HNP-10-097, Relief Request from ASME Boiler and Pressure Vessel Code, Section XI Requirements for the Service Water System

Relief Request from ASME Boiler and Pressure Vessel Code,Section XI Requirements for the Service Water System

ML102740038
Person / Time
Site:	Harris
Issue date:	09/24/2010
From:	Caves J Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
HNP-10-097
Download: ML102740038 (16)

Text

Serial: HNP-10-097 10-CFR 50.55a jProgress Energy SEP 2 4 2010 U.S. Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET-NO. 50-400/RENEWED LICENSE NO. NPF-63 RELIEF REQUEST FROM ASME BOILER AND PRESSURE VESSEL CODE, SECTION XI REQUIREMENTS FOR THE SERVICE WATER SYSTEM Ladies and Gentlemen:

In accordance with the Code of Federal Regulations, Title 10, Part 50.55a, "Codes and Standards," paragraph (g)(5)(iii), the Harris Nuclear Plant (HNP) of Carolina Power and Light Company, doing business as Progress Energy Carolinas, Inc., submitsthe following request for relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2001 Edition with addenda through 2003.

Approval is requested for deferral of code repair of a flaw in an ASME Code Class 3 piping supply line in the HNP Service Water (SW) system. Slight moisture accumulation on "A" Train Emergency Service Water (ESW) supply pipe 3SWI-1 14SA-l indicates a leak point at the interface of a sockolet and sockolet-to-pipe weld. The flaw is located in a section of piping that cannot be isolated to complete a code repair within the time period permitted by the applicable Technical Specifications (TS) Limiting Condition for Operation (LCO).

Code repair of the identified flaw at this time is impractical since a plant shutdown would be required. Evaluation of the flaw in accordance with the fracture mechanics methodology provided in GL 90-05 has determined that the structural integrity of the SW piping is not adversely affected by this flaw. Therefore, HNP requests NRC approval to defer implementation of code repairs to no later than the next scheduled refueling outage, as permitted by GL 90-05.

The attached relief request addresses the present condition of the weld and implementation of the compensatory actions taken per GL 90-05. Operability and functionality of the system have been maintained and HNP has concluded that deferring repair of the flaw will not affect the health and safety of the public. Since compliance with the specified Code requirements would result in unnecessary hardship without a compensating increase in the level of quality and safety, HNP requests approval of this relief request pursuant to 10 CFR 50.55a(g)(5)(iii).

Progress Energy Carolinas, Inc.

Harris Nuclear Plant P.0. Box 165 New Hill, NC 27562 47

HNP-10-097 Page 2 contains proposed HNP relief request 13R-08. contains the Regulatory Commitments associated with this request.

Please refer any questions regarding this submittal me at (919) 362-3137.

Sincerely, J.R. Caves Supervisor - Licensing/Regulatory Programs Harris Nuclear Plant JRC/kab

Enclosures:

1. 10 CFR 50.55a Request: 13R-08

2. List of Regulatory Commitments cc: Mr. J. D. Austin, NRC Sr. Resident Inspector, HNP Ms. M. G. Vaaler, NRC Project Manager, HNP Mr. L. A. Reyes, NRC Regional Administrator, Region II

Enclosure Ito SERIAL: I-TNP-10-0971 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a REQUEST: 13R-08 Revision 0 Request for Relief for Temporary Non-Code Repair of Service Water Return Piping Line in Accordance with 10 CFR 50.55a(g)(5)(iii) 1.0 ASME CODE COMPONENT AFFECTED (a)

Description:

Interface of a sockolet and sockolet-to-pipe weld on line 3SW1-114SA-1, a 1-inch carbon steel supply line to root valve 1SWf62 and a downstream pressure transmitter connection point off of the "A" Train Emergency Service Water (ESW) return piping from the "A" Component Cooling Water Heat Exchanger (CCW HX).

(b) Function:

The ESW system provides cooling water to remove heat from essential plant heat loads associated with reactor auxiliary components for dissipation in the plant ultimate heat sink during emergency operation. The Operability of the ESW System ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions.

Line 3SWI-1 14SA-1 must maintain its structural integrity so that minimum flow can be maintained on all "A" ESW loads.

