HNP-10-004, Relief Request from ASME Boiler and Pressure Vessel Code, Section XI Requirements for the Service Water System
| ML101170058 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 04/15/2010 |
| From: | Corlett D Progress Energy Carolinas, Progress Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| HNP-10-004 | |
| Download: ML101170058 (16) | |
Text
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¶vagmss n'wfl Serial: HNP-10-004 10 CFR 50.55a APR. 1 5 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., submits the 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. This Relief Request is associated with HNP's Third Ten-Year inservice inspection (ISI) interval.
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 3SW3/4-1630SA-1 indicates a leak point at the interface between 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).
In accordance with Generic Letter (GL) 90-05, 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). contains proposed HNP relief request 13R-06.
Progress Energy Carolinas, Inc.
Harris Nuclear Plant P. 0. Box 165 New Hill, NC 27562 a
HNP-10-004 Page 2 contains the Regulatory Commitments associated with this request.
Please refer any questions regarding this submittal me at (919) 362-3137.
Sincerely, D. H. Corlett Supervisor - Licensing/Regulatory Programs Harris Nuclear Plant DHC/kms
Enclosures:
- 1.
10 CFR 50.55a Request: 13R-06
- 2.
List of Regulatory Commitments cc:
Mr. J. D. Austin, NRC Sr. Resident Inspector, HNP Mr. L. A. Reyes, NRC Regional Administrator, Region II Ms. M. G. Vaaler, NRC Project Manager, HNP
Enclosure I to SERIAL: HNP-10-004 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a REQUEST: 13R-06 Revision 0 Request for Relief for Temporary Non-Code Repair of Service Water Supply Piping Line in Accordance with 10 CFR 50.55a(g)(5)(iii)
--Inservice Inspection Impracticality-1.0 ASME CODE COMPONENT AFFECTED (a)
==
Description:==
Interface between a sockolet and sockolet-to-pipe carbon steel weld on line 3SW3/4-1630SA-1, a three-quarter inch supply piping to root valve ISW-1344.
This piping is an instrument test connection line off of the "A" Train Emergency Service Water (ESW) supply piping to the "A" Essential Service Chilled Water Condenser.
(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 uitimate 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.
(c)
Class:
ASME Code Class 3 (d)
Description of the flaw:
A through-wall leak was found on the carbon steel sockolet and sockolet-to-pipe weld interface on instrument test connection line 3SW3/4-1630SA-1 off the "A" Train ESW supply piping to the "A" Essential Service Chilled Water Condenser.
Only slight moisture accumulation can be seen at the leak point with no actual quantifiable leak rate with the "A" ESW header supplied by Normal Service Water (NSW) and pressure at approximately 80 psig.
Although the three-quarter inch piping and downstream valve were recently replaced during RFO-15 (4/2009), the sockolet itself was not replaced during this outage. The paint was removed from the area directly surrounding the leak, revealing an axial indication along the weld surface indicative of the start/stop location of the weld.
Page 1 of 13
Enclosure Ito SERIAL: HNP-10-004 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a REQUEST: 13R-06 Revision 0 Based on the proximity of the leak to this location, it is likely that a lack of fusion or other defect existed here following installation and a small crevice was created partially or fully between the ID and OD of the weld/sockolet interface.
The cause of this defect could be a result of impurities, work practices or workmanship. Based on the significant number of similar welds performed on the ESW system during RFO-15 and previous outages, with no other failures identified, there is no programmatic or widespread deficiency present. This is considered an isolated incident.
(e)
Flaw Detection:
The flaw was identified on December 7, 2009, during operator rounds. The plant was in Mode 1 at 100% 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 REQUIREMENT Per NRC Inspection Manual Part 9900 Technical Guidance, "Operability Determinations
& Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety," Section C. 12, "If a leak is discovered in a Class 1, 2, or 3 component while conducting an inservice 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 components and their supports. This is used whenever a flaw is discovered that does not meet the ASME requirements.
