ML051400273
| ML051400273 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 05/11/2005 |
| From: | Jamil D Duke Energy Corp |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML051400273 (13) | |
Text
Dukte D.M. JAMIL Dukoer Vice President Duke Power Catawba Nuclear Station 4800 Concord Road / CNO1 VP York, SC 29745-9635 803 831 4251 803 831 3221 fax May 11, 2005 U.S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, D.C. 20555
Subject:
Duke Energy Corporation Catawba Nuclear Station, Unit 2 Docket Number 50-414 Request for Relief Number 05-CN-003 Limited Volumetric Inspection Coverages During End-of-Cycle 12 and End-of-Cycle 13 Refueling Outages Pursuant to 10 CFR 50.55a(g)(5)(iii), please find attached Request for Relief 05-CN-003.
This request for relief is associated with limited volumetric inspection coverages during the subject outages.
While this request primarily addresses examinations during the End-of-Cycle 13 outage, information is also being provided regarding four steam generator nozzle inside radii inspections which were reported as limited in the End-of-Cycle 12 outage inservice inspection report.
These radii were all re-examined in the End-of-Cycle 13 outage and 100% coverage was achieved.
The attachment to this letter contains all technical information necessary in support of this request for relief.
Duke Energy Corporation is requesting NRC review and approval of this request at your earliest opportunity.
There are no regulatory commitments contained in this letter or its attachment.
If you have any questions concerning this material, please call L.J. Rudy at (803) 831-3084.
AoYQl Catawba Nudear Station20th Annitrmary www.duepowercom 1985-2005
Document Control Desk Page 2 May 11, 2005 Very truly yours, D.M. Jamil LJR/s Attachment xc (with attachment):
W.D. Travers, Regional Administrator U.S. Nuclear Regulatory Commission, Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 E.F. Guthrie, Senior Resident Inspector U.S. Nuclear Regulatory Commission Catawba Nuclear Station S.E. Peters, Project Manager (addressee only)
U.S. Nuclear Regulatory Commission Mail Stop 0-8 G9 Washington, D.C. 20555-0001
Note:
Throughout this request for relief, the following Duke Energy system designations are utilized:
EMF -
Process and Area Radiation Monitoring System FW -
Refueling Water System FWST -
Refueling Water Storage Tank NC -
Low Head Injection/Residual Heat Removal System NV -
High Head Injection/Centrifugal Charging System
Request for Relief 05-CN-003 Revision 0 Page 1 of 10 Proposed Relief in Accordance with 10 CFR 50.55a(g)(5)(iii)
Inservice Inspection Impracticality Duke Energy Corporation Catawba Nuclear Station - Unit 2 (EOC-12 for four Inside Radius Inspections; See Section IX of this Relief Request) and (EOC13 for 11 welds)
Second 10-Year Interval - Inservice Inspection Plan Second Interval Start Date - August 19, 1996 Second Interval End Date - August 19, 2006 ASME Section XI Code - 1989 Edition with No Addenda II.IV.
&V.
VI.
Vll.
Vill.
Request Limited System Code Requirement from Impracticality/
Proposed Implementation Justification Number Area/Weld Component for Which Relief is Requested:
Burden Caused Alternate Schedule and for Granting I.D.
Which Relief is 100% Exam Volume Coverage by Code Examinations Duration Relief Number Requested:
Exam Category Compliance or Testing Area or Weld to be Item No.
Examined Fig. No.
Limitation Percentage 2SGA4INLET Steam Generator 2A Category B-D, Item Number Examination Data for ID. Number 2SGA-INLET is shown in Attachment Primary Nozzle (Inside B03.140.001 A-EOC 12 Radius)
This Item was reported as limited In Outage EOC12, (66.74%). Re.
Examination Data for ID. Number 2S-A-iNLET (Item Number examination in EOC13 achieved 100%
B03.140.001A) is shown in Attachment B-EOC13) coverage.
See Section IX for an explanation.
- 2.