(c) Class: ASME Section III, Class 3 (d) Description of the flaw:

A through-wall pinhole leak was found on the carbon steel sockolet and sockolet-to-pipe weld interface on pressure transmitter connection line 3SWI-1 14SA-1 off the "A" Train ESW return piping to the "A" CCW HX. Only slight moisture accumulation can be seen at the leak point with no measurable leak rate. The "A" ESW header is supplied by Normal Service Water (NSW) with an operating temperature of 1000 F and operating pressure of approximately 80 psig.

The 1-inch piping and downstream valve were most recently replaced during RFO- 13 (April 2006); however, the sockolet itself was not replaced during that outage. Non-destructive (NDE) ultrasonic testing (UT) was conducted on the affected area. The thickness data was reviewed by HNP's Mechanical/Civil Design group. The data shows that the areas surrounding the pinhole are near nominal thickness values. The thickness values of the sockolet vary, but were all Page 1 of 13

Enclosure Ito SERIAL: HNP-10-097 SHEARON HARRIS NUC.LEAR POWER PLANT, UNITNO. 1Y DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50:55a REQUEST: 13R-08 Revision 0 recorded'to'b'greater than 0.200:inyches: T..s thickness readings ofthe:l -inch i K piping adjacent tothe sockolet were recorded near nominalvalues. No areas of generalized thinning were identified.

(e) - Flaw Detection:

C.

The flaW was identified: on August, 8, 2010,:during operator rounds., The plant was in Mode 1 at 100 percent power.

2.0 APPLICABLE CODE EDITION AND ADDENDA ASME Boiler and Pressure Vessel Code;,Section XI, 2001 Edition with addenda through 2003.-

3.0 APPLICABLE CODE REOUIREMENT Per NRC Inspection Manual. Part 9900. Technical Guidance, "OperabilityDeterminations

& Functionality, Assessments for Resolution of Degraded or Nonconforming Conditions

'Adverse to;Quality orSafety,, Section C.12, "If aleak is:discovered in a Class 1, 2, or 3 component while conducting an in-service inspection, maintenance activity, or during facility operation, any corrective measures to repair or replace the leaking component must be performed in accordance with IWA-4000 of Section XI"..

'Article IWA-4000(Repair/Replacement Activities) provides the requirements for performing repair/replacement activities on corponents and the-r:supports. This is used

,.whenever a flaw. is discovered that does not.meetthe ASME~requirements.:

Per IWA-4 l0:ofiIW.A-4000 (Scope): ., "  :

(a) The requirements of this Article apply regardless of the reason for the repair/replacementactiy.ity or the method that detected:the condition requiring the repair/replacement activity.

(b) This Article provides requirements for repair/replacement activities associated with pressure retaining Components and their supports,4ncluding appurtenances,.

subassemblies, parts of a component, core support structures, metal containments and their:integral, attachments, and metallic portions of Class CC containments, and their integral attachments. Repair/replacement activities include welding, brazing, defect removal, metal removal by thermal means;, rerating, and removing, adding,,and-Page 2of 13'

'4 "'"' 'y' Enclosure Ito SERIAL: HNP-10-097 SHEARON HARRIS NUCLEARPOWER PLANT, UNIT NO. I DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50i55a:REQUEST: 13R-08 Reyision 0 modifýing i items or, systems.-: Thesexre'quirements are applicable-to procurement, design, installation, examination, and pressure testing of items withinjthe scope of%,this Division.

HNP is requesting relief from these Article IWA-4000 requirements to defer the code repair of the identified through-wall flaw until the next refueling-outage;,RFO (November 2010), provided the conditions of Generic Letter (GL) 90-05, "Guidance for Perfoming* Temporary Non-Code Repair*of ASME Code. Class 1, 2, and 3 Piping," are met.

4.0 IMPRACTICALITY OF COMPLIANCE Per GL 90-05, an ASME Code repair is required for. Code Class 1, 2, and 3 piping unless specific written relief is granted by the NRC. Relief is appropriate when performing the repair at the time of discovery is determined to be impractical.

In accordance with this GL, impracticality. is definedto exist if:.

.: The flaw detected during plant-operation is in a section of Class 3piping that cannot

. .beisolatedto complete a code repairswithin the time period permitted by-the limiting condition of-operation of the affected system as specified in the plant Technic'al Specifications, and .. -.

