Mechanical Clamp Outline Drawing INST (1) 1/8 NPT INJ PT CENTERED IN CLAMPMACHINE .110 DP CRUNCH TEETH THRU BOREAS SHOVWN WITH 03/16 THRU4,00CNS-f 2 :CNS-15-0209 UNLESS OTHERWISE SPECIFIED, ALL DIMENSIONS IN INCHESMACHINED SURFACES

-/-BREAK SHARP CORNERSTOLERANCES:

3 PLACE DECIMAL !,0052 PLACE DECIMAL e.01I PLACE DECIMAL ,1ANGLES 11/2FRACTIONS

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