Per IWA-41 10 of IWA-4000 (Scope):
(a) The requirements of this Article apply regardless of the reason for the repair /replacement activity 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, including appurtenances, Page 2 of 13 to SERIAL: HNP-10-004 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a REQUEST: 13R-06 Revision 0 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 modifying items or systems. These requirements are applicable to procurement, design, installation, examination, and pressure testing of items within the 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 outage of sufficient duration, but no later than the next refueling outage, provided the conditions of Generic Letter (GL) 90-05, "Guidance for Performing 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 defined to exist if:
" The flaw detected during plant operation is in a section of Class 3 piping that cannot be isolated to complete a code repair within the time period permitted by the limiting condition of operation of the affected system as specified in the plant Technical Specifications, and
" Performance of code repair necessitates a plant shutdown.
The identified flaw is a pinhole leak at the interface between a sockolet and sockolet-to-pipe weld on 3 SW3/4-1630SA-1, the supply line to root valve I SW-1344 and a downstream pressure indicator test connection point. This three-quarter inch instrument test connection line is off of the 'A' Train Emergency Service Water (ESW) supply piping to the 'A' Essential Service Chilled Water Condenser. To repair this weld, isolation of the "A" ESW system would require header depressurization and potentially freeze sealing.
The HNP Technical Specifications (TS) Limiting Condition for Operation (LCO) associated with the ESW System is:
3.7.4 At least two independent emergency service water loops shall be OPERABLE.
Page 3 of 13
Enclosure Ito SERIAL: HNP-10-004 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a REQUEST: 13R-06 Revision 0 APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With only one emergency service water loop OPERABLE, 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Since the identified repair cannot be performed in the above TS timeframe allowed for operating with one train of the emergency service water loops out of service, impracticality exists in accordance with the above GL 90-05 definition.
5.0 BURDEN CAUSED BY COMPLIANCE In order to comply with the Code requirement, the plant would need to be shut down to perform the repair. As noted in GL 90-05, "The rather frequent instances of small leaks in 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."
Per this NRC determination that temporary non-code repair of Class 3 piping that cannot be isolated without a plant shutdown is justified in some instances, HNP requests approval for this temporary non-code repair of code Class 3 piping based on the impracticality in performing an ASME Code repair while the plant is operating.
6.0 PROPOSED ALTERNATIVE AND BASIS FOR USE In accordance with the guidelines of GL 90-05, HNP is proposing to defer repair of the identified flaw until the next scheduled outage exceeding 30 days but no later than the next refueling outage, currently scheduled to begin in October 2010.
To ensure that the acceptance criteria of GL 90-05 continue to be met, HNP has implemented compensatory actions to detect changes in the condition of the identified defect.
Page 4 of 13
Enclosure Ito SERIAL: HNP-10-004 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a REQUEST: 13R-06 Revision 0 6.1 SCOPE An indication of a through-wall leak was found on a sockolet fitting at the interface between a sockolet and socket weld on 3SW3/4-1630SA-1, the supply line to root valve 1SW-I 344 and a downstream pressure indicator test connection point. This three-quarter inch instrument test connection line is off of the "A" Train Emergency Service Water (ESW) supply piping to the "A" Essential Service Chilled Water Condenser.
Deferral of the code repair of the identified through-wall flaw will not impact the capability of the ESW system to perform its intended safety-related function, based on the following:
- The current loss of flow from the ESW system is negligible compared to the total system flow. The leak rate is less than one drop per minute at normal operating pressure with NSW supplying the "A" ESW header. ESW system flow is typically 13,000 to 18,000 gpm;
- Based on past ESW through-wall leak analyses, a weld leak is indicative of crevice corrosion which typically creates a localized flaw with no mechanism to propagate rapidly into the adjoining pipe or tee section;
- Based on Non-Destructive Evaluation (NDE) Ultrasonic Testing (UT) measurements adjacent to the weld and on the HNP Civil/Structural Design Engineering evaluation, there is no impact on the structural integrity of the ESW line involved;
" Due to the small leakage rate, the leak is not affecting any other equipment important to safety in the immediate area. If the leakage rate were to increase, housekeeping devices could be installed to shield adjacent equipment;
- In accordance with the guidance in GL 90-05, an augmented inspection of five additional susceptible locations was performed within fifteen days of flaw detection.
