View source for RS-08-128, Relief Request CR-26, Inservice Inspection Program Relief Regarding Examination Coverage for Second 10-Year Inservice Inspection Interval

{{Adams
| number = ML082760276
| issue date = 10/01/2008
| title = Relief Request CR-26, Inservice Inspection Program Relief Regarding Examination Coverage for Second 10-Year Inservice Inspection Interval
| author name = Simpson P R
| author affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000373, 05000374
| license number = NPF-011, NPF-018
| contact person = 
| case reference number = RS-08-128
| document type = Letter type:RS
| page count = 16
| project = 
| stage = Other
}}

=Text=
{{#Wiki_filter:Exelon Generation www.exeloncorp.co m 4300 Winfield Road Warrenville, IL 60555 RS-08-128 October 1, 2008 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374 

==Subject:==
Relief Request CR-26, Inservice Inspection Program Relief Regarding Examination Coverage for Second 10-Year Inservice Inspection Interval Exelon 10 CFR 50.55a In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (g)(5)(iii), Exelon Generation Company, LLC (EGC), requests NRC approval to use a proposed alternative to the existing American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," on the basis that compliance with the specified requirements is impractical due to plant design. EGC is submitting this relief request for those ASME Section XI weld examinations performed during the Second 10-Year Inservice Inspection (ISI) Interval where the inspection coverage achieved was less than or equal to 90%. The reduced examination coverage is due to the original design of these welds, physical obstructions, and geometric interference. The relief request supports the second ISI interval for both Unit 1 and Unit 2. The second ISI interval began on November 23, 1994 and ended October 1, 2007. Nuclear This letter contains no regulatory commitments. EGC requests approval of this relief request by October 1, 2009. If you have any questions regarding this letter, please contact Ms. Tricia A. Mattson at (630) 657-2813. Patrick R. Simpson Manager - Licensing Attachment
: Relief Request CR-26  
: 1. ASME Code Components Affected 2. Applicable Code Edition an d Addenda 3. Applicable Code. Requirement 
: 4. ImpracticalitYof Compliance ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 1 of 15 Code Class: 1 and 2 References
: IWB-2500, Table IWB-2500-1, Code Case N-509, IWC-2500, Table IWC-2500-1 Examination Category: B-A, B-D, B-H, B-K-1, C-B, C-C, C-G Item Number: 81.12, 131.21, 131.22, 81.30, 131.40, B3.90, 138.10, 810.10, 810.20, C2.21, C3.10, C3.20, C6.10 Description
: Limited Examination Coverage Component Numbers: See attached Table CR-26.1 and Table CR-26.2 for Component IDs The applicable American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," is the 1989 edition which requires essentially 100% examination coverage of required surfaces and volumes of various Class 1 and 2 components. Section XI, 1989 Edition requires a volumetric and/or surface examination, which includes essentially 100% of the weld and applicable base metal, for the affected examination categories. Additionally, at the time of these limited examinations, LaSalle County Station was using Code Case N-509 as an alternative to the requirements of Tables IWB-2500-1 Categories B-H and B-K-1 and IWC-2500-1, Category C-C of the 1989 Edition of ASME Section XI. Code Case N-509, Table IWB-2500-1 and IWC-2500-1 Categories/Item Numbers B-K, 1310.10, B10.20; and C-C, C3.10 & C3.20 require a surface examination over 100% of the weld of each integrally welded attachment. LCS invoked ASME Section XI Code Case N-460, "Alternative Examination Coverage for Class 1 and Class 2 welds, Section XI Division 1." Code Case N-460 states in part, "...when the entire examination volume or area cannot be examined ... a reduction in examination coverage ... may be accepted provided the reduction in coverage for that weld is less than 10%." NRC Information Notice 98-42, "Implementation of 10 CFR 50.55a(g) Inservice Inspection Requirements," termed the reduction in coverage of less than 10% to be "essentially 100 percent." Information Notice 98-42 states in part, "The NRC has adopted and further refined the definition of "essentially 100 percent" to mean "greater than 90 percent" ... has been applied to all examinations of welds or other areas required by ASME Section XI." Pursuant to 10 CFR 50.55a(g)(5)(iii), relief is being requested on the basis that compliance with the specified Code requirement has been determined to be impractical. Due to the original