, Performance of code itepair necessitatIs a plant shutdown...'

The identified flaw -isa-piItnhole' leak on the sockolet fitiing :atlthe interfacd of a sockolet

-and sockoletýto-pipeweld on* 3SW1'- 114SA-I1 the 'supply liie't0 root-valve;1 SW-62 and a downstream pressure 'transmitter connection point. 'This 1-,inch 6onnection line'is off of the "A" Train ESW return piping to the "A" CCW HX. The HNP Technical Specifications (TS) Limiting Condition for Operation (LCO) associdted twith "the-ESW System is:

'3,7.4i:- At least two independent emergeriny service water loops 'shall 'be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION,: :With-only one emergency service water loop-OPERABLEE restore at least two loops to OPERABLE'status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or'be in at least HOT, STANDBY within the next:6-hours and in COLD SHUTDOWN -withinthe.following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Repair of the defect requires -header.depressurization. The section~of piping containing the flaw cannot be isolated from the system without a plant shutdown. Freeze sealing of Page 3.3of 13

Enclosure I to SERIAL: HNP-10-097 SHEARON.HARRIS NUCLEAR POWER PLANT, UNIT NO., 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a'REQUEST: .13R-08 Revision :0

.. '.theline' is an, alternative method-to teriporarily isolate the line;' h6wever'it is.iunlikely that "this method canbe executed to; allow completion ofthe'repair within the':72-hour LCO

....,time window. Since. there is. little' confidence, that the affected section of piping' can be isolated :for the completion of a code repair withih the.time period permitted by the above TS; impracticality exists in accordance. with the above GL 90-05.,definition.

5.0 BURDEN CAUSED BY COMPLIANCE In orderto comply with ASME Code requirements and the applicable LCO, the plant

_would need to be shutdown to perform the repair. As noted in .GL.90-05, "The rather

'frequent-instances of small leaksin some Class 3 systems' such as service water systems, could lead to an excessive number of plant start-up and shutdown cycles.with undue and unnecessary stress on facility systems and components if the facilities were to perform a

-,code repair when the leakage is identified.". .';

6.0 PROPOSED ALTERNATIVE'AND BASIS FOR USE-*":,

In accordance with the guidelines of GL 90:05,,I-fNP is proposing'to defer repair of the identified. flaw; leaving the'piping "as,-is',, until the next outage -exceeding 30' days, which is expectedto be'the.:next refuelingoutage, RFO,16,.scheduled-to begin in October 2010.

To ensure ,that the.acceptance criteria-of GL 9-05 continue to be 'met, HNP'has.,

implemented compensatory actions to: detect changes in the condition of the, identified

'defect., ,.

6.1 SCOPE. ...

An indicationt of a.through-wall leak was found on a' sockolet fitting'at the interface between a sock*olet and socket weld. on 3SWI.-l!14SA- 1, fthe ýsupply line to, root valve 1SW-62 and a downstream pressure transmitter connection point,. This, 1-inch line is off of the "A" Train ESW return piping to the "A" CCW HX. The flaw on this carbon steel, ASMECodep Class..3 piping was discovyeredon August 8, 2010,, during normal.plant

• .:operations-.and isbelieved.,to= have'originatedon the inner dianmeter,(ID) of the affected piping. .

6.2 SPECIFIC CONSIDERATIONS ,, .

The following considerations have been 'made to 'assure,the structural integrity. of the affected piping: '- -

Page. 4 of 13

Enclosure I to SERIAL: HNP-10-097 SHEARONTHARRIS NUCLEARTPOWER PLANT, UNIT NO. #1 DOCKET-NO. 50'400/RENEWEDLICENSE NO. NPF-'63 10 CFR 50.55a REQUEST: 13R-08 Revision 0

. onsequences of'Floodng.'A wakdown.,of line 3SWl414SA-'1 wth'the. A" CCOWHX C

.--in-service an'd-NSW supplying "' ESW revealed little-to no vibration pr&sent.on: the line that could be felt by touch. This line. ext~nds: horiz0ntally'frpm 24-inch line 3 SW24-

• 67SA-1..; The :1-inch-line is less than -4foot -in length and is therefore quite rigid. The 24-inch line is' supported by a .large strudtuisal support'approximately 6 feet above the 1-inch line. Based on this, vibration induced fatigue cracking of the line, and consequential flooding of the area, is not a concern.