2SGA-OULET Steam Generator 2A Category B-D, Item Number Examination Data for ID. Number 2SGA-OUTLET is shown in Attachment Primary Nozzle (Inside B03.140.002 C-EOC 12 Radius)
This item was reportedi as imited R n Examination Data for ID. Number 2SGA-OUTLET (Item Number examination In EOC13 achieved 100%
B03.140.002A) is shown in Attachment D-EOC13) coverage.
See Section IX for an explanation.
- 3.
2SGD.INLET Steam Generator 2D Category B-D, Item Number Examination Data for ID. Number 2SGD-INLET is shown in Attachment E Primary Nozzle (Inside B03.140.007
- EOC12 Radius)
This item was reported as limited In Examination Data for ID. Number 2SGD-INLET (Item Number B03.140.007A) examination in EOC13 achieved 100%
is shown in Attachment F - EOCI3 coverage.
See Section IX for an explanation.
- 4.
2SGD-OUJLEr Steam Generator 2D Category B-D, Item Number (Examination Data for ID. Number 2SGD-OUTLET is shown in Primary Nozzle (Inside B03.140.008 Attachment G - EOC 12)
Radius)
This Item was reported as limited in Outage EOC12, (66.74%). Re-(Examination Data for ID. Number 2SGD-OUTLET (Item Number examination In EOCI3 achieved 100%
1B03.140.008A) is shown in Attachment H - EOC13) coverage.
See Section IX for an explanation.
Request for Relief 05-CN-003 Revision 0 Page 2 of 10 II.
ILII.
IV. &V.
VI.
Vll.
Vill.
Request Limited System Code Requirement from Impracticality/
Proposed Implementation Justification Number Area/Weld Component for Which Relief is Requested:
Burden Caused Alternate Schedule and for Granting I.D.
Which Relief is 100% Exam Volume Coverage by Code Examinations Duration Relief Number Requested:
Exam Category Compliance or Testing Area or Weld to be Item No.
Examined Fig. No.
Limitation Percentage 5
2RPV-101-141 Reactor Vessel Category B-A, Item Number See Paragraph "A" See Paragraph "F" See Paragraph "G" See Paragraph Shell-to-Lower Head 1301.011.001 "H"
Circumferential Weld Figure Number IWB-2500-1(b)
Volume Limitation 76.8%
(Attachment 1) 6 2RPV-101-142A Reactor Vessel Category B.A. Item Number See Paragraph "A" See Paragraph "F" See Paragraph "G" See Paragraph Lower Shell Longitudinal I301.012.007 "H"
Seam Weld @ 60 Degrees Figure Number IWD-2500-2 Volume Limitation 77.8%
(Attachment J) 7 2RPV-10l-142B Reactor Vessel Category B-A, Item Number See Paragraph "A" See Paragraph "F' See Paragraph "G" See Paragraph Lower Shell Longitudinal B01.012.008 "H"
Seam Weld @ 180 Degrees Figure Number IWB-2500-2 Volume Limitation 77.8%
(Attachment K) 8 2RPV-101-142C Reactor Vessel Category B-A, Item Number See Paragraph "A" See Paragraph "F' See Paragraph "G" See Paragraph Lower Shell Longitudinal 1301.012.009 "H"
Seam Weld 0 300 Degrees Figure Number IWB-2500-2 Volume Limitation 77.8%
(Attachment L) 9 2RPV-101-151 Reactor Vessel Category B-A, Item Number See Paragraph "A" See Paragraph "F' See Paragraph "G" See Paragraph Shell-to-Lower Head B01.021.001 "H"
Circumferential Weld Figure Number 1W1-2500-3 Volume Limitation 53.3%
(Attachment M) 10 Pressurizer Category B-D, Item Number See Paragraph "B" See Paragraph "F' See Paragraph "G" See Paragraph 2PZR-W3 Safety Nozzle to Upper B03.110.003 IT Head Figure Number IWB-2500-7(b)
ASME Section V. Article 4, T-424.1 Volume Limitation 79.2%
(Attachment N)
Request for Relief 05-CN-003 Revision 0 Page 3 of 10 1.II.
III.
IV. &V.
VI.
VII.
l VIll.