design of these welds, it is not feasible to effectively perform examinations of 100% of the volume and/or surface area of the welds. Therefore, relief is requested on the basis ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 2 of 15 that the Code requirements to examine essentially 100% of the welds' volume and/or surface area is impractical due to physical obstructions and geometric interference. To perform volumetric and/or surface examination of these welds would require major hardware modifications without a corresponding increase in the level of quality and safety. Table CR-26.1 and CR-26.2 indicate the outage these welds were examined in and the coverage percentages obtained for Unit 1 and Unit 2, respectively. Based on the above explanation, LCS requests relief on the performance of these examinations without achieving 100% volume and/or surface area coverage when 100% coverage is impractical. 5. Burden Caused. tar Compliance Compliance with the examination requirements of ASME Section XI would require modification of plant components to remove obstructions, redesign of plant systems, and replacement of components where geometry is inherent to component design. 6. Proposed_
Alternative_
and Basis for Use In accordance with 10 CFR 50.55a(g)(iii), relief is requested on the basis that the required "essentially 100%" coverage examination is impractical due to physical obstructions and limitations imposed by design, geometry and materials of construction for the components of Table CR-26.1 and Table CR-26.2. LCS proposes to perform the Code required volumetric and/or surface examinations to the maximum extent possible. Because of the design of these welds, there are no alternative examination techniques currently available to increase the examination volume. All components received as a minimum, the required examination(s) applicable to the extent practical due to the limited or lack of access available. The examinations conducted, confirmed satisfactory results evidencing no unacceptable flaws present, even though essentially 100% coverage was not attained. Additionally, a VT-2 examination performed on the subject components during the system pressure test per examination category B-P each refueling outage and category C-H each period provides additional assurance that the structural integrity of the subject components is maintained. 7. Duration of Proposed Alternative Relief is requested for the second ten-year inspection interval of the Inservice Inspection Interval (ISI) Program for LCS Units 1 and 2. The second ISI interval began on November 23, 1994 and ended October 1, 2007.
ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 3 of 15 At the end of the first fuel cycle in the first interval, Unit 1 had a refueling outage that lasted 329 consecutive days. Paragraph IWA-2430(e) of ASME Section XI allows an inspection interval to be extended for a period of time equivalent to the duration of a continuous outage that exceeds six months. The start date of the Unit 1 second inspection interval noted above reflects an extension of the first inspection interval by 329 days. The second ISI interval is divided into three successive inspection periods as determined by calendar years of plant service within the inspection interval. Identified below are the period dates for the second ISI interval. In accordance with Paragraph IWB-2412(b) of ASME Section XI, the inspection periods specified below may be decreased or extended by as much as one year to enable inspections to coincide with LCS's refueling outages. Unit 1, Period 1 (November 23, 1994 through October 11, 1999) Unit 1, Period 2 (October 12, 1999 through October 11, 2003) Unit 1, Period 3 (October 12, 2003 through October 11, 2006) Unit 2, Period 1 (October 17, 1994 through July 4, 2000) Unit 2, Period 2 (July 5, 2000 through July 4, 2004) Unit 2, Period 3 (July 5, 2004 through July 4, 2007) Period 1 for Unit 1 reflects a 690 day extension allowed by IWA-2430(e) for a forced outage (i.e., L1 F35) and an extension per IWB-2412(b) to allow completion of ongoing system pressure testing. System pressure testing was completed on November 19, 1999 for Unit 1. Period 1 for Unit 2 reflects a 932 day extension allowed by IWA-2430(e) for an extended refueling outage (i.e., L 9R07) and an extension per IWB-2412(b) to allow completion of ongoing system pressure testing. System pressure testing was completed and on November 20, 2000 for Unit 2. Period 2 for Unit 2 was extended per the ASME Code IWB-2412(b) due to deferral of Category B-D nozzle to vessel welds. As a result of these extensions of the second ISI intervals as shown in the table below, the date to start both Unit 1 and Unit 2 ISI Third Interval was on October 01, 2007. The extensions are all within the required maximum of one year per IWA-2430(d). 8. Precedents The NRC has previously approved similar relief for Quad Cities Nuclear Power Station, Units 1 and 2, and Dresden Nuclear Power Station, Units 2 and 3. The NRC granted relief for Quad Cities Nuclear Power Station in Reference 1, and the Dresden Nuclear Power Station in Reference 