.. . ,- .. . . 5 - * .- -,. ,- 1.,

Consequences of Spraying Water on Other Equipment. With only slight moisture accumulation at the leak point and-no, quantifiable leak rate (less than 1 drop per minute),

the-leak is not affecting any other equipment important'to ýafety. in the immediate 'area. If the leakage rate were to increase, housekeeping devices, could be installed to shield adjacent equipment.'

Significant of Loss of Flow. The current loss of flow-from the ESW system is negligible compared to the total system flow. The leak rate is less than 1 drop per minute under normal operating conditions of 1004F and 80 psig, with NSW supplying the "A" ESW header. ESW system flow is typically.. 3,000,to.!j8,000-gpm.. '

-Design Loading Evaluation. A:stress analyysis (Attacrhment 2) was performed by HNP

. , Mechanical/Civil Engineering-that evaluated whether-the UT-thickness data for the affected area.impacted the'design- load stresses, including deadweight Pressure, thermal expafision and seismic loads of the affected piping:1This'analysis concluded that because

!: the: current pipe thickness of the, affected: aiea -was at least 90 percent of the nominal pipe -

thickness, there is a negligible effect on the calculated stresses for the pipe and no impact on the integrity of the piping system.

Ihtegrity of the Temporary Non-code Repair. A code repair will be completed n6later -

than the next refueling outage, RFO-16, scheduled to begin in October 2010. RFO-16 anrd the: subsequent repair of this flaw, will: begin withiný 90 day's ýof thediscovery, of this leak; eliminating the need to schedule additional UT nieasu~ements of the affected area every three months."(- ""'

' - v . - '- , - -'. . ,, -

...' - - - - . . 2*

..* ') ) '"J  :' ',. * , * .' .'; :" ' *[ *. 5 '

Q.u'alitativeAssessments.to Monitor*Degradation.A compensatory' action has been initiated for Operdt-ionsý to perform aqualitatiVe assessment-via-waikdown inspections of the leakage from the flaw to be completed at least weekly, meeting the GL 90"05 requirements. .

EngineeringEvaluation. UT data shows the areas surrounding the pinhole'are near nominal thickness values. The thickness of the actual sockolet varies, although all measurements were greater than-0.-200 inches..-. The 1-inch pipe adjacent-to the sokolet was recorded with thicknesses near nominal values. No general area thinning was Page 5 of 13

Enclosure Ito SERIAL: HNP-10-097 SHEARON HARRIS' NUCLEAR POWER PLANT; UNIT NO. 1' DOCKET NO. 50-400/RENEWED LICENSE NO. ,NPF'-63 10 CFR 50.55a' REQUEST: -13R-08 Revision 0 identified: HtNP Engineering Evaluation has determined that this'gis.:ikelV an igolated occurrence -withno remedial measures required.: .'

6.3 'CAUSE OFLEAK , '.* ,'.  : '.

The pinhole is on a sockolet fitting at the interface of the sockolet and sockolet'-to-pipe weld. While the 1-inch piping and downstream valve were: replaced during RFO-13 (April 2006), the sockolet itself was notreplaced. NDE-(UT)'measurements were taken for the'-affected area. of pipe.The thickness data:wasreviewed by. HNP's,. ,-,- . ,

Mechanical/Civil Design Engineering group. The data shows that .the areas surrounding the pinhole are-near nominal thickness. The thickness values of the sockolet vary, but were all greater than. 0200 inches. The -thickness readings for.the.J-inch!piping adjacent to the sockolet-wererecorded near nominal thickness. No areas of generalized-thinning were identified- . , - ...- .. ... .

Although theeexact, cause of the weld defect cannot be determined, it could be a result of impurities, work practices, orworkmanship.-Welds and' their surrounding heat affected zones are -particularly.susceptible to local corrosion attack: ..Sufficient:-selective attack in these regions results in oxide removal, thereby enhancing local corrosionrate-and the potential for formation of a crevice. Once a crevice forms and becomes wetted by service water its passive film starts to break down and the surrounding region begins to corrode in a mechanism similar to pitting corrosion. This type of -orrosio;'ltends to-remain -'

localized and not propagate rapidly into adjoining regions. Because of the leak location, it is believed that a crevice formed'on the ID of-the:weld area:due to 'localized corrosion.

