ML15082A074

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 Station Duke Energy CNO1VP 1 4800 Concord Road York, SC 29745 o: 803.701.4251 CNS-15-028 f: 803.701.3221 March 19, 2015 10 CFR 50.55a U.S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001

Subject:

Duke Energy Carolinas, LLC (Duke Energy)Catawba Nuclear Station, Unit 1 Docket Number 50-413 Relief Request Serial Number 15-CN-001, Proposed Alternative Repair for Main Steam 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. The basis 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 Station LJR/s Enclosure www.duke-energy.com U.S. Nuclear Regulatory Commission March 19, 2015 Page 2 xc (with enclosure):

V.M. McCree Regional Administrator U.S. Nuclear Regulatory Commission

-Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 G.A. Hutto, III NRC Senior Resident Inspector Catawba Nuclear Station G.E. Miller (addressee only)NRC Project Manager (Catawba)U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 8 G9A 11555 Rockville Pike Rockville, MD 20852-2738 Enclosure Duke Energy Carolinas, LLC Catawba Nuclear Station, Unit 1 Relief Request Serial No. 15-CN-001 Proposed Alternative Repair for Main Steam System Braided Flex-Hose, Submitted Pursuant to 10 CFR 50.55a(z)(2)

Relief Request 15-CN-001 Page 2 of 5 1.0 ASME Code Component Affected Catawba 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 psia Design Temperature:

600°F Pipe Size and Material:

1/2"-Sch. 80 / SA376 TP304 Flex-Hose Size and Material:

1/2"-2500# / SA213 TP304 Root 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 is located 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 the Main Steam 1A Containment Penetration (M113) identified as Item No. 91 within UFSAR Table 6-77, Unit 1 Containment Isolation Valves Data.2.0 Applicable Code Edition and Addenda ASME Boiler and Pressure Vessel Code, Section Xl, 1998 Edition with the 2000 Addenda 3.0 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 that the standards of IWB-3522 may be applied. Duke Energy has chosen to apply the acceptance standards of IWB-3522 to address leakage from this Class 2 component.

3.2 IWA-4400 specifies requirements for welding, brazing, defect removal, and installation 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 Boundary provides the requirements for using and designing mechanical clamping devices for piping pressure boundary.4.0 Reason for Request 4.1 On August 12, 2014 a steam leak was discovered from a braided flex-hose part of level instrument 1 SMLS-571 0. This level switch is located in the Unit 1 Exterior Doghouse and controls valve 1SM-89 which dumps accumulated condensate to the Unit 1 main condenser as required.

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

-The structural integrity of the ASME Code Class 1, 2, and 3 components shall be maintained." Once isolated it was discovered that minor leakage still occurred from the flex-hose due to valve Relief Request 15-CN-001 Page 3 of 5 seat leakage past the upstream root valves. To enable isolation of the leak to facilitate a code repair/replacement of the leaking flex-hose in accordance with IWA-4000, the alternative documented in this request is proposed.

The proposed alternative consists of monitoring the leak until corrective action is required, then installing a mechanical clamping device using ASME Section XI, Appendix IX and injecting sealant into the pipe between the two root valves to fully isolate the leaking component.

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

Because Appendix IX prohibits the use of mechanical clamping devices on portions of piping systems that form the containment boundary, relief is required to permit use of this mechanical clamping device to facilitate the isolation of these valves to replace the leaking flex-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 rate tested. 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 primary side immediately after a LOCA occurs. Any leakage between the primary and secondary 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 pressure against 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 Basis Accident.

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

The presence of this steam leak does not invalidate any Safety Analysis calculations with regards to SGTR (i.e., an unanalyzed condition does not exist).4.3 NRC Inspection Manual, Part 9900 Technical Guidance, Appendix C.12 Operational 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 in accordance with the intent of the ASME Code or construction code and, therefore, would not meet code requirements, even though the system or component may demonstrate adequate structural integrity." The guidance provided in Part 9900 implies that the NRC does not accept that IWC-3000 of the ASME Code, Section Xl allows through wall leakage in Class 2 components.