Request Limited System
/
Code Requirement from Impracticality/
Proposed Implementation Justification Number Area/hVeld Component for Which Which Relief is Requested:
Burden Caused Alternate Schedule and for Granting I.D.
Relief is Requested:
100% Exam Volume Coverage by Compliance Examinations Duration Relief Number Area or Weld to be Exam Category or Testing Examined Item No.
Fig. No.
Limitation Percentage 11 Pressurizer Category B-D, Item Number See Paragraph "B" See Paragraph "F' See Paragraph "G" See Paragraph "I" 2PZR-W4A Safety Nozzle to Upper Head B03.110.004 Figure Number IWB-2500-7(b)
ASME Section V. Article 4, T-424.1 Volume Limitation 79.2%
(Attachment 0) 12 Pressurizer Category 13-D, Item Number See Paragraph "B" See Paragraph "F' See Paragraph "G" See Paragraph "I" 2PZR-W4B Safety Nozzle to Upper Head B03.110.005 Figure Number IWB-2500-7(b)
ASME Section V, Article 4, T-424.1 Volume Limitation 79.2%
(Attachment P) 13 Steam Generator 2C Lower Category C-A, Item Number See Paragraph "C" See Paragraph "F' See Paragraph "G" See Paragraphs "J" 2SGC-04B-05 Shell-to-Transition Cone COI.010.002 Figure Number 1WC-2500-1(c)
Volume Limitation 48.3%
(Attachment Q) 14 FW System Category C-F-I, Item Number See Paragraph "D" See Paragraph "F' See Paragraph "G" See Paragraphs "K" 2FWV76-6 Valve 2FW028-to-Pipe C05.011.032 Circumferential Weld Figure Number IWC-2500-7(a)
Volume Limitation 30.8%
(Attachment R) 15 NV System Category C-F-I. Item Number See Paragraph "E' See Paragraph "F' See Paragraph "G" See Paragraphs "L" 2NV34-11 Elbow-to-Tee C05.021.241 Circumferential Weld Figure Number IVC-2500-7(a)
Volume Limitation 85.8%
(Attachment S)_
Request for Relief 05-CN-003 Revision 0 Page 4 of 10 IV. & V. Impracticality/Burden Caused by Code Compliance Paragraph A:
Scanning limitations were caused by the core support lugs, which prevented scanning 100% of the weld length from four orthogonal directions. The procedure, qualified through the Performance Demonstration Initiative requires scanning in four orthogonal directions using 45'single element shear waves, 450 single element refracted longitudinal waves (RL), and 450 dual element RL waves.
For ID Number 2RPV-101-151, Reactor Vessel Shell to Lower Head Circumferential Weld. Scanning limitations were caused by proximity of the bottom mounted instrument tubes, which prevented scanning 100% of the weld length from four orthogonal directions. The procedure, qualified through the Performance Demonstration Initiative requires scanning in four orthogonal directions using 450single element shear waves, 450 single element refracted longitudinal waves (RL), and 450 dual element RL waves.
The percent of coverage reported represents the aggregate coverage from all scans performed on the welds. In order to scan all of the required surfaces for the inspection of the shell to lower head welds and the lower shell longitudinal welds, the interferences would have to be moved to allow scanning the full length of the welds, which is impractical.
These examinations were performed using personnel, procedures and equipment qualified in accordance with ASME Section XI, Appendix VIII, 1995 Edition through the 1996 Addenda as administered through the Performance Demonstration Initiative (PDI).
(Examination Data is shown in Attachments I, J, K, L and M)
Paragraph B:
Scanning limitations were caused by the nozzle geometry that restricts scanning from the nozzle side. The percent coverage reported represents the aggregate coverage of all scans performed. The examination volume was scanned using 45°; 600 shear waves, and straight beam longitudinal waves in accordance with ASME Section V, Article 4, T-441.3.2.1.
The 450 beam covered 92.0% of the examination volume perpendicular to the weld from the vessel head side and 66.7% of the examination volume perpendicular to the weld from the nozzle side. Scans parallel to the weld with the 450 beam covered 79.6% of the examination volume in two opposite directions.