With no areas. of generalized thinning ideitified, this flaw has been determined to be an isolated occurrence:- ' -~ ' - , -'..

6.4 STRUCTURAL INTEGRITY OF LINES NDE (UT) measurements were taken -n the; affected area byý a qualified H4NP: Quality Control (QC) inspector on August 9, 2010, and reviewed by HNP Mechanical/Civil Design Engineering. - .

The UT data (Attachnment 1) shows that the pipe thickness in,theiarea surrounding the pinhole is near nominal thickness. The thickness of the actual sockolet varies, however all measurements 'v'ere recorded greater. than 0.200 inches. Near nominal. thickness values were recorded for the UT readings on the 1-inch pipe and the 24-inch pipe adjacent to the sockolet, with no general area thinning identified. ,  :

Page 6 of 13

Enclosure I to SERIAL: HNP-10-097 SHEARONHARRIS NUCLEAR,;POWER PLANT,, UNIT NO. 1 DOCKET NO.-50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50,:55a'REQUEST: I3R-08 Revision 0 Aý Flaw. Eyaluation. (Section 6.5), performed in accordancewith the:GL 90-05 ;"Through-Wall Flaw Approach," demonstrates-that the piping,conmection containing the pinhole leak is structurally adequate.

A walkdown of line 3SWI-1 14SA-1 with the "A" Component.ýCo6ling Water fleat Exchanger in service and NSW supplying "A" ESW revealed little to no vibration present on the line (i.e. could not be feltiby touch). vTheiline'extends horizontally from24-inch line 3SW24-67SA-1. This. 14-inch line is less..than one foot in length and is~therefore quite rigid. Additionally, the 24.inch line~issupported by a large structural suppdrt-located approximately' 6 feet; above the 1-inch line. Therefrie, vibration induced fatigue Wear of the-line is'not a concern, .

Based on the-above evaluation;, this leak does not represent an operability concern for the structural integrity of the 1-inch supply line:3SW1 -114SA-1 for all modes of operation.

Besides monitoring, no remedial measures are required to ensure integrity is maintained.

Further analysis to consider the impact of a complete line break is unwarranted.

Furthermore; the identified pinhole leak does not prevent the ESW. system from

.perforniing'its-safety function to provide cooling: water and remove heat -from essential

, .plantheat loads: associated with reactor auxiliary components during emergency

.operation. ..

6.5 FLAW EVALUATION'

-Per GL 90-05, "'thrdugh-wall;.flaw evaluation criteria were used to evaluate the pinhole

-leak in ESW.line.3SWI-I 14SA-1. The GL:90-05 criteria are applicable sincethis line is ASME Section III, Class 3 piping per the HNP Engineering Database (EDB). For conservatism purposes, analyses assumes the through-wall flaw length "2a" to be in the circumferential direction and the stress "s" is assumed to be bending stress.

The pipe properties for line 3SW1-1 14sA-1, (1-inch, schedule 80, ASTM A106 Grade B i'pipe)per EDB, NAVCOPipinig ýDatalog & ASME are:

Outside Diameter: D,= 1.315 in Nominal Pipe Thickness::, -. ,tnom 0.1 79 in

-Measured Pipe Thickness,: , t= 0.179in. (which-is greate" than tmin)

Design Temperature: . T-.440VF Design Pressure: p = 150 psig Page 7. of 13

Enclosure Ito SERIAL: HNP-10-097 SHEARONHARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKETNO. 50-400/RENEWED LICENSE NO.,NPF-63 A10CFR 50.55aREQUEST: 13R-08 Revision ,0 Allowable Stress: , '.S 000;psi ,;ýr.,."

Allowable Stress Intensity Factor: K-<35kSi(in).