6.2 SPECIFIC CONSIDERATIONS There are no other identified leaks in the ESW System at this time. An augmented inspection of five other similar locations was performed in accordance with GL 90-05 to determine the extent of condition. 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 three-quarter inch pipe and 12 inch pipe adjacent to the sockolet were recorded with thicknesses near nominal values. There was no general area thinning identified.
Page 5 of 13
Enclosure Ito SERIAL: HNP-10-004 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a REQUEST: 13R-06 Revision 0 A walkdown of line 3SW3/4-1630SA-1 with WC-2A in-service and NSW supplying "A" ESW revealed little to no vibration present on the line that could be felt by touch. This line extends vertically from 12 inch line 3SW12-83SA-1. Since the three-quarter inch line is less than 2 feet in length and approximately 20 pounds, it is therefore quite rigid.
The 12 inch line is supported by a 3-way hanger within 2 feet axially of the three-quarter inch line. Based on this, vibration induced fatigue wear of the line is not a concern.
Based on the GL 90-05 evaluation, a code repair will be completed no later than the next rescheduled outage (RFO-16). A compensatory action has been initiated for Operations to perform a qualitative assessment via a walkdown inspection of the leakage from the flaw at least weekly, meeting the GL 90-05 requirements. In accordance with the guidance in GL 90-05, UT measurements of the area of the leak will be taken every three months to assess the integrity of the piping until the leak is repaired.
6.3 CAUSE OF LEAK The pinhole is on a sockolet fitting at the interface of the sockolet and sockolet-to-pipe weld. While the three-quarter inch piping and downstream valve were replaced during RFO-15 (April 2009), the sockolet itself was not replaced during this outage. When the paint was removed from the area directly surrounding the leak, an axial indication along the weld surface indicative of the start/stop location of the weld was revealed. Based on the proximity of the leak to this location, it is likely that a lack of fusion existing here and a small crevice was created partially between the ID and OD of the weld/sockolet interface.
Although the exact cause of the weld defect cannot be determined, it could be a result of impurities, work practices, or workmanship. If the initial flaw was not present through the entire thickness of the sockolet, then the crevice or pit likely propagated through a corrosion process. Once a small crevice becomes wetted by service water, its passive film begins to break down. The immediate area surrounding the region then exhibits corrosion in a mechanism similar to pitting corrosion, until a through wall leak is present.
This corrosion process is normal on carbon steel components in service water applications.
Although welds and their surrounding heat affected zones are particularly susceptible to local corrosion attack, corrosion cells formed by this mechanism tend to remain localized and not propagate rapidly into adjacent areas.
Page 6 of 13 to SERIAL: HNP-10-004 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a REQUEST: 13R-06 Revision 0 6.4 STRUCTURAL INTEGRITY OF LINES NDE (UT) measurements were taken on the affected area by a qualified HNP Quality Control (QC) inspector on December 08, 2009, and reviewed by Harris Mechanical/Civil Design Engineering.
The UT data (Attachment 2) shows the areas surrounding the pinhole are near nominal thickness values. The thickness of the actual sockolet varies, although all measurements were recorded greater than 0.200". Near nominal thickness values were recorded for the three-quarter inch pipe and the 12 inch pipe adjacent to the sockolet, with no general area thinning identified.
The leak and surrounding UT data have been evaluated by HNP Mechanical/Civil Design Engineering. The maximum pipe stress ratio from Calculation 8050-79 at this location is 0.466.
A Flaw Evaluation, performed in accordance with the GL 90-05 "Through-Wall Flaw Approach," demonstrates that the piping connection containing the pinhole leak is structurally adequate.
A walkdown of line 3SW3/4-1630SA-1 with WC-2A in-service and NSW supplying "A" ESW revealed little to no vibration present on the line (i.e. could not be felt by touch).
This line extends vertically from 12 inch line 3SW12-83SA-1. Since the three-quarter inch line is less than two feet in length and approximately 20 pounds, it is therefore quite rigid. Additionally, the 12" line is supported by a 3-way hanger within two feet axially of the three-quarter inch line. Therefore, vibration induced fatigue wear of the line is not a concern.