The 60° beam covered 94.2% of the examination volume perpendicular to the weld from the vessel head side and 60.5% of the examination volume perpendicular to the weld from the nozzle side. Scans parallel to the weld with the 600 beam covered 79.6% of the examination volume in two opposite directions.
The straight beam covered 79.6% of the examination volume. In order to achieve more coverage, the nozzle would have to be re-designed to all scanning in four orthogonal directions. This examination was performed using procedures prepared in accordance with ASME Section V, Article 4 using personnel qualified in accordance with ASME Section XI, BVA-2300, including Appendix VII, 1995 Edition through the 1996 Addenda.
(Examination Data is shown in Attachments N, 0 and P)
Paragraph C:
Scanning limitations were caused by the proximity of nine restraint lugs, which prevented scanning of the full volume from four orthogonal directions. The percent coverage reported represents the aggregate coverage of all scans performed. The examination volume was scanned using 350, 450and 600 shear waves, and straight beam longitudinal waves in accordance with ASME Section V, Article 4, T-441.3.2.1.
Request for Relief O5-CN-003 Revision 0 Page 5 of 10 The 350 beam covered 43.7% of the required volume from the shell side perpendicular to the weld and 41.1% from 2 opposite directions parallel to the weld.
The 450 beam covered 42.2% of the required volume from the shell side perpendicular to the weld and 77.0% of the required volume from the cone side. Scans parallel to the weld covered 41.1% of the required volume from two opposite directions.
The 600 beam covered 87.8% of the required volume from the cone side perpendicular to the weld. In order to achieve more coverage the restraint lugs would have to be removed, which is impractical.
(Examination Data is shown in Attachment Q)
Paragraph D:
Single sided access caused by the valve configuration prevented scanning from the valve side of the weld. The percent coverage reported represents the aggregate coverage of all scans performed. The examination volume was scanned using 450, 600 and 700 shear waves.
The 450 beam covered 43% of the required volume in two opposite circumferential directions. The 60° beam covered 37.2% of the required volume in one axial direction from the pipe side of the weld. A 70° shear wave best effort scan covered 62.8% of the required volume in one axial direction from the pipe side of the weld but was not included in the coverage calculation because of the requirements of IOCFR50.55a(b)(2)(xv)(A)(2).
In order to achieve more coverage, the weld would have to be re-designed to allow scanning from four orthogonal directions. This examination was performed using personnel, procedures and equipment qualified in accordance with ASME Section XI, Appendix VIII, 1995 Edition through the 1996 Addenda as administered through the Performance Demonstration Initiative (PDI).
(Examination Data is shown in Attachment R)
Paragraph E:
Single sided access caused by the elbow-to-tee configuration prevented scanning from the tee side of the weld. The percent coverage reported represents the aggregate coverage of all scans performed. The examination volume was scanned using 450, 600 and 70° shear waves.
The 450 beam covered 100% of the required volume in two opposite circumferential directions. The 60° beam covered 63.6% of the required volume in one axial direction from the pipe side of the weld. A 700 shear wave best effort scan covered 56.2% of the required volume in one axial direction from the pipe side of the weld but was not included in the coverage calculation because of the requirements of 10CFR50.55a(b)(2)(xv)(A)(2).
In order to achieve more coverage, the weld would have to be re-designed to allow scanning from four orthogonal directions. This examination was performed using personnel, procedures and equipment qualified in accordance with ASME Section XI, Appendix VIII, 1995 Edition through the 1996 Addenda as administered through the Performance Demonstration Initiative (PDI).
(Examination Data is shown in Attachment S)
Request for Relief O5-CN-003 Revision 0 Pagc6of10 VI.
Proposed Alternate Examinations or Testingt Paragraph F:
The scheduled 10-year code examination was performed on the referenced area / weld and resulted in the noted limited coverage of the required ultrasonic volume. No alternate examinations or testing is planned for the area I weld during the current inspection interval.
VII. Implementation Schedule and Duration Paragraph G:
The scheduled second 10-year interval plan code examination was performed on the referenced welds and components resulting in limited volumetric coverage. No additional examinations are planned for the welds and components during the current inspection interval. The same welds and components may be examined again as part of the next (third) 10-year interval plan, depending on the applicable code year edition and addenda requirements adopted in the future.