Minimum Required.Wall Thickness, t,ý fo" hoop:stress per ASME Section III is:

tmin - --- 0.007 in

- .2(S+0.4p)

"Through -Wall" Flaw 'Evaluation 2 The following through-wall flaw evaluation is performedin' accordance with GL 90-05, , Section C.3.a. The stress intensity factor for through-wall flaw (including safety factor of 1.4) is:

Stress Intensity Factor: K 1.4 (s)(F)(3.1416*a)0s, where:

Combined Bending Stresses: s = MA + MBE = 1617 psi (per Stress Evaluation)

Moment StresS(Deadweight & Pressure)K  :

MA 757.ps" Moment Siress(DBE Seismic): MDBE -86 psi.

Geometry Factor: F- 1 (A)(c)1:5 c 3.736

'A,= 3.26543 ,+.1.52784r,- 0.072698r, +o0.0016011r 3 =-627.031._',

3 =

B = 11.36322.-73,91412r + 0.18619 r 2 -0.004099r 1.-601.x 103 C = -3.18609 + 3.84763 r 0.18304r 2 +. 0.00403r 3 = 1.583 x 103 a

c(3.1416)(R)

~ 0.028 R = mean pipe radius = Do: tnom - 0.568 ir.

2 R

r S- tm in... - 86.733 in ., . . ." . .. * .._ ,: . .'

Flaw Length, 2a = the diameter of the pinhole = 0.100 in a = 0.050 in Page 8,of 13

Enclosure Ito SERIAL: HNP-10-097 SHEARONUHARRIS NUCLEARPOWER PLANT, UNlT.NO. -1 DOCKET.NOU. 50-400/RENEWED LIC-ENSE"NO. NPF-63 10 CFR 50.55a REQUEST.: 13R-08 Revision.O ,

Flaw Length, 2a, was conservatively'rassunied to be 0:100 inches.,A flaw of this dimension would have a visible, steady leakage rate. With no visible leakage rate and only moisture accumulation as in'dication of this through-wall flaw, aflaw;Iength'.of 0.100 inches is acceptable. Since "2a" is less than 3 inches and less than 15 percent of the pipe circumference-', this, "Through-Wall' Flaw" approach ilsacceptablein assessing the structural integrity of the flaw.

K = 1.4 (s)(F)(3.14i6*a)O.s = 3352 psi(in)°.5 3.352 ksi(in) 0 :s5 . .

Since the evaluated stress intensity factor of 3.352 ksi(in)0 ,5 is less,'than the allowable stress intenfsity facto6r0'o 35 ksi(in)°5 for carbon steel, this flaw meets the GL 90-05 criteria for proposal of a temporary non-code repair of the Clas ý3 piping.

6.6 AUGMENTED INSPECTIONS "

Since the flaw has been evaluated and found acceptable by the G- 90-05 "Through-Wall FlaW" approach, augmented UT inspections of the five most susceptible (and accessible) locations were performed to assess the ov,erall degradation of the affected system, per GL 90-05 requirements for moderate energy systems. The locations selected. for the augmented inspection were irithe'ih'at affected zones of welds near five robt valves on 1-.

inch, carbon steel ASME Class Code 3 service water piping with design temperature and pressure equal to that,ofthe flaw-under evaluation: As-shown in Table I below, none of the measured thicknesses,, tmin, were less than the ASME, code .required minimum.

thickness of0.007 irnches. The average wall thickness for ea'chlication was nearnominal thickness. No significant thinning was identified at any location, indicating that the flaw.

is most likely a'solated failure and that"ssem wide degradation is not present.

' Tible 1. Augmented UT Ins'ection Dat'a' -

Valve ,D tnominal tmin taverage 1SW- 1058 0.179 0.148 0.179 1SW- 1056 0.179 0.159 0.174

'1SW- 1057 0.179 0.-162:- 0.176 -

1SW-1065 0.179 0.175 0.182 1SW-1067 0.179 0.171 0.177

6.7 CONCLUSION

The flaw evaluation performed for the pinhole flaw demonstrates that the piping, including the weld joining the pipe to the sockolet for line 3SWl-1 14SA-1, is structurally adequate in Page 9 of 13

Enclosure 1 to SERIAL: HNP-10-097 SHEARON HARRIS NUCLEAR POWER PLANT,; UNIT NO. 1 DOCKET NO.- 50-400/RENEWED: LICENSE NO. 'NPF-63 10 CFR50.55a R.EQUEST:;'13R-08 RevisionD0 accordance with the guidance -provided-in GL"90-,05.: The.flaw wasý evaluated using the through-wall flaw fracture mechanics methodology provided by NRC Generic Letter 90-05, which is a conservative approach.to, evaluating the stability of the flaw. -

Since the flaw in line 3SWI-I 14SA-1 satisfies the criteria presented in GL 90-05, it is acceptable to~propose..a: temporaD', non-cdde repair of this code Class 3 piping.