Based on the above, this leak does not represent a concern for the structural integrity of the three-quarter inch supply line 3SW3/4-1630SA-1. Other than monitoring, no remedial measures are required to ensure integrity is maintained.
6.5 FLAW EVALUATION GL 90-05 "through-wall" flaw evaluation criteria was used to evaluate the pinhole leak in ESW line 3SW3/4-1630SA-1. The GL 90-05 criteria are applicable since this line is ASME Section III, Class 3 piping per the EDB. For conservatism purposes, analyses using both the circumferential and axial flaw methodologies were performed.
This three-quarter inch line, 3SW3/4-1630SA-1, is the supply piping to root valve 1SW-1344 and a downstream pressure indicator test connection point. Leakage, as indicated Page 7 of 13
Enclosure Ito SERIAL: HNP-10-004 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a REQUEST: 13R-06 Revision 0 by slight moisture accumulations, was reported on a sockolet fitting at the interface between a sockolet and sockolet-to-pipe weld. There is no actual quantifiable leak rate with the "A" Emergency Service Water (ESW) header supplied by Normal Service Water (NSW) at approximately 80 psig pressure.
The allowable stress for the pipe material, ASTM A106, Grade B per EDB and stress calc 8050-79, is:
S = 15000 psi reference ASME Section III Appendices The pipe properties for 3SW3/4-1630SA-1 (three-quarter inch, schedule 40 pipe) are:
Do = 1.050 in t = 0.100 in tnom = 0.113 in (reference: Engineering Design Basis (EDB), Calculation 8050-79, NAVCO Piping Datalog)
Actual measured thickness used for flaw evaluation which is greater than tmin The operating temperature is 125 deg F and design pressure is 150 psig (reference EDB):
p=150psi T=125deg The minimum required wall thickness for hoop stress per ASME Section III is:
tm =
pDo
= 0. 005 in 2(S + 0.4p)
Per the applicable pipe stress calculation (8050-79), the maximum stress ratio is 0.466.
The section modulus for three-quarter inch pipe, schedule 40, is 0.0706 in3. The minimum wall thickness measured is 0.103 inch.
Page 8 of 13
Enclosure Ito SERIAL: HNP-10-004 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a REQUEST: 13R-06 Revision 0 The "through-wall" flaw evaluation is as follows:
Allowable Stress Intensity Factor ksi = 1000 psi The allowable stress intensity factor (K) for determining the acceptability of flawed piping under the applied load is:
K < 35 ksi(in)°.5 CS allowable stress intensity factor for flaws (reference NRC GL 90-05)
Circumferential Flaw Evaluation The following through-wall flaw evaluation is performed in accordance with GL 90-05,, Section C.3.a. The stress intensity factor for through-wall flaw (including safety factor of 1.4) is:
K = 1.4 (s) (F) (3.1416*a)°.s, where:
s (combination of deadweight, pressure, thermal, seismic stresses) = MA + MBE = 7461 psi Moment Stress due to Deadweight and Pressure:
MA = 3278 psi Moment Stress due to DBE Seismic:
MBE = 4183 psi F (geometry factor) = 1 + (A)(c)1.5 + (B)(c) 2.5 + (C)(c) 3.5 = 5.024 Nomenclature:
A = -3.26543 + 1.52784 ()R 0.072698 (R
+ 0.0016011 R 3= 701.572 B = 11.36322 - 3.91412 (-R ) + 0.18619 (--) - 0.004099(R
= -1.793 x 103 C = -3.18609 + 3.84763 -
0.18304(-
+ 0.00403
= 1.771 x 103 a
C a
= 0.034 (3.1416)(R)
Page 9 of 13
Enclosure Ito SERIAL: HNP-10-004 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a REQUEST: 13R-06 Revision 0 R = mean pipe radius = Do-tnom = 0.469 in.
2 2a (the flaw length, in.) = the diameter of the pinhole (0.1"), reference GL 90-05 a = 0.050 in Therefore, K = 1.4 (s) (F) (3.1416*a)0.5 = 20.8 ksi(in)0.5 Since the evaluated K is less than the stress intensity factor of 35 ksi(in)°5 for the CS material, this flaw meets the GL 90-05 criteria for a temporary non-code repair of the Class 3 piping.