VIII. Justification for Granting Relief Paragraph H:
Design and fabrication of the reactor vessels is carried out in strict accordance with ASME Code,Section III, Class I requirements. The head flanges and nozzles are manufactured as forgings. The Unit 2 vessel is made up of several shells, each consisting of formed plates joined by full penetration longitudinal weld seams. The hemispherical heads are made from dished plates. The reactor vessel parts are joined by welding, using the single or multiple wire submerged arc and the shielded metal arc processes. The non-destructive examination of the reactor vessel and its appurtenances is conducted in accordance with the ASME Code Section III requirements; also numerous examinations are performed in addition to ASME Code Section III requirements. Nondestructive examination of the vessel is discussed in UFSAR Section 5.3.1.3 and the reactor vessel quality assurance program is given in UFSAR Table 5-11.
The reactor vessel materials meet the fracture toughness requirements of IOCFR 50, Appendix G, to the extent possible. A summary of the fracture toughness data is given in UFSAR Table 5-13 and 5-15. Appendix G of 10 CFR 50 requires heatup and cooldown of the reactor vessel be accomplished within established pressure-temperature limits.
The heatup and cooldown rates imposed by plant operating procedures are 50'F per hour and 800F per hour, respectively, for normal operation. The heatup and cooldown rate limits are 60 'F per hour and 100 'F per hour, respectively, for abnormal or emergency conditions. The rate of 100°F per hours is the vessel design specification as a normal condition for conservatism for both heatup and cooldown. For Catawba Unit 2, the heatup and cooldown limit curves for normal operation provides a predicted operation window that is sufficient to conduct heatup and cooldowns.
In addition, cyclic loads are introduced by normal power changes, reactor trips, and startup / shutdown operations.
These design base cycles are selected for fatigue evaluation and constitute a conservative design envelope for the projected plant life. Vessel analysis results in a usage factor that is less than 1.
Plant Technical Specifications dictate that a reactor coolant system water inventory balance be performed on a regular basis (i.e. at least once every three days). The normal operating practice is to perform this computer based program on a daily frequency and/or whenever the operators suspect any abnormal changes to other leakage detection systems.
Plant Technical Specification requires system leak from "unidentified" sources be maintained below I gpm; however, plant operation procedures (PT/I(2)1A141501001D, NC System Leakage Calculation) establish an administrative limit of 0.15 gpm above which source of leakage will be investigated. Leakage as a result of a failed weld discussed in this section would show up as unidentified leakage and subject to the 0.15 gpm administrative limit. The water inventory
Request for Relief 05-CN-003 Revision 0 Page 7 of 10 balance provides repeatable results less than the 0.15 gpm administrative limit; however, an evaluation of sensitivity below this leak rate level has not been performed. No analysis has been done to quantify the flaw size in the reactor vessel which would be detectable by the leakage detection system.
Other leakage detection systems available to the operator and dictated per plant technical specifications are:
Containment Atmosphere Particulate Radioactivity (EMF 38) Monitoring System which would detect airborne radiological activity; Containment Ventilation Unit Condensate Drain Tank Level Monitoring which collects and measures as unidentified leakage the moisture removed from the containment atmosphere.
The above leakage detection methods are dependant upon the containment atmosphere cooling via the Lower Containment Ventilation System, which provides for forced circulation of cooling air across the reactor vessel and for subsequent air return to lower containment. This provides the motive force for transporting moisture and radioactivity from any through wall leak in the reactor vessel to the above described leakage detection monitors.
Paragraph I:
100% bare metal visual (BMV) examinations were performed on all Alloy 600 1 82/182 weld locations on the pressurizer during 2EOC13 (Ref. Duke's Response to NRC Bulletin 2004-01, dated February 24, 2005 (See Attachment T), W/O 98642405-01, and W/O 98636907-01). The inspection scope included the three (3) pressurizer safety valve nozzle to upper head locations. No evidence of leakage or boric acid deposits was observed at any of these locations. There was no wastage identified on the external surface of the pressurizer vessel head.