Since, a through-wall flaw in a flex-hose cannot be evaluated using an applicable and NRC endorsed code case, relief is required to comply with this guidance.5.0 Proposed Alternative and Basis for Use 5.1 Proposed Alternative In lieu of the requirement of IWB-3522.1 to correct the degraded condition prior to 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 5 activity 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 during Operations rounds to confirm that the leakage from the flex-hose has not increased significantly.

2. If a significant increase in leakage is detected, a non-code repair shall be performed to stop the leakage using a mechanical clamp, 1/8" NPT injection valve, and injection sealant. An ASME class 2 mechanical clamp shall be installed followed by installation of an ASME class 1 injection valve. After installing the injection valve, a 3/16" diameter hole shall be drilled in the pipe to facilitate sealant injection.

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

A drawing of the mechanical clamp is provided in Attachment

4. Additionally, after verification that the leak has been fully isolated, 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. These caps (or plugs) will also serve as part of the class 2 pressure boundary until the flex hose, affected root valves and piping can be replaced.3. The mechanical clamp, injection valve and sealant injection may be used between one or both sets of root valves located upstream and downstream of the 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 are available to isolate this component are not leak-tight and are located upstream of the main steam isolation valves (MSIVs). Therefore, in order to isolate the affected component, a unit shutdown would be required to facilitate 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 to depressurize the line and replace the affected components.

Compliance with the 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 the leak identifies any significant increase in leakage rate, the proposed clamp and leak injection alternative (and installation of temporary pressure boundary 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 the leaking component to be fully isolated and depressurized to permit removal of the leaking flex-hose and installation of pipe caps (or plugs) to restore the leak-tight integrity of the system. The piping between the leaking root valves has been evaluated and the structural and leak-tight integrity of this piping shall 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 5 6.0 Duration of Proposed Alternative The proposed alternatives to the ASME Code are applicable for the third 10-year Inservice Inspection (ISI) Interval at Catawba Nuclear Station, Unit 1.* The Catawba Unit 1 third Inservice Inspection Interval began on June 29, 2005 and 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 before refueling outage 1EOC22.7.0 References 7.1 1998 Edition through 2000 Addenda, ASME Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components." 7.2 US NRC Regulatory Issue Summary 2005-20, Rev. 1, Revision to NRC Inspection 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 into ADAMS.These drawings, can be accessed within the ADAMS package or by performing a search on the Document/Report Number.DOI -D03X Attachment I Flow Diagram of Main Steam System (SM)Coordinates E-1 Attachment 2 Instrumentation Weld Isometric Drawing Attachment 3 Weld Isometric Drawing Attachment 4 Mechanical Clamp Outline Drawing INST (1) 1/8 NPT INJ PT CENTERED IN CLAMP MACHINE .110 DP CRUNCH TEETH THRU BORE AS SHOVWN WITH 03/16 THRU 4,00 CNS-f 2 :CNS-15-0209 UNLESS OTHERWISE SPECIFIED, ALL DIMENSIONS IN INCHES MACHINED SURFACES -/-BREAK SHARP CORNERS TOLERANCES:

3 PLACE DECIMAL !,005 2 PLACE DECIMAL e.01 I PLACE DECIMAL ,1 ANGLES 11/2 FRACTIONS

-1/32.PMA J0224J wf" -3.96 LBS 1l; I -.16 CUIH vmý DM 1/225/15 1/2' HOT TAP BAR UTLITIS SUPPRT SPECIALIST UTILITIE INC.P.O. BOX 338 VALDESE, NC 28690-0338 PH, 704-327-8744 A I !l2.00--l I'll-- -I ----, -, ý, -=A J 3/4 HOLE'5/8 STUD (2 PLCS)3-1/8' LONG MATERIAL.PLATE: PIPEt EARS, STUDS: NUTSs SA-516/675 CR70 SA-516/675 CR70 SA-193 07 SA-194 2H 1.25-1.25--2.50 SHEET I OF I