6.6 AUGMENTED INSPECTIONS Since the flaw has been evaluated and found acceptable by the GL 90-05 "Through-Wall Flaw" approach, an augmented Ultrasonic (UT) inspection was performed to assess the overall degradation of the affected system. In accordance with GL 90-05 Section C.4, inspection of at least five most susceptible (and accessible) locations were performed.
In addition, measurements of the area where the flaw is located will be performed every three months, as required by GL 90-05.
There are no operating mode restrictions associated with this condition. The leak will be visually inspected weekly by Operations until it is repaired in RFO-16. Normal walkdowns are performed by system engineers once every quarter in accordance with 10 CFR 50.65, Maintenance Rule. This increase in periodicity of inspection is in accordance with the criteria of GL 90-05, which recommends that a qualitative assessment of leakage be performed at least weekly to determine any degradation of structural integrity.
The leak will be permanently repaired by piping replacement or a code weld repair in RFO-16.
6.7 CONCLUSION
The minimum wall thickness, maximum bending stress and through-wall flaw evaluations performed for the pinhole flaw demonstrate that the piping, including the weld joining the pipe to the sockolet for line 3SW3/4-1630SA-1, is structurally adequate in accordance with the guidance provided in GL 90-05. The flaw was evaluated using the through-wall flaw Page 10 of 13
Enclosure Ito SERIAL: HNP-10-004 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a REQUEST: 13R-06 Revision 0 fracture mechanics methodology provided by NRC Generic Letter 90-05, which is a conservative approach to evaluating the pinhole leak.
Since the flaw in line 3SW3/4-1630SA-1 satisfies the criteria of the above evaluation approach, it is acceptable to propose a temporary non-code repair of the code Class 3 piping in accordance with GL 90-05. Furthermore, the provisions of the through-wall flaw evaluation per GL 90-05 demonstrate that the structural integrity of the ESW piping components is not adversely affected by this defect.
Based on the above analysis, HNP is requesting NRC approval per 10 CFR 50.55a(g)(5)(iii) to defer ASME Section XI IWA-4000 repair/replacement requirements for the identified flaw in accordance with the guidance provided in GL 90-05.
7.0 DURATION OF PROPOSED ALTERNATIVE Repair of the defect will be deferred until the next scheduled outage exceeding 30 days, but no later than the next refueling outage, provided the condition continues to meet the acceptance criteria of Generic Letter 90-05. HNP is currently monitoring the leak location.
HNP's next refueling outage, RFO-16, is currently scheduled to begin in October 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 Page 11 of 13 to SERIAL: HNP-10-004 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a REQUEST: 13R-06 Revision 0
'QA UT-8 Rev 7
~2Progress Energy GGMUP NDE Report#
N/A Page 1 of 2
DIGITAL ULTRASONIC THICKNESS NDE REPORT Plant 0 BNP C] CR3 0 HNP 0- RNP WO:
N/A Unito 1 02 03 Date:
12/09/09 Component I Item Tested:
Vent line for 1 SW-1344 NDE Procedure:
No.:
42R7 Rev.:
7 TR: N/A Component Material:
Expected 0 C/S SA-106 0 SS N/A0 Other (Specify):
N/A Nominal T Range:
Various See spread sheet
-7ype
__________________(Specify):_
Thickness Gauge:
Couplant:
Mfg.: panametrics Model: 36DL+
S/N: 32386301 Software Rev. No.:2,07/1.2 Brand:
Ultral Gel Batch No.: 041250 Calibration I Reference Std.:
0 Test Item-Mic./Caliper No.:
NIA Primary Cal. Thickness: 0.100 to 0.500 0 Step Block S/N: CT-2076
[] CIS 0 SIS 0 Other (Describe Below)
Cal. Check Thickness: 0.100to0.500 Transducer.
D798 79797
.200 7.5 MHZ Mfg.: Panametrics Model:
D790SM S/N:
119783 Diameter:
.312 Freq.:
5.0 MHZ
[]Single 09 Dual Component Conditions:
Inst. Receiver Gain Setting Auto dB Other Test Conditions:
High Temp:
0 Yes 0 No Technique: 0 Single-Echo 0 Thru.Coat Cal Block-UTC: 1745764 Coated/Painted: 0 Yes 0 No 0 Multiple-Echo Sketch component or item and area tested. Include thickness data.