Plant Technical Specifications dictate that a reactor coolant system water inventory balance be performed on a regular basis (i.e. at least once every three days). The normal operating practice is to perform this computer based program on a daily frequency and/or whenever the operators suspect any abnormal changes to other leakage detection systems.
Plant Technical Specification requires system leak from "unidentified" sources be maintained below 1 gpm; however, plant operation procedures (PT/I1(2)1A/4150/OOID, NC System Leakage Calculation) establish an administrative limit of 0.15 gpm above which source of leakage will be investigated. Leakage as a result of a failed weld discussed in this section would show up as unidentified leakage and subject to the 0.15 gpm administrative limit. The water inventory balance provides repeatable results less than the 0.15 gpm administrative limit; however, an evaluation of sensitivity below this leak rate level has not been performed.
Other leakage detection systems available to the operator and dictated per plant technical specifications are:
Containment Atmosphere Gaseous and Particulate Radioactivity Monitoring System (EMF monitors 38 &
- 39) which would detect airborne radiological activity; Containment Ventilation Unit Condensate Drain Tank Level Monitoring Subsystem which collects and measures as unidentified leakage the moisture removed from the containment atmosphere; Containment Floor and Equipment Sump Level and Flow Monitoring Subsystem where unidentified accumulated water on the containment floor would be monitored and evaluated as sump level changes.
Paragraph J:
All pressure boundary materials used in the steam generator are selected and fabricated in accordance with the requirements of Section III of the ASME code. Although the ASME classification for the secondary side is specified to be Class 2, the current philosophy is to design all pressure retaining parts of the steam generator, and thus both the primary and secondary pressure boundaries, to satisfy the criteria specified in Section 111 of the ASME Code for Class I components. The design stress limits, transient conditions and combined loading conditions applicable to the steam
Request for Relief 05-CN-003 Revision 0 Page 8 of 10 generator are discussed in UFSAR Section 3.9.1. The fracture toughness of the materials is discussed in UFSAR Section 5.2.3.3.
The weld is located on the secondary side at the interface between the Steam Generator Lower Shell to Transition Cone. Any through-wall leak at this location would result in a loss of secondary feedwater either as fluid leakage to the containment floor or as steam leakage to increase containment humidly. The leakage detection systems available to the operator to detect secondary feedwater leakage are:
Containment Ventilation Unit Condensate Drain Tank Level Monitoring Subsystem which collects and measures as unidentified leakage the moisture removed from the containment atmosphere; Containment Floor and Equipment Sump Level Monitoring Subsystem where unidentified accumulated water on the containment floor would be monitored and evaluated as sump level changes.
In addition, each steam generator enclosure compartment is equipped with temperature monitors. Any increase in indicated temperature would alert the control room operator to high energy fluid leakage within the associated steam generator enclosure compartment.
Paragraph K:
The Refueling Water (FW) System is periodically tested (i.e., once every 18 months) to reduce leakage to as low as possible. Visual inspection of the FW System is performed, in part, by aligning the Refueling Water Storage Tank (FWST) for recirculation and separately starting each FW recirculation pump. The recirculation flow path is visually inspected for leakage. Components with identified leakage are required to have a Work Request/Order submitted to repair the leak, a corrective action program entry is initiated to track the status of the leak, and boric acid corrosion concerns reported to Maintenance. This periodic test includes verification of leakage at the weld, which is located in Auxiliary Building Room 110 on the FWST side of isolation valve 2FW028 (Reference Operating Procedure OP/2/A14207/007, Enclosure 13.2).
The weld is on the FWVST side of check valve 2FW28, which provides a suction source from the FWST to ND Pump 2A. This piping is typically maintained in standby readiness and at a pressure equivalent to the fluid head from the FWST. As such, there are no other direct means employed to ensure system integrity at this weld location.