Therefore, HNP is requesting NRC approval per 10 -CFR 50:55a(g)(5).(iii) to defer ASME Section XI IWA-4000 repair/replacement requirements forlthe identified flaw in accordance with the guidance provided inGL 90-05. .

7.0 DURATION OF PROPOSED ALTERNATIVE Repair of'the defect will be deferred until the next refueling outage, RFO-1 6, provided the condition continues to meet the acceptance criteria of Generic Letter 90-05. HNP is currently monitoring the leak location. RFOA16 is scheduled to begin in Qctober 2010.

8.0 PRECEDENT Similar requests for relief were approved for:

Harris Nuclear Plant, October 30, 2009, ML093010584 South Texas Project Unit 2, November 30, 2007, ML073120446

!, * .,i *, :'*: * " ...

4*U Page, 10 of,13

Enclosure I to SERIAL: HNP-10-097 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT.NO- )I DOCKET NO. 50-400/RENEWED LICENSE NO. NP.F-:63 10 CFR 50.55a REQUEST: 13R-08

'Revision,0.

Attachment l. Inspecetiofi'Findifng(for Fiaw ,on"Line,3SW1-114SA-1 CQAUT4-Rev 7 m ejrss~~ JG NDE REP-21tA Flan' ICR3. PAHNP L)N QR*KP WO:: LL j,ý Uniblý(1: [22 03 Daile ofokbi.-

GomilrpenlteitemIsleJ NDE Procoduro L!j%I A~s~ k 7.No.: L-11 Rev .: TR- 5 ccmponent Iit~a:

~C'S~1Ls~ ~ 0ona Speco 1lAina IkanyL~P.t~

1hldt' ~  !AA I.~e , NQ.ILtý1 M~~~:~~~A~~ YPeG~Ot q.?ZSU&rIe4 GaIibfai-d, Ree~~ %W,.:. [- i,[ II,, ri,.t.,ij! MiN Ma . 9 j Pi riaay -Dal.Tliinkiýa;:O-(~~ .O' 0 Sep~Io1~hl:~SiS'O C~ L~sn Uti,. O~eai- ei~~, . Ca!. uli'ei ý Otý#'esvciyf Te;,,,,Iuer..

MqPpO1mellroc f.,1hl tl' 1 Rh 4 T1¶rZ.

7- ~ ~L..... ri,4,10 row-t D bft Hibh3ntnp:, *.0 YeR - Ne- I U a-Ccat Lj""~~1.ge-4c 0

Coatedi Paint-e.d: NoI Sketch ocmparknt or item Mid ~~raoteoled. kidade k~c~ date:

• i~Z1' 00-PM'-(~ 6r&.-

J+

Oceificatian Level Da7te RsvvzSwAO d'r Titis Dam NGGr0M,P,,1,Jl11 APPENDIX/'

Page 11 of1r3 1)

Enclosure 1 to SERIAL: HNP-10-097 SHEARON HARRIS NUCLEAR POWER PLANT,. UNIT NO. 1 DOCKET NO. 50-:400/RENEWED LICENSE NO.: NPF-63 10 CFR 50.55a REQUEST:,. 13R-08 Revision'0 Attachment 2. Stress Evaluation for -Line 3SWI-114SA-1 prepared by:. Daryl Hughes 8/9/10 Verified by: Aaron Borudotsky 8M910 Stress evaluation: . .

The pin hole leak is in socket weld of the 1"pipe, 3SW1-1 14SA-1, to the coupling on the header pipe, 3SW24-67SA-1, near the outlet of the A7 CCW Heat-Exchanger. The_ 1. pipe is associated with valve ISW-62 (Ebasco tag number 3SW-V56SA- 1), refelrence drawings 5-G-0047 &5-S-0547 and stress calculation 80150-4. Since-the flaw is in the weld a 90-05 evaluation will be performed.