See following pages for exam results.
3W4 line Leak LOCwtiof 1
2 Headew Inspector:
Dave Gerber Certification Level Date Level 1 A-Scan 12/9/09 Inspector: N/A Certification Level Date N/A N/A Reviewed By (If Required)
Title Date
_______________-_. 1 Dte NGGM-PM-0O01 1 APPENDIX A Page 12 of 13 to SERIAL: HNP-10-004 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a REQUEST: 13R-06 Revision 0 Report No.:
N/A Page 2
of 2
Sys. / Comp. ID:
4085 Material:
Carbon Steel High Reading: *0.138" Exam Item:
Size:
314" Scr 40 Low Reading: "0.103" DWG No.:
8-G-498S02, 5-S-998S02 Thickness:
00,113" Grid Size:
I" WR / Mod:
N/A Configuration:
Pipe Datum Point:
C/L US Socr-o.et Category:
Section II ASME Class:
3 Acceptance Standard:
0.080" Per Engineennn UT thickness request Procedure:
NDEP 427 Rev.:
7 Socket and 3/4" Vent Base metal (12" pipe)
A B
C
,D Min
-Mak-_Avg.
Aý ',:
l C
Min Maxii Avg 1
0.265 0,248 0.241 0.226 0.226 0,265 0.245 1'.
0.436 0.437 07432 0.434 0.432 0A437 0.435 2
0.109 0.112 0.119 0.127 0.109 0.127 0.117 2.- 0.442 0.402 0.406 0.436 0.402 0.442 0.422 3.
0,103 0,112 0.127 0.124 0.103 0.127 0.117 3,*.
0.438 0.407 0.397 0.451 0.397 0.451 0.423 4
0.117 0.124 0.133 0.126 0.117 0,133 0.125 Min 0.436 0.402 0.397 0.434 5
0,118 0.129 0.138 0.129 0.118 0.138 0.129 Max 0,442 0.437 0432 0.451 6
0.128 0.129 0.138 0.125 0.125 0.138 0.13 Avg 0.439 0.415 0,412 0,44 7
0.129 0.122 0.128 0.126 0.122 0.129 0.126 TMin 0.397 Min 0.103 0.112 0.119 0.124
,TMax 0.451 Max 0.265 0.248 0.241 0.226 TAv9 0.427 Avg 0.138 0.139 0.146 0.14 TMin 0.103 TMax 0.265 TAvg 0.141 Remarks:
-Note that the High and Low reading on this report Examiner:
Dave Gerber Level: 1 Ascan Date' 12/08/09 is for the 3/4" sen 40 pipe only.
Examiner:
N/A Level:
NIA Date:
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N/A Page 13 of 13 to SERIAL: HNP-10-004 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 10 CFR 50.55a RELIEF REQUEST: 13R-06 LIST OF REGULATORY COMMITMENTS The actions in this document committed to by Harris Nuclear Plant (HNP) regarding 10 CFR 50.55a Request 13R-06, "Request for Relief for Temporary Non-Code Repair of Service Water Supply Piping Line in Accordance with 10 CFR 50.55a(g)(5)(iii)," are identified in the following table. Statements in this submittal, with the exception of those in the table below, are provided for information purposes and are not considered commitments. Please direct any questions regarding this document or any associated regulatory commitments to the Supervisor, Licensing/Regulatory Affairs.
Item C-commitment 0
Copetion Date 1
Replace temporary non-code repair of defect in weld on line 3SW3/4-1630SA-1 with a permanent repair. Temporary non-code repair consists of deferral of code repair until the next RFO-16 (Nov. 2010) scheduled outage exceeding 30 days but no later than the next scheduled refueling outage, provided the 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 structural integrity and the monitoring frequency will be re-RFO-16 (Nov. 2010) evaluated if significant changes are found in the condition of the weld area during this monitoring.
3 Perform ultrasonic measurements of the area where the flaw is located at least once every 90 days.
RFO-16 (Nov. 2010)
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