Paragraph L:
The Chemical and Volume Control (NV) System outside containment is periodically tested (i.e., once every 18 months) to reduce leakage to as low as possible. The NV System is aligned to pressurize the system and a visual inspection of system components performed to identify leakage. Components with identified leakage are required to have a Work Request/Order submitted to repair the leak, a corrective action program entry is initiated to track the status of the leak, and boric acid corrosion concerns reported to Maintenance. This periodic test includes verification of leakage at the weld, which is located in Auxiliary Building Room 243 on the high pressure side of isolation valve 2NV307 (Reference Operating Procedure OPt2/A14206/006, Enclosure 13.2).
In addition, any leakage at the weld would be detected via the reactor coolant system water inventory balance test.
Plant Technical Specifications dictate that a water inventory balance be performed on a regular basis (i.e. at least once every three days). The normal operating practice is to perform this computer based program on a daily frequency and/or whenever the operators suspect any abnormal changes to other leakage detection systems. Plant Technical Specification requires system leak from "unidentified" sources be maintained below 1 gpm; however, plant operation procedures (PT/1(2)/A/41I50/00D, NC System Leakage Calculation) establish an administrative limit of 0.15 gpm above which source of leakage will be investigated. Leakage as a result of a failed weld discussed in this section would show up as unidentified leakage and subject to the 0.15 gpm administrative limit. The water inventory balance provides repeatable results less than the 0.15 gpm administrative limit; however, an evaluation of sensitivity below this leak rate level has not been performed. No analysis has been done to quantify the flaw size at the weld which would be detectable by the leakage detection system.
Request for Relief 05-CN-003 Revision 0 Page 9of10 IX. Other Information UTexaminations were performed on Steam Generators 2A and 2D Inlet and Outlet Nozzle Inside Radius Sections during the EOC12 Refueling Outage. The examinations were recorded as having a limited volumetric coverage of 66.74% in the Outage Summary Report following the 2003 EOC12 Refueling Outage. Further, the report stated that the examination results would be submitted as part of a relief request. Subsequently to report submittal, it was determined that a different UT examination procedure may provide better results. The new examination was conducted during the next outage (2004 EOC13 Refueling Outage) where 100% coverage was obtained. Although no action is needed from the NRC, the examination result was recorded here for documentation closeout.
The following individuals contributed to the development of this relief request:
Jim McArdle (Principle NDE Level III Inspector) provided UT related information for Sections III through VI Steve Mays (Mechanical Engineer) provided information for Section VIII Andy Hogge (Sponsor) compiled the remaining sections 112911 Sponsored By:
Approved By:
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Attachment A Attachment B Attachment C Attachment D Attachment E Attachment F Attachment G Attachment H Attachment I Attachment J Attachment K Attachment L Attachment M Attachment N Attachment P Component ID Number 2SGA-INLET Examination Data EOC12 Component ID Number 2SGA-INLET Examination Data EOC13 Component ID Number 2SGA-OUTLET Examination Data EOC12 Component ID Number 2SGA-OUTLET Examination Data EOC 13 Component ID Number 2SGD-INLET Examination Data EOC12 Component ID Number 2SGD-INLET Examination Data EOC13 Component ID Number 2SGD-OUTLET Examination Data EOC12 Component ID Number 2SGD-OUTLET Examination Data EOC13 Weld ID Number 2RPV-101-141 Examination Data Weld ID Number 2RPV-10I-142A Examination Data Weld ID Number 2RPV-101-142B Examination Data Weld ID Number 2RPV-101-142C Examination Data Weld ID Number 2RPV-101-151 Examination Data Weld ID Number 2PZR-W3 Examination Data Weld ID Number 2PZR-W4A Examination Data Weld ID Number 2PZR-W4B Examination Data
Request for Relief 05-CN-003 Revision 0 Page 10 of 10 Attachment Q Attachment R Attachment S Attachment T Weld ID Number 2SGC-04B-05 Examination Data Weld ID Number 2FW76-6 Examination Data Weld ID Number 2NV34-11 Examination Data D. M. Jamil to U. S. Nuclear Regulatory Commission letter dated February 24, 2005, Duke Energy Corporation Catawba Nuclear Station Unit 2 Docket No. 50-414 Response to NRC Bulletin 2004-01 Inspection of Alloy 82/182/600 Materials Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping Connections at Pressurizer-water Reactors