The 1' pipe is a short ca tleverttansiions totibinbg The 1I pipe is not iod*led in the stress analysis- However, from page 55 of calculation 8050-4. theseismic accelerations at node 1784 (represents the location of the 1I pipe to the header pipe) are:

Gx = 0.863 Gy =0.770 Gz =0.831 Pipe 3SW1-1 14SA-1 is 1', schedule 80, material ASTM A-106 per EDB_ The design pressure and temperature are 150 psig and 140 degF. respectively. The pipe properties per NAVCO Piping Datalog, Edition 1I are.

OD =1.3157 Wall thickness =0.179" 2.4b -.

Pipe weight including water 2-48 Ill 0.207 bAn Moment of Inertia = 0.105 in4 3 Section Modulus of 1' OD pipe = 0-160 in Cantilever length L =7- as-built measurement SIF=.2-1 I Calculate Bending rr*ient at the socketl let c:nection,"refere-ne AISC:

. ue to pipewe.ght= wl.2' -={.207 x (7?/2 =5.1inl Mduetovalveweight=Pxb-b=16x4.57==2i.72 b . . -

Total Bending Momerit due to lg = 5.1:+72 = 77.1 in--lb Bending stress due to DW = 77.1/0.160 = 482 psi

2. Calculate Pressure stress = pd/4t = 150 psi x 1.315/(4 xO.179) = 275 psi

3. Calculate pipe frequency based on deflection, reference AISC:

a = 2_5 Valve WL 16 #

L-W b = 4.57 1= 7' Page 1.2 of 13

Enclosure Ito SERIAL: HNP-10-097 SHEARON HARRIS NUCLEARPOWER PLANT, UNIT NO..:I DOCKET:NO. 50-400/RENEWED LIC-ENSE NO :NPF-63 10 CFR:50.55a.,REQUEST: .13R-0:8

.Revision0O Att;chmnent-2 (Continued) 4 2 A (w0L 8E) +-Pb I6EI (3L-b)

A= (((0270 (7f)4 8(29 x 10f)0.05"] ÷ [16 (4.5f 16(29 x 10 4)0.1051(3 x7-4.5)

A =O0.0O28" f=(lI2rr)? (386-41. A)!i f 187 HZ Since.-1tpe is rg, use accelerations-at node -1748 fromn 05o-4,:page 55 to, determine bending.stress in the 1 pipe.

Gx = 0.863g (Noflti) Gy =0-770g Gz =0_831g

4. Calculate DBE Stew (Emergency condilion):- -

Pipe is running nortt/south. Gx does not need to be included. The resultant acceleration for bending stress is:

Gr = (0.7702 +0.831) m = 1.133g DBE bending Stress = (77-1 x{1 75x2_1 /0160).x 1.133.=860 psi

.J 5- Total stress = Pressure + DW + DBE = 27'5-- 482 & D60"-1617 psi The thickness of the pipe at D, 90, 180 and 27D degrees are 0-191, 0-174', 0D226" and 0.162", reslpectiel using U measurements technique. A generalscan of the pipe.

showed thickes ranging frm"0.16W"to 0.175" The wall thickness, in genera3, is 90% or greater than nominal- Therefore, no inrcrease in stress is nece§sary- The stress value above is used in the attached GU' 90-05 calcuti*oins -

", I,.  !- : . =.  : : ' .. ..  :

Page I3 ofl3

(

Enclosure 2 to SERIAL: HNP-10-097 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a RELIEF REQUEST: 13R-08 LIST OF REGULATORY COMMITMENTS The following table identifies the actions in this document to which the Harris Nuclear Plant has committed. Statements in this submittal with the exception of those in the table below are provided for information purposes and are not considered commitments.

'Item commitment Completion Date' I Replace temporary non-code repair of defect in weld on line 3SWI-1 14SA-1 with a permanent repair. Temporary non-code repair consists of deferral of code repair until the next RFO-16 scheduled outage exceeding 30 days, RFO-16, provided the (November 2010) condition continues to meet the acceptance criteria of Generic Letter 90-05.

2 Perform weekly inspections of location to detect changes in size or leakage of weld until code, repair is performed. The RFO-16 structural integrity and the monitoring frequency will be re- (November 2010) evaluated if significant changes are found in the condition of the weld area during this monitoring.

Page' 1 of 1