: 2. Unit No. Pro ram Inclusive Dates Days Extended 1 ISI Oct. 11,_200_6 to Set. 30, 2_007_ 354 Da y s 2 ISI July 04, 2007 to Sept. 30, 2007 88 Days  
: 9. References ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 4 of 15 1. Letter from U.S. NRC to O. D. Kingsley (Commonwealth Edison Company), "Quad Cities, Units 1 and 2 - Relief Request CR-32 for Third 10-Year Inservice Inspection Interval," dated September 6, 2000 2. Letter from U.S. NRC to C. M. Crane (Exelon Generation Company, LLC), "Dresden Nuclear Power Station, Units 2 and 3 - Relief Request CR-26 for Third 10-Year Inservice Inspection Interval," dated October 1, 2004 ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 5 of 15 Table CR-26.1 LASALLE COUNTY STATION Unit 1 List of Welds with Limited Examination Coverage Component ID Outage Actual Remarks Examined Coverage GEL-1009-AG L1 R07 71.90% RPV Head to Flange 0-180 Deg. limited UT scanning from nozzle side of weld due to nozzle design. GEL-1009-DM L1 R07 77.20% RPV Head Meridional limited UT scanning due to lifting lugs. GEL-1009-DP L1 R07 77.20% RPV Head Meridional limited UT scanning due to lifting lugs. LCS-1-AF L1 R07 55.40% RPV Shell to Flange limited UT scanning from Flange side of weld. LCS-1-NIA L1 R07 57.94% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N1 B L1 R07 58.60% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N3A L1 R07 59.95% Main Steam nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N3B L1 R07 59.95% Main Steam nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N3C L1 R07 59.95% Main Steam nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N3D L1 R07 59.95% Main Steam nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design.
ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 6 of 15 Component ID Outage Actual Remarks Examined Coverage RH-HX1 B-1 L1 R07 80.00% Residual Heat Removal Heat Exchanger nozzle to Head limited UT scanning from nozzle side of weld due to nozzle design. HP-PUl-05 L1 R07 72.00% High Pressure Core Spray Pump PT limited due to design of pump base support. LCS-1-N6A L1 R08 69.5% Residual Heat Removal nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N6C L1 R08 69.5% Residual Heat Removal nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N10 L1 R09 72.4% Control Rod Drive nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N2B L1 R09 70.9% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N2C L1 R09 70.9% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N2D L1 R09 70.9% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N2E L1 R09 70.9% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N2F L1 R09 70.9% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N2J L1 R09 70.9% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design.
ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested in Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 7 of 15 Component ID Outage Actual Remarks Examined Coverage LCS-1-N2K L1 R09 70.9% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N4A L1 R09 72.3% Feedwater nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N4B L1 R09 72.3% Feedwater nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N4C L1 R09 72.3% Feedwater nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N4D L1 R09 72.3% Feedwater nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N4E L1 R09 72.3% Feedwater nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N4F L1 R09 72.3% Feedwater nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. IRH-HX1 B-08A L1 R09 76% Residual Heat Removal Heat Exchanger Attachment limited MT/PT due to structural steel. GEL-1009-AG L1 R11 49% RPV Flange to Head 180-360 Deg. limited UT scanning from Flange side of weld. GEL-1006-AJ L1 R11 19% RPV to Bottom Head Weld limited UT scanning from bottom side due to RPV Skirt. LCS-1-BG L1 R11 75.3% Reactor Pressure Vessel Vertical Weld limited UT scanning due to proximity of instrument nozzle to RPV weld. LCS-1-13H LIR11 88.8% Reactor Pressure Vessel Vertical Weld limited UT scanning due to proximity of thermocouple attachments to RPV.
ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 8 of 15 Component ID Outage Actual Remarks Examined Coverage GEL-1009-DR L1 R11 68% RPV Head Meridional limited UT scanning due to lifting lugs. GEL-1009-DT L1 R11 68% RPV Head Meridional limited UT scanning due to lifting lugs. LCS-1-N2A L1 R11 83.1% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N2G L1 R11 83.1% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N2H L1 R11 83.1% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N5 L1R11 85.5% Low Pressure Core Spray nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N6B L1 R11 83.7% Residual Heat Removal nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N7 L1R11 87% Reactor Core Isolation Cooling nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N8 L1 R11 79% Reactor Head Vent nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N16 L1 R11 85.5% High Pressure Core Spray nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-1-N18 L1 R11 87% Spare nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. RPV-SS-1 L1 R11 66% RPV Stabilizer Lug limited MT/PT due to structural steel.
ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 9 of 15 Component ID Outage Actual Remarks Examined Coverage IHP-PUl-04 L1 R11 16% High Pressure Core Spray Pump Weld MT limited due to design of um base su " ort. ILP-PU-04 L1 R11 24% Low Pressure Core Spray Pump Weld MT limited due to design of um base support. IRH-PU1 C-04 L1 R11 21% Residual Heat Removal Pump Weld MT limited due to design of um base su q ort. IHP-PU1-7A L1 R11 0% High Pressure Core Spray Pump Weld encased in concrete. IHP-PU1-7B L1 R11 0% High Pressure Core Spray Pump Weld encased in concrete. IHP-PU1-7C L1 R11 0% High Pressure Core Spray Pump Weld encased in concrete. IHP-PU1-8A L1 R11 0% High Pressure Core Spray Pump Weld encased in concrete. IHP-PU1-8B L1 R11 0% High Pressure Core Spray Pump Weld encased in concrete. IHP-PU1-8C L1 R11 0% High Pressure Core Spray Pump Weld encased in concrete. IHP-PU1-9 L1 R11 0% High Pressure Core Spray Pump Weld encased in concrete. ILP-PU1-7 L1R11 0% Low Pressure Core Spray Pump Weld encased in concrete. ILP-PU1-8 L1 R11 0% Low Pressure Core Spray Pump Weld encased in concrete. ILP-PU1-9 L1R11 0% Low Pressure Core Spray Pump Weld encased in concrete. IRH-PU1C-7A L1R11 0% Residual Heat Removal Pump Weld encased in concrete. IRH-PU1C-7B L1R11 0% Residual Heat Removal Pump Weld encased in concrete. IRP-PU1C-7C L1R11 0% Residual Heat Removal Pump Weld encased in concrete. IRH-PU1C-8A L1R11 0% Residual Heat Removal Pump Weld encased in concrete.
ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 10 of 15 Component ID Outage Actual Remarks Examined Coverage IRH-PU1C-8B L1R11 0% Residual Heat Removal Pump Weld encased in concrete. IRH-PU1C-8C L1R11 0% Residual Heat Removal Pump Weld encased in concrete. IRH-PU1C-9 L1R11 0% Residual Heat Removal Pump Weld encased in concrete.
ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 11 of 15 Table CR-26.2 LASALLE COUNTY STATION Unit 2 List of Welds with Limited Examination Coverage Component ID Outage Actual Remarks Examined Coverage GEL-1060-AG L2R07 65.3% RPV Head to Flange 0-180 Deg. limited UT scanning from Flange side of weld due to flange configuration. LCS-2-AE L2R07 54.6% RPV Shell to Flange 180 Deg. t o 360 Deg. limited UT scanning from Flange side of weld due to flange configuration. LCS-2-N1 A L2R07 64.7% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N3C L2R07 62.1% Main Steam nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N3D L 2R07 62.1% Main Steam nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N7 L 2R07 63.2% Reactor Core Isolation Cooling nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N8 L 9R07 64.1% Reactor Head Vent nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N10A L2R07 68% Capped Control Rod Drive Return nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle desi n. LCS-2-N18 L 2R07 63.2% Spare nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N1 B L9R08 60.4% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design.
ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 12 of 15 Component ID Outage Actual Remarks Examined Coverage LCS-2-N6A L 9R08 68.4% Residual Heat Removal nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N6C L2R08 68.4% Residual Heat Removal nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N4A L 2R09 78% Feedwater nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N413 L 2R09 78% Feedwater nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N4C L2R09 78% Feedwater nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N4D LPR09 78% Feedwater nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N4E L 2R09 78% Feedwater nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N4F 1-21109 78% Feedwater nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. IRH-HX2B-08D L 2R09 74% Residual Heat Removal Heat Exchanger Attachment limited MT/PT due to structural steel. LCS-2-AE LPR10 58% RPV Shell to Flange 0 Deg. t o 180 Deg. limited UT scanning from Flange side of weld due to configuration. GEL-1061-DA L2R10 66% RPV Bottom Head Meridional Weld limited UT scanning due to RPV Skirt. GEL-1061-DE L2R10 66% RPV Bottom Head Meridional Weld limited UT scanning due to RPV Skirt. GEL-1061-DF L2R10 66% RPV Bottom Head Meridional Weld limited UT scanning due to RPV Skirt.
ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 13 of 15 Component ID Outage Actual Remarks Examined Coverage LCS-2-N2B LPR10 70% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N2C LPR10 70% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N2F L2R10 70% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N2G LPR10 70% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N2H LPR10 70% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N2J LPR10 70% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N2K L2R10 70% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N3A L2R10 68% Main Steam nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N3B L2R10 68% Main Steam nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. 2RPV-SS-2 LPR10 52% Reactor Pressure Vessel Stabilizer Lug limited MT/PT due to structural steel. GEL-1060-AG L 9R11 80.6% RPV Top Head to Flange 180 to 360 Deg. limited UT scanning from nozzle side of weld due to nozzle design.
ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 14 of 15 Component ID Outage Actual Remarks Examined Coverage LCS-2-N16A L2R11 70.5% High Pressure Core Spray nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle desi n. LCS-2-N2A L2R11 70.5% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N2D LPR11 70.3% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N2E 12R11 70.5% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N5A 12R11 70.5% Low Pressure Core Spray nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N6B L2R11 70.5% Reactor Recirculation nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N9A L 2R11 82.9% Jet Pump Instrument nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. LCS-2-N9B L 9R11 82.9% Jet Pump Instrument nozzle to RPV limited UT scanning from nozzle side of weld due to nozzle design. IHP-PU2-7A I PR11 0% High Pressure Core Spray Pump Weld encased in concrete. IHP-PU2-7B LPR11 0% High Pressure Core Spray Pump Weld encased in concrete. IHP-PU2-7C L2R11 0% High Pressure Core Spray Pump Weld encased in concrete. IHP-PU2-8A 1-21111 0% High Pressure Core Spray Pump Weld encased in concrete.
ATTACHMENT 10 CFR 50.55a Request Number CR-26 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)
Page 15 of 15 Component ID Outage Actual Remarks Examined Coverage IHP-PU2-8B L2R11 0% High Pressure Core Spray Pump Weld encased in concrete. IHP-PU2-8C L2R11 0% High Pressure Core Spray Pump Weld encased in concrete. IHP-PU2-9 L 2R11 0% High Pressure Core Spray Pump Weld encased in concrete. ILP-PU2-7 L2R11 0% Low Pressure Core Spray Pump Weld encased in concrete. ILP-PU2-8 LPR11 0% Low Pressure Core Spray Pump Weld encased in concrete. ILP-PU2-9 L2R11 0% Low Pressure Core Spray Pump Weld encased in concrete. IRH-PU2C-7A L2R11 0% Residual Heat Removal Pump Weld encased in concrete. IRH-PU2C-7B L2R11 0% Residual Heat Removal Pump Weld encased in concrete. IRP-PU2C-7C L 2R11 0% Residual Heat Removal Pump Weld encased in concrete. IRH-PU2C-8A L2R11 0% Residual Heat Removal Pump Weld encased in concrete. IRH-PU2C-8B LPR11 0% Residual Heat Removal Pump Weld encased in concrete. IRH-PU2C-8C L ?R11 0% Residual Heat Removal Pump Weld encased in concrete. IRH-PU2C-9 LPR11 0% Residual Heat Removal Pump Weld encased in